WorldWideScience

Sample records for range safety criteria

  1. Laser Safety Inspection Criteria

    International Nuclear Information System (INIS)

    Barat, K

    2005-01-01

    A responsibility of the Laser Safety Officer (LSO) is to perform laser safety audits. The American National Standard Z136.1 Safe use of Lasers references this requirement in several sections: (1) Section 1.3.2 LSO Specific Responsibilities states under Hazard Evaluation, ''The LSO shall be responsible for hazards evaluation of laser work areas''; (2) Section 1.3.2.8, Safety Features Audits, ''The LSO shall ensure that the safety features of the laser installation facilities and laser equipment are audited periodically to assure proper operation''; and (3) Appendix D, under Survey and Inspections, it states, ''the LSO will survey by inspection, as considered necessary, all areas where laser equipment is used''. Therefore, for facilities using Class 3B and or Class 4 lasers, audits for laser safety compliance are expected to be conducted. The composition, frequency and rigueur of that inspection/audit rests in the hands of the LSO. A common practice for institutions is to develop laser audit checklists or survey forms. In many institutions, a sole Laser Safety Officer (LSO) or a number of Deputy LSO's perform these audits. For that matter, there are institutions that request users to perform a self-assessment audit. Many items on the common audit list and the associated findings are subjective because they are based on the experience and interest of the LSO or auditor in particular items on the checklist. Beam block usage is an example; to one set of eyes a particular arrangement might be completely adequate, while to another the installation may be inadequate. In order to provide more consistency, the National Ignition Facility Directorate at Lawrence Livermore National Laboratory (NIF-LLNL) has established criteria for a number of items found on the typical laser safety audit form. These criteria are distributed to laser users, and they serve two broad purposes: first, it gives the user an expectation of what will be reviewed by an auditor, and second, it is an

  2. Laser Safety Inspection Criteria

    International Nuclear Information System (INIS)

    Barat, K.

    2005-01-01

    A responsibility of the Laser Safety Officer (LSO) is to perform laser audits. The American National Standard Z136.1 Safe Use of Lasers references this requirement through several sections. One such reference is Section 1.3.2.8, Safety Features Audits, ''The LSO shall ensure that the safety features of the laser installation facilities and laser equipment are audited periodically to assure proper operation''. The composition, frequency and rigor of that inspection/audit rests in the hands of the LSO. A common practice for institutions is to develop laser audit checklists or survey forms It is common for audit findings from one inspector or inspection to the next to vary even when reviewing the same material. How often has one heard a comment, ''well this area has been inspected several times over the years and no one ever said this or that was a problem before''. A great number of audit items, and therefore findings, are subjective because they are based on the experience and interest of the auditor to particular items on the checklist. Beam block usage, to one set of eyes might be completely adequate, while to another, inadequate. In order to provide consistency, the Laser Safety Office of the National Ignition Facility Directorate has established criteria for a number of items found on the typical laser safety audit form. The criteria are distributed to laser users. It serves two broad purposes; first, it gives the user an expectation of what will be reviewed by an auditor. Second, it is an opportunity to explain audit items to the laser user and thus the reasons for some of these items, such as labelling of beam blocks

  3. Safety and reliability criteria

    International Nuclear Information System (INIS)

    O'Neil, R.

    1978-01-01

    Nuclear power plants and, in particular, reactor pressure boundary components have unique reliability requirements, in that usually no significant redundancy is possible, and a single failure can give rise to possible widespread core damage and fission product release. Reliability may be required for availability or safety reasons, but in the case of the pressure boundary and certain other systems safety may dominate. Possible Safety and Reliability (S and R) criteria are proposed which would produce acceptable reactor design. Without some S and R requirement the designer has no way of knowing how far he must go in analysing his system or component, or whether his proposed solution is likely to gain acceptance. The paper shows how reliability targets for given components and systems can be individually considered against the derived S and R criteria at the design and construction stage. Since in the case of nuclear pressure boundary components there is often very little direct experience on which to base reliability studies, relevant non-nuclear experience is examined. (author)

  4. Panel 1: Safety design criteria

    International Nuclear Information System (INIS)

    Yllera, Javier

    2013-01-01

    There is general consensus in the nuclear community, and more after the Fukushima accident, that the deployment of nuclear energy has to be done at the highest levels of nuclear safety and that safety cannot be compromised by other factors. It is well understood that reactors that are being licensed and the new generations of reactors that will be constructed in the future will need to reach higher safety levels than the existing ones. Several countries and international organizations or international groups are launching initiatives to harmonise safety goals, safety requirements, safety objectives, regulations, criteria or safety reference levels. There are differences in the meanings of these terms and the working approaches, but the overall purpose is the same: to specify how new plants can be safer. In this context, the IAEA has an statutory function for developing international nuclear safety standards. The IAEA safety standards are per se not mandatory for IAEA Member States. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA’s standards for use in their national regulations in different ways. The IAEA Safety Standards represent international consensus on what must constitute a high level of safety for nuclear installations. In the area of NPP design, IAEA safety standards that are published are intended to apply primarily to new plants. It might not be practicable to apply all the requirements to plants that are already in operation. In addition, the focus is primarily on plants with water cooled reactors

  5. Range Flight Safety Requirements

    Science.gov (United States)

    Loftin, Charles E.; Hudson, Sandra M.

    2018-01-01

    The purpose of this NASA Technical Standard is to provide the technical requirements for the NPR 8715.5, Range Flight Safety Program, in regards to protection of the public, the NASA workforce, and property as it pertains to risk analysis, Flight Safety Systems (FSS), and range flight operations. This standard is approved for use by NASA Headquarters and NASA Centers, including Component Facilities and Technical and Service Support Centers, and may be cited in contract, program, and other Agency documents as a technical requirement. This standard may also apply to the Jet Propulsion Laboratory or to other contractors, grant recipients, or parties to agreements to the extent specified or referenced in their contracts, grants, or agreements, when these organizations conduct or participate in missions that involve range flight operations as defined by NPR 8715.5.1.2.2 In this standard, all mandatory actions (i.e., requirements) are denoted by statements containing the term “shall.”1.3 TailoringTailoring of this standard for application to a specific program or project shall be formally documented as part of program or project requirements and approved by the responsible Technical Authority in accordance with NPR 8715.3, NASA General Safety Program Requirements.

  6. Common Risk Criteria Standards for National Test Ranges

    Science.gov (United States)

    2016-08-01

    supplemental) document to RCC Document 321. a. Modified aircraft vulnerability criteria for business class jets. b. Modified the aircraft vulnerability... successful , the logical relationships among criteria used at the test ranges and across different hazards are often difficult to comprehend. The...provides a common set of range safety policies, risk criteria, and guidelines for managing risk to people and assets during manned and unmanned

  7. NSSS supplier's response to differing safety criteria

    Energy Technology Data Exchange (ETDEWEB)

    Cremades, J; Filkin, R; Franke, T [Westinghouse Electric Nuclear Energy Systems Europe (WENESE), Brussels (Belgium)

    1980-11-01

    The limited progress achieved to date in harmonizing national criteria has led to the development of designs which include the most common national requirements. Progress towards harmonization of safety criteria can be accelerated by expanding the IAEA leadership and co-ordination activities, and implementing an integrated approach to criteria development. National and International safety criteria are examined.

  8. Ferrocyanide Safety Program: Safety criteria for ferrocyanide watch list tanks

    International Nuclear Information System (INIS)

    Postma, A.K.; Meacham, J.E.; Barney, G.S.

    1994-01-01

    This report provides a technical basis for closing the ferrocyanide Unreviewed Safety Question (USQ) at the Hanford Site. Three work efforts were performed in developing this technical basis. The efforts described herein are: 1. The formulation of criteria for ranking the relative safety of waste in each ferrocyanide tank. 2. The current classification of tanks into safety categories by comparing available information on tank contents with the safety criteria; 3. The identification of additional information required to resolve the ferrocyanide safety issue

  9. TAPS safety evaluation criteria for reload fueling

    International Nuclear Information System (INIS)

    Mahendra Nath; Veeraraghavan, N.

    1976-01-01

    To improve operating performance of Tarapur reactors, several proposals are under consideration such as core expansion, change-over to an improved fuel design with lower heat rating, extension of fuel cycle lengths etc., which have a bearing on overall plant operating characteristics and reactor safety. For evaluating safety implications of the various proposals, it is necessary to formulate safety evaluation criteria for reload fuelling. Salient features of these criteria are discussed. (author)

  10. Safety criteria of uranium enrichment plants

    International Nuclear Information System (INIS)

    Nardocci, A.C.; Oliveira Neto, J.M. de

    1994-01-01

    The applicability of nuclear reactor safety criteria applied to uranium enrichment plants is discussed, and a new criterion based on the soluble uranium compounds and hexafluoride chemical toxicities is presented. (L.C.J.A.). 21 refs, 4 tabs

  11. The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

  12. Criteria for safety-related operator actions

    International Nuclear Information System (INIS)

    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

  13. Squale: evaluation criteria of functioning safety

    International Nuclear Information System (INIS)

    Deswarte, Y.; Kaaniche, M.; Benoit, P.

    1998-05-01

    The SQUALE (security, safety and quality evaluation for dependable systems) project is part of the ACTS (advanced communications, technologies and services) European program. Its aim is to develop confidence evaluation criteria to test the functioning safety of systems. All industrial sectors that use critical applications (nuclear, railway, aerospace..) are concerned. SQUALE evaluation criteria differ from the classical evaluation methods: they are independent of the application domains and industrial sectors, they take into account the overall functioning safety attributes, and they can progressively change according to the level of severity required. In order to validate the approach and to refine the criteria, a first experiment is in progress with the METEOR automatic underground railway and another will be carried out on a telecommunication system developed by Bouygues company. (J.S.)

  14. Review of fuel safety criteria in France

    Energy Technology Data Exchange (ETDEWEB)

    Boutin, Sandrine; Graff, Stephanie; Foucher-Taisne, Aude; Dubois, Olivier [Institut de Radioprotection et du Surete Nucleaire, Fontenay-aux-Roses (France)

    2018-01-15

    Fuel safety criteria for the first barrier, based on state-of-the-art at the time, were first defined in the 1970s and came from the United States, when the French nuclear program was initiated. Since then, there has been continuous progress in knowledge and in collecting experimental results thanks to the experiments carried out by utilities and research institutes, to the operating experience, as well as to the generic R and D programs, which aim notably at improving computation methodologies, especially in Reactivity-Initiated accident and Loss-of-Coolant Accident conditions. In this context, the French utility EDF proposed new fuel safety criteria, or reviewed and completed existing safety demonstration covering the normal operating, incidental and accidental conditions of Pressurised Water Reactors. IRSN assessed EDF's proposals and presented its conclusions to the Advisory Committee for Reactors Safety of the Nuclear Safety Authority in June 2017. This review focused on the relevance of historical limit values or parameters of fuel safety criteria and their adequacy with the state-of-the-art concerning fuel physical phenomena (e.g. Pellet-Cladding Mechanical Interaction in incidental conditions, clad embrittlement due to high temperature oxidation in accidental conditions, clad ballooning and burst during boiling crisis and fuel melting).

  15. Review of fuel safety criteria in France

    International Nuclear Information System (INIS)

    Boutin, Sandrine; Graff, Stephanie; Foucher-Taisne, Aude; Dubois, Olivier

    2018-01-01

    Fuel safety criteria for the first barrier, based on state-of-the-art at the time, were first defined in the 1970s and came from the United States, when the French nuclear program was initiated. Since then, there has been continuous progress in knowledge and in collecting experimental results thanks to the experiments carried out by utilities and research institutes, to the operating experience, as well as to the generic R and D programs, which aim notably at improving computation methodologies, especially in Reactivity-Initiated accident and Loss-of-Coolant Accident conditions. In this context, the French utility EDF proposed new fuel safety criteria, or reviewed and completed existing safety demonstration covering the normal operating, incidental and accidental conditions of Pressurised Water Reactors. IRSN assessed EDF's proposals and presented its conclusions to the Advisory Committee for Reactors Safety of the Nuclear Safety Authority in June 2017. This review focused on the relevance of historical limit values or parameters of fuel safety criteria and their adequacy with the state-of-the-art concerning fuel physical phenomena (e.g. Pellet-Cladding Mechanical Interaction in incidental conditions, clad embrittlement due to high temperature oxidation in accidental conditions, clad ballooning and burst during boiling crisis and fuel melting).

  16. Autonomous system for launch vehicle range safety

    Science.gov (United States)

    Ferrell, Bob; Haley, Sam

    2001-02-01

    The Autonomous Flight Safety System (AFSS) is a launch vehicle subsystem whose ultimate goal is an autonomous capability to assure range safety (people and valuable resources), flight personnel safety, flight assets safety (recovery of valuable vehicles and cargo), and global coverage with a dramatic simplification of range infrastructure. The AFSS is capable of determining current vehicle position and predicting the impact point with respect to flight restriction zones. Additionally, it is able to discern whether or not the launch vehicle is an immediate threat to public safety, and initiate the appropriate range safety response. These features provide for a dramatic cost reduction in range operations and improved reliability of mission success. .

  17. Safety criteria for nuclear chemical plants

    International Nuclear Information System (INIS)

    Ball, P.W.; Curtis, L.M.

    1983-01-01

    Safety measures have always been required to limit the hazards due to accidental release of radioactive substances from nuclear power plants and chemical plants. The risk associated with the discharge of radioactive substances during normal operation has also to be kept acceptably low. BNFL (British Nuclear Fuels Ltd.) are developing risk criteria as targets for safe plant design and operation. The numerical values derived are compared with these criteria to see if plants are 'acceptably safe'. However, the criteria are not mandatory and may be exceeded if this can be justified. The risk assessments are subject to independent review and audit. The Nuclear Installations Inspectorate also has to pass the plants as safe. The assessment principles it uses are stated. The development of risk criteria for a multiplant site (nuclear chemical plants tend to be sited with many others which are related functionally) is discussed. This covers individual members of the general public, societal risks, risks to the workforce and external hazards. (U.K.)

  18. A NSSS supplier's response to differing safety criteria

    International Nuclear Information System (INIS)

    Cremades, J.; Filkin, R.; Franke, Th.

    1980-01-01

    The limited progress achieved to date in harmonizing national criteria has led to the development of designs which include the most common national requirements. Progress towards harmonization of safety criteria can be accelerated by expanding the IAEA leadership and co-ordination activities, and implementing an integrated approach to criteria development. National and International safety criteria are examined. (author)

  19. Radiation protection criteria in the long-range view

    International Nuclear Information System (INIS)

    Snihs, J.O.; Bergman, C.

    1989-01-01

    The report presents by way of introduction radiation protection criteria applied to radiological activities and to disposal of low-level and intermediate-level radioactive waste. In these cases it is primarily short-range views that are relevant, up to a few thousand years as a maximum. In the case of high-level wastes where the views may extend to more than hundreds of thousands years, there are not for the present any equally well stablished criteria. Based upon preliminary results from a Nordic team for criteria for high-level radioactive wastes, dose estimates in the long-range view and alternative assessment criteria are discussed. Proposals are also presented for 12 criteria that may be applicable. As the work is not yet finshed, the criteria are however merely preliminary

  20. Probabilistic safety criteria at the safety function/system level

    International Nuclear Information System (INIS)

    1989-09-01

    A Technical Committee Meeting was held in Vienna, Austria, from 26-30 January 1987. The objectives of the meeting were: to review the national developments of PSC at the level of safety functions/systems including future trends; to analyse basic principles, assumptions, and objectives; to compare numerical values and the rationale for choosing them; to compile the experience with use of such PSC; to analyse the role of uncertainties in particular regarding procedures for showing compliance. The general objective of establishing PSC at the level of safety functions/systems is to provide a pragmatic tool to evaluate plant safety which is placing emphasis on the prevention principle. Such criteria could thus lead to a better understanding of the importance to safety of the various functions which have to be performed to ensure the safety of the plant, and the engineering means of performing these functions. They would reflect the state-of-the-art in modern PSAs and could contribute to a balance in system design. This report, prepared by the participants of the meeting, reviews the current status and future trends in the field and should assist Member States in developing their national approaches. The draft of this document was also submitted to INSAG to be considered in its work to prepare a document on safety principles for nuclear power plants. Five papers presented at the meeting are also included in this publication. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  1. Safety Criteria for the Private Spaceflight Industry

    Science.gov (United States)

    Quinn, Andy; Maropoulos, Paul

    2010-09-01

    The Federal Aviation Administration(FAA) Office of Commercial Space Transportation(AST) has set specific rules and generic guidelines to cover experimental and operational flights by industry forerunners such as Virgin Galactic and XCOR. One such guideline Advisory Circular(AC) 437.55-1[1] contains exemplar hazard analyses for spacecraft designers and operators to follow under an experimental permit. The FAA’s rules and guidelines have also been ratified in a report to the United States Congress, Analysis of Human Space Flight Safety[2] which cites that the industry is too immature and has ‘insufficient data’ to be proscriptive and that ‘defining a minimum set of criteria for human spaceflight service providers is potentially problematic’ in order not to ‘stifle the emerging industry’. The authors of this paper acknowledge the immaturity of the industry and discuss the problematic issues that Design Organisations and Operators now face.

  2. 2011 NASA Range Safety Annual Report

    Science.gov (United States)

    Dumont, Alan G.

    2012-01-01

    Welcome to the 2011 edition of the NASA Range Safety Annual Report. Funded by NASA Headquarters, this report provides a NASA Range Safety overview for current and potential range users. As is typical with odd year editions, this is an abbreviated Range Safety Annual Report providing updates and links to full articles from the previous year's report. It also provides more complete articles covering new subject areas, summaries of various NASA Range Safety Program activities conducted during the past year, and information on several projects that may have a profound impact on the way business will be done in the future. Specific topics discussed and updated in the 2011 NASA Range Safety Annual Report include a program overview and 2011 highlights; Range Safety Training; Range Safety Policy revision; Independent Assessments; Support to Program Operations at all ranges conducting NASA launch/flight operations; a continuing overview of emerging range safety-related technologies; and status reports from all of the NASA Centers that have Range Safety responsibilities. Every effort has been made to include the most current information available. We recommend this report be used only for guidance and that the validity and accuracy of all articles be verified for updates. Once again the web-based format was used to present the annual report. We continually receive positive feedback on the web-based edition and hope you enjoy this year's product as well. As is the case each year, contributors to this report are too numerous to mention, but we thank individuals from the NASA Centers, the Department of Defense, and civilian organizations for their contributions. In conclusion, it has been a busy and productive year. I'd like to extend a personal Thank You to everyone who contributed to make this year a successful one, and I look forward to working with all of you in the upcoming year.

  3. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

    2013-01-01

    • Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

  4. Development of small reactor safety criteria in Canada

    International Nuclear Information System (INIS)

    Ernst, P.C.; French, P.M.; Axford, D.J.; Snell, V.G.

    1990-01-01

    A number of new small reactor designs have been proposed in Canada over the last several years and some have reached the stage where licensing discussions have been initiated with the Atomic Energy Control Board (AECB). An inter-organizational Small Reactor Criteria (SRC) working group was formed in 1988 to propose safety and licensing criteria for these small reactors. Two levels of criteria are proposed. The first level forms a safety philosophy and the second is a set of criteria for specific reactor applications. The safety philosophy consists of three basic safety objectives together with evaluation criteria, and fourteen fundamental principles measured by specific criteria, which must be implemented to meet the safety objectives. Two of the fourteen principles are prime: defence in depth, and safety culture; the other twelve principles can be seen as deriving from them. A benefit of this approach is that the concepts of defence in depth and safety culture become well-defined. The objectives and principles are presented in the paper and their criteria are summarized. The second level of criteria, under development, will form a safety application set and will provide small reactor criteria in a number of general areas, such as regulatory process and safety assessment, as well as for specific reactor life-cycle activities, from siting through to decommissioning. The criteria are largely deterministic. However, the frequencies and consequences of postulated accidents are assessed against numerical criteria to assist in judging the acceptability of plant design, operation, and proposed siting. All criteria proposed are designed to be testable in some evidentiary fashion, readily enabling an assessment of compliance for a given proposal

  5. Licensing procedures and safety criteria for research reactors in France

    International Nuclear Information System (INIS)

    Berry, J.L.; Lerouge, B.

    1980-11-01

    This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. Some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors. At the end, a few considerations are given to the consequences of the Osiris core conversion

  6. Nuclear Fuel Safety Criteria Technical Review - Second edition

    International Nuclear Information System (INIS)

    Beck, Winfried; Blanpain, Patrick; Fuketa, Toyoshi; Gorzel, Andreas; Hozer, Zoltan; Kamimura, Katsuichiro; Koo, Yang-Hyun; Maertens, Dietmar; Nechaeva, Olga; Petit, Marc; Rehacek, Radomir; Rey-Gayo, Jose Maria; Sairanen, Risto; Sonnenburg, Heinz-Guenther; Valach, Mojmir; Waeckel, Nicolas; Yueh, Ken; Zhang, Jinzhao; Voglewede, John

    2012-01-01

    Most of the current nuclear fuel safety criteria were established during the 1960's and early 1970's. Although these criteria were validated against experiments with fuel designs available at that time, a number of tests were based on unirradiated fuels. Additional verification was performed as these designs evolved, but mostly with the aim of showing that the new designs adequately complied with existing criteria, and not to establish new limits. In 1996, the OECD Nuclear Energy Agency (NEA) reviewed existing fuel safety criteria, focusing on new fuel and core designs, new cladding materials and industry manufacturing processes. The results were published in the Nuclear Fuel Safety Criteria Technical Review of 2001. The NEA has since re-examined the criteria. A brief description of each criterion and its rationale are presented in this second edition, which will be of interest to both regulators and industry (fuel vendors, utilities)

  7. The study on safety facility criteria for radioactive waste repository

    International Nuclear Information System (INIS)

    Lee, S. H.; Choi, M. H.; Han, S. H. and others

    1992-12-01

    The radioactive waste repository are necessary to install the engineered safety systems to secure the safety for operation of the repository in the event of fire and earthquake. Since the development of safety facility criteria requires a thorough understanding about the characteristics of the engineered safety systems, we should investigate by means of literature survey and visit SKB. In particular, definition, composition of the systems, functional requirement of the systems, engineered safety systems of foreign countries, system design, operation and maintenance requirement should be investigated : fire protection system, ventilation system, drainage system, I and C system, electric system, radiation monitoring system. This proposed criteria consist of purpose, scope of application, ventilation system, fire protection system, drainage system, electric system and this proposed criteria can be applied as a basic reference for the final criteria

  8. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

  9. Safety Criteria and Standards for Bearing Capacity of Foundation

    Directory of Open Access Journals (Sweden)

    Yanlong Li

    2017-01-01

    Full Text Available This paper focuses on the evaluation standards of factor of safety for foundation stability analysis. The problem of foundation stability is analyzed via the methods of risk analysis of engineering structures and reliability-based design, and the factor of safety for foundation stability is determined by using bearing capacity safety-factor method (BSFM and strength safety-factor method (SSFM. Based on a typical example, the admissible factors of safety were calibrated with a target reliability index specified in relevant standards. Two safety criteria and their standards of bearing capacity of foundation for these two methods (BSFM and SSFM were established. The universality of the safety criteria and their standards for foundation reliability was verified based on the concept of the ratio of safety margin (RSM.

  10. Safety criteria for advanced HTGR concepts

    International Nuclear Information System (INIS)

    Kroeger, W.

    1989-01-01

    It is commonly agreed that advanced HTGR concepts must be licensable, which means that they must fulfil existing regulatory requirements. Furthermore, it is necessary to improve their public acceptance and they must even be suitable for urban sites. Therefore, they should be 'safer' than existing plants, which mainly means with respect to low-frequency or beyond-design severe accidents. Last but not least, the realization of advanced HTGR would be easier if commonly shared safety principles could be stated ensuring this further increased level of safety internationally. These qualitative statements need to be cast into quantitative guidelines which can be used as a rationale for safety evaluation. This paper tries to describe the status reached and to stimulate international activities. (author). 12 refs, 4 figs, 3 tabs

  11. Safety principles and design criteria for nuclear power stations

    International Nuclear Information System (INIS)

    Gazit, M.

    1982-01-01

    The criteria and safety principles for the design of nuclear power stations are presented from the viewpoint of a nuclear engineer. The design, construction and operation of nuclear power stations should be carried out according to these criteria and safety principles to ensure, to a reasonable degree, that the likelihood of release of radioactivity as a result of component failure or human error should be minimized. (author)

  12. Safety assessment of outdoor live fire range

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-05-01

    The following Safety Assessment (SA) pertains to the outdoor live fire range facility (LFR). The purpose of this facility is to supplement the indoor LFR. In particular it provides capacity for exercises that would be inappropriate on the indoor range. This SA examines the risks that are attendant to the training on the outdoor LFR. The outdoor LFR used by EG&G Mound is privately owned. It is identified as the Miami Valley Shooting Grounds. Mondays are leased for the exclusive use of EG&G Mound.

  13. Safety criteria from the public viewpoint

    International Nuclear Information System (INIS)

    Renn, O.

    1994-01-01

    The paper attempts to outline the scope and limits of a consensus for the evaluation of energy systems, particularly nuclear energy. It is divided into four sections. The first section deals with factual acceptance of technology, while the second inquires into the specific acceptance of nuclear energy, i.e., public perception and valuation of nuclear energy today. The third section discusses criteria of acceptability. In the fourth section, finally, the author deals with questions concerning an energy consensus and presents his own model for approaching this issue. (orig.) [de

  14. Guidance for the definition and application of probabilistic safety criteria

    International Nuclear Information System (INIS)

    Holmberg, J.-E.; Knochenhauer, M.

    2011-05-01

    The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

  15. Guidance for the definition and application of probabilistic safety criteria

    Energy Technology Data Exchange (ETDEWEB)

    Holmberg, J.-E. (VTT Technical Research Centre of Finland (Finland)); Knochenhauer, M. (Scandpower AB (Sweden))

    2011-05-15

    The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

  16. Systems engineered health and safety criteria for safety analysis reports

    International Nuclear Information System (INIS)

    Beitel, G.A.; Morcos, N.

    1993-01-01

    The world of safety analysis is filled with ambiguous words: codes and standards, consequences and risks, hazard and accident, and health and safety. These words have been subject to disparate interpretations by safety analysis report (SAR) writers, readers, and users. open-quotes Principal health and safety criteriaclose quotes has been one of the most frequently misused phrases; rarely is it used consistently or effectively. This paper offers an easily understood definition for open-quotes principal health and safety criteriaclose quotes and uses systems engineering to convert an otherwise mysterious topic into the primary means of producing an integrated SAR. This paper is based on SARs being written for environmental restoration and waste management activities for the U.S. Department of Energy (DOE). Requirements for these SARs are prescribed in DOE Order 5480-23, open-quotes Nuclear Safety Analysis Reports.close quotes

  17. Discussions about safety criteria and guidelines for radioactive waste management.

    Science.gov (United States)

    Yamamoto, Masafumi

    2011-07-01

    In Japan, the clearance levels for uranium-bearing waste have been established by the Nuclear Safety Commission (NSC). The criteria for uranium-bearing waste disposal are also necessary; however, the NSC has not concluded the discussion on this subject. Meanwhile, the General Administrative Group of the Radiation Council has concluded the revision of its former recommendation 'Regulatory exemption dose for radioactive solid waste disposal', the dose criteria after the institutional control period for a repository. The Standardization Committee on Radiation Protection in the Japan Health Physics Society (The Committee) also has developed the relevant safety criteria and guidelines for existing exposure situations, which are potentially applicable to uranium-bearing waste disposal. A new working group established by The Committee was initially aimed at developing criteria and guidelines specifically for uranium-bearing waste disposal; however, the aim has been shifted to broader criteria applicable to any radioactive wastes.

  18. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  19. Fuel safety criteria and review by OECD / CSNI task force

    International Nuclear Information System (INIS)

    Van Doesburg, W.

    1999-01-01

    Full text of publication follows: with the advent of advanced fuel and core designs, and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a general feeling that safety margins have been or are being reduced. Historically, fuel safety margins were defined by adding conservatism to the safety limits, which in turn were also fixed in a conservative manner, here, the expression 'conservatism' expresses the fact that bounding or limiting numbers were chosen for model parameters, plant and fuel design data, and fuel operating history values. Unfortunately, as these conservatisms were not quantified (or quantifiable), the amount of safety available or the reduction thereof is difficult to substantiate. For the regulator, it is important to know the margin available with the utilities' request for approval of new fuel or methods; likewise, for the utility and vendor it is important to know what margins exist and what they are based on, to identify in which direction they can make further progress and optimize fuel and fuel cycle cost. Naturally, each party involved will have to decide on how much margin should be in place, to establish operational criteria and ensure that these can actually be met during operation. To assess the margins issue, safety criteria themselves need to be reviewed first. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel available at that time - mostly at zero exposure. Of course, verification was performed as designs progressed in later years, primarily with the aim to be able to prove that safety criteria were adequate as long as the said conservatisms would be retained, and not with the aim to reestablish limits. The mandate to the OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) is to assess the adequacy of existing fuel safety criteria, in view of the 'new design' elements (new

  20. Criteria for optimizing cortical hierarchies with continuous ranges

    Directory of Open Access Journals (Sweden)

    Antje Krumnack

    2010-03-01

    Full Text Available In a recent paper (Reid et al.; 2009, NeuroImage we introduced a method to calculate optimal hierarchies in the visual network that utilizes continuous, rather than discrete, hierarchical levels, and permits a range of acceptable values rather than attempting to fit fixed hierarchical distances. There, to obtain a hierarchy, the sum of deviations from the constraints that define the hierarchy was minimized using linear optimization. In the short time since publication of that paper we noticed that many colleagues misinterpreted the meaning of the term optimal hierarchy. In particular, a majority of them were under the impression that there was perhaps only one optimal hierarchy, but a substantial difficulty in finding that one. However, there is not only more than one optimal hierarchy but also more than one option for defining optimality. Continuing the line of this work we look at additional options for optimizing the visual hierarchy: minimizing the number of violated constraints and minimizing the maximal size of a constraint violation using linear optimization and mixed integer programming. The implementation of both optimization criteria is explained in detail. In addition, using constraint sets based on the data from Felleman and Van Essen, optimal hierarchies for the visual network are calculated for both optimization methods.

  1. Safety criteria related to microheterogeneities in LWR mixed oxide fuels

    International Nuclear Information System (INIS)

    Renard, A.; Mostin, N.

    1978-01-01

    The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)

  2. Fuel safety criteria technical review - Results of OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria

    International Nuclear Information System (INIS)

    Hollasky, N.; Valtonen, K.; Hache, G.; Gross, H.; Bakker, K.; Recio, M.; Bart, G.; Zimmermann, M.; Van Doesburg, W.; Killeen, J.; Meyer, R.O.; Speis, T.

    2000-01-01

    With the advent of advanced fuel and core designs, the adoption of more aggressive operational modes and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a concern if safety margins have remained adequate. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel that was available at that time, mostly with unirradiated specimens. Verification was of course performed as designs progressed in later years, however mostly with the aim to be able to prove that these designs adequately complied with existing criteria, and not to establish new limits. The OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) was therefore given the mandate to technically review the existing fuel safety criteria, focusing on the 'new design' elements (new fuel and core design, cladding materials, manufacturing processes, high burnup, MOX, etc.) introduced by the industry. It should also identify if additional efforts may be required (experimental, analytical) to ensure that the basis for fuel safety criteria is adequate to address the relevant safety issues. In this report, fuel-related criteria are discussed without attempting to categorize them according to event type or risk significance. For each of these 20 criteria, we present a brief description of the criterion as it is used in several applications along with the rationale for having such a criterion. New design elements, such as different cladding materials, higher burnup, and the use of MOX fuels, can affect fuel-related margins and, in some cases, the criteria themselves. Some of the more important effects are mentioned in order to indicate whether the criteria need to be re-evaluated. The discussion may not cover all possible effects, but should be sufficient to identify those criteria that need to be addressed. A summary of these discussions is given in Section 7. As part

  3. Licensing procedures and safety criteria for research reactors in France

    International Nuclear Information System (INIS)

    Berry, J.L.; Lerouge, B.

    1983-01-01

    From the very beginning of the CEA up to now, a great deal of work has been devoted to the development and utilization of research reactors in France for the needs of fundamental and applied research, production of radioisotopes, and training. In recent years, new reactors were commissioned while others were decommissioned. Moreover some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors (Osiris and Isis). This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. At the end, a few considerations are given to the consequences of the Osiris core conversion. Safety of research reactors has been studied in detail and many improvements have been brought due to: implementation of a specific experimental program, and adaptation of safety principles and rules elaborated for power reactors. Research reactors in operation in France have been built within a 22 year period. Meanwhile, safety rules have been improved. Old reactors do not comply with all the new rules but modifications are continuously made: after analysis of incidents, when replacement of equipment has to be carried out, when an important modification (fuel conversion for example) is decided upon

  4. Licensing procedures and safety criteria for research reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Berry, J L; Lerouge, B [Centre d' Etudes Nucleaires de Saclay (France)

    1983-08-01

    From the very beginning of the CEA up to now, a great deal of work has been devoted to the development and utilization of research reactors in France for the needs of fundamental and applied research, production of radioisotopes, and training. In recent years, new reactors were commissioned while others were decommissioned. Moreover some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors (Osiris and Isis). This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. At the end, a few considerations are given to the consequences of the Osiris core conversion. Safety of research reactors has been studied in detail and many improvements have been brought due to: implementation of a specific experimental program, and adaptation of safety principles and rules elaborated for power reactors. Research reactors in operation in France have been built within a 22 year period. Meanwhile, safety rules have been improved. Old reactors do not comply with all the new rules but modifications are continuously made: after analysis of incidents, when replacement of equipment has to be carried out, when an important modification (fuel conversion for example) is decided upon.

  5. Criteria for guidance in the safety assessment of nuclear installations in the United Kingdom

    International Nuclear Information System (INIS)

    Gausden, R.; Fryer, D.R.H.

    1977-01-01

    There is an increasing appreciation of the need for a consistent approach to nuclear safety between various groups having an interest in safety and between various types of installation. Licensing for construction and ultimate approval to operate any nuclear installation depend in the United Kingdom upon a searching assessment of the design, construction and operation of the proposed plant. Criteria of the kind discussed in this paper have been used by the Nuclear Installations Inspectorate in this assessment process. From time to time they are subject to comments from other bodies in the U.K. One aim of the criteria is to set out the broad objectives that should be met regarding the magnitude of radiological consequences of accidents or normal operation. In addition, the criteria give guidance on the design philosophy for nuclear safety and the principles of fault evaluation. Criteria must be conceived so that while maintaining safety standards their application does not frustrate design and development. It is also important that undue formalism is not induced in the assessment process at the expense of inhibiting the judgement of safety assessors. A balance must, therefore, be struck between detailed and generalised guidance. It is also accepted that experience in the use and interpretation of criteria will indicate a need for improvement and additions: the criteria are, therefore, regarded as living rather than fixed statements which are expected to develop in response to any need for change in a safe direction that may arise. In developing them, the Inspectorate has drawn heavily upon the experience accumulated during its 16 years of operation and has also referred to criteria published by other organisations. The paper deals specifically with certain of the most important sections of the criteria and indicates the total range of subjects which need to be included in such criteria

  6. Priming patient safety: A middle-range theory of safety goal priming via safety culture communication.

    Science.gov (United States)

    Groves, Patricia S; Bunch, Jacinda L

    2018-05-18

    The aim of this paper is discussion of a new middle-range theory of patient safety goal priming via safety culture communication. Bedside nurses are key to safe care, but there is little theory about how organizations can influence nursing behavior through safety culture to improve patient safety outcomes. We theorize patient safety goal priming via safety culture communication may support organizations in this endeavor. According to this theory, hospital safety culture communication activates a previously held patient safety goal and increases the perceived value of actions nurses can take to achieve that goal. Nurses subsequently prioritize and are motivated to perform tasks and risk assessment related to achieving patient safety. These efforts continue until nurses mitigate or ameliorate identified risks and hazards during the patient care encounter. Critically, this process requires nurses to have a previously held safety goal associated with a repertoire of appropriate actions. This theory suggests undergraduate educators should foster an outcomes focus emphasizing the connections between nursing interventions and safety outcomes, hospitals should strategically structure patient safety primes into communicative activities, and organizations should support professional development including new skills and the latest evidence supporting nursing practice for patient safety. © 2018 John Wiley & Sons Ltd.

  7. Methods of checking general safety criteria in UML statechart specifications

    International Nuclear Information System (INIS)

    Pap, Zsigmond; Majzik, Istvan; Pataricza, Andras; Szegi, Andras

    2005-01-01

    This paper describes methods and tools for safety analysis of UML statechart specifications. A comprehensive set of general safety criteria including completeness and consistency is applied in automated analysis. Analysis techniques are based on OCL expressions, graph transformations and reachability analysis. Two canonical intermediate representations of the statechart specification are introduced. They are suitable for straightforward implementation of checker methods and for the support of the proof of the correctness and soundness of the applied analysis. One of them also serves as a basis of the metamodel of a variant of UML statecharts proposed for the specification of safety-critical control systems. The analysis is extended to object-oriented specifications. Examples illustrate the application of the checker methods implemented by an automated tool-set

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

    International Nuclear Information System (INIS)

    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

  9. Common Risk Criteria Standards for National Test Ranges

    Science.gov (United States)

    2017-09-01

    capability greater than 150 kilometers (km), ranges should coordinate with the Joint Space Operations Squadron (JSpOC) for conjunction assessment if...insurance to cover such potential mishaps and has historically not required conjunction assessments for mission assurance or unmanned asset protection...into a sustainable orbit, the duration of the conjunction assessment required for manned and active spacecraft protection shall be applied from

  10. Methods and criteria for safety analysis (FIN L2535)

    International Nuclear Information System (INIS)

    1992-12-01

    In response to the NRC request for a proposal dated October 20, 1992, Westinghouse Savannah River Company (WSRC) submit this proposal to provide contractural assistance for FIN L2535, ''Methods and Criteria for Safety Analysis,'' as specified in the Statement of Work attached to the request for proposal. The Statement of Work involves development of safety analysis guidance for NRC licensees, arranging a workshop on this guidance, and revising NRC Regulatory Guide 3.52. This response to the request for proposal offers for consideration the following advantages of WSRC in performing this work: Experience, Qualification of Personnel and Resource Commitment, Technical and Organizational Approach, Mobilization Plan, Key Personnel and Resumes. In addition, attached are the following items required by the NRC: Schedule II, Savannah River Site - Job Cost Estimate, NRC Form 189, Project and Budget Proposal for NRC Work, page 1, NRC Form 189, Project and Budget Proposal for NRC Work, page 2, Project Description

  11. Safety analysis, risk assessment, and risk acceptance criteria

    International Nuclear Information System (INIS)

    Jamali, K.

    1997-01-01

    This paper discusses a number of topics that relate safety analysis as documented in the Department of Energy (DOE) safety analysis reports (SARs), probabilistic risk assessments (PRA) as characterized primarily in the context of the techniques that have assumed some level of formality in commercial nuclear power plant applications, and risk acceptance criteria as an outgrowth of PRA applications. DOE SARs of interest are those that are prepared for DOE facilities under DOE Order 5480.23 and the implementing guidance in DOE STD-3009-94. It must be noted that the primary area of application for DOE STD-3009 is existing DOE facilities and that certain modifications of the STD-3009 approach are necessary in SARs for new facilities. Moreover, it is the hazard analysis (HA) and accident analysis (AA) portions of these SARs that are relevant to the present discussions. Although PRAs can be qualitative in nature, PRA as used in this paper refers more generally to all quantitative risk assessments and their underlying methods. HA as used in this paper refers more generally to all qualitative risk assessments and their underlying methods that have been in use in hazardous facilities other than nuclear power plants. This discussion includes both quantitative and qualitative risk assessment methods. PRA has been used, improved, developed, and refined since the Reactor Safety Study (WASH-1400) was published in 1975 by the Nuclear Regulatory Commission (NRC). Much debate has ensued since WASH-1400 on exactly what the role of PRA should be in plant design, reactor licensing, 'ensuring' plant and process safety, and a large number of other decisions that must be made for potentially hazardous activities. Of particular interest in this area is whether the risks quantified using PRA should be compared with numerical risk acceptance criteria (RACs) to determine whether a facility is 'safe.' Use of RACs requires quantitative estimates of consequence frequency and magnitude

  12. Selection of tolerable risk criteria for dam safety decision making

    International Nuclear Information System (INIS)

    Nielsen, N.M.; Hartford, D.N.D.; MacDonald, T.F.

    1994-01-01

    Risk assessment has received increasing attention in recent years as a means of aiding decision making on dams by providing systematic and rational methods for dealing with risk and uncertainty. Risk assessment is controversial and decisions affecting risk to life are the most controversial. Tolerable criteria, based on the risks that society is prepared to accept in order to avoid excessive costs, set bounds within which risk-based decisions may be made. The components of risk associated with dam safety are addressed on an individual basis and criteria established for each component, thereby permitting flexibility in the balance between component risk and avoiding the problems of placing a monetary value on life. The guiding principle of individual risk is that dams do not impose intolerable risks on any individual. A risk to life of 1 in 10 4 per annum is generally considered the maximum tolerable risk. When considering societal risk, the safety of a dam should be proportional to the consequences of its failure. Risks of financial losses beyond the corporation's ability to finance should be so low as to be considered negligible. 17 refs., 3 figs

  13. Occupational safety and health criteria for responsible development of nanotechnology

    Science.gov (United States)

    Schulte, P. A.; Geraci, C. L.; Murashov, V.; Kuempel, E. D.; Zumwalde, R. D.; Castranova, V.; Hoover, M. D.; Hodson, L.; Martinez, K. F.

    2014-01-01

    Organizations around the world have called for the responsible development of nanotechnology. The goals of this approach are to emphasize the importance of considering and controlling the potential adverse impacts of nanotechnology in order to develop its capabilities and benefits. A primary area of concern is the potential adverse impact on workers, since they are the first people in society who are exposed to the potential hazards of nanotechnology. Occupational safety and health criteria for defining what constitutes responsible development of nanotechnology are needed. This article presents five criterion actions that should be practiced by decision-makers at the business and societal levels—if nanotechnology is to be developed responsibly. These include (1) anticipate, identify, and track potentially hazardous nanomaterials in the workplace; (2) assess workers' exposures to nanomaterials; (3) assess and communicate hazards and risks to workers; (4) manage occupational safety and health risks; and (5) foster the safe development of nanotechnology and realization of its societal and commercial benefits. All these criteria are necessary for responsible development to occur. Since it is early in the commercialization of nanotechnology, there are still many unknowns and concerns about nanomaterials. Therefore, it is prudent to treat them as potentially hazardous until sufficient toxicology, and exposure data are gathered for nanomaterial-specific hazard and risk assessments. In this emergent period, it is necessary to be clear about the extent of uncertainty and the need for prudent actions.

  14. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  15. Transportation of Organs by Air: Safety, Quality, and Sustainability Criteria.

    Science.gov (United States)

    Mantecchini, L; Paganelli, F; Morabito, V; Ricci, A; Peritore, D; Trapani, S; Montemurro, A; Rizzo, A; Del Sordo, E; Gaeta, A; Rizzato, L; Nanni Costa, A

    2016-03-01

    The outcomes of organ transplantation activities are greatly affected by the ability to haul organs and medical teams quickly and safely. Organ allocation and usage criteria have greatly improved over time, whereas the same result has not been achieved so far from the transport point of view. Safety and the highest level of service and efficiency must be reached to grant transplant recipients the healthiest outcome. The Italian National Transplant Centre (CNT), in partnership with the regions and the University of Bologna, has promoted a thorough analysis of all stages of organ transportation logistics chains to produce homogeneous and shared guidelines throughout the national territory, capable of ensuring safety, reliability, and sustainability at the highest levels. The mapping of all 44 transplant centers and the pertaining airport network has been implemented. An analysis of technical requirements among organ shipping agents at both national and international level has been promoted. A national campaign of real-time monitoring of organ transport activities at all stages of the supply chain has been implemented. Parameters investigated have been hospital and region of both origin and destination, number and type of organs involved, transport type (with or without medical team), stations of arrival and departure, and shipping agents, as well as actual times of activities involved. National guidelines have been issued to select organ storage units and shipping agents on the basis of evaluation of efficiency, reliability, and equipment with reference to organ type and ischemia time. Guidelines provide EU-level standards on technical equipment of aircrafts, professional requirements of shipping agencies and cabin crew, and requirements on service provision, including pricing criteria. The introduction in the Italian legislation of guidelines issuing minimum requirements on topics such as the medical team, packaging, labeling, safety and integrity, identification

  16. Probabilistic safety criteria on high burnup HWR fuels

    International Nuclear Information System (INIS)

    Marino, A.C.

    2002-01-01

    BACO is a code for the simulation of the thermo-mechanical and fission gas behaviour of a cylindrical fuel rod under operation conditions. Their input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA-CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. The techniques for sensitivity analysis defined in BACO are: the 'extreme case analysis', the 'parametric analysis' and the 'probabilistic (or statistics) analysis'. We analyse the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs taking account probabilistic safety criteria and using the BACO's sensitivity analysis. (author)

  17. Safety criteria for the next generation of European reactors

    International Nuclear Information System (INIS)

    Dominguez Bautista, M.T.

    1995-01-01

    For the next generation of reactors, European companies operating in the electricity sector have drawn up a document called European Utilities Requirement (EUR), which sets out the requirements to be met by the designers of future reactors. The main objective of these new requirements is to increase the safety in existing reactors, making good use of operating experience available and the technological developments of the last decade. This paper offers an in-depth analysis of the most significant characteristics, describing how the EUR requirements have been prepared and how they are being implemented by the designers. Areas covered are: - Combining deterministic and probabilistic criteria - Automation of control systems - Design extension for severe accidents - Containment design - Emergency plans - Autonomy versus manual operation

  18. Argentine criteria on nuclear safety and emergencies: their impact on the Argos PHWR 380 design

    International Nuclear Information System (INIS)

    Gonzalez, A. J.

    1988-01-01

    This paper describes first the safety criteria of the Argentine regulatory authority with emphasis on the probabilistic safety criteria based on a limitation of individual risks. Then, it is presented a discussion on emergency criteria in relation to evacuation and relocation measures. Finally, the paper briefly describes the design of an Argentine offer for a safer heavy water reactor where these criteria are applied. 9 figs., 1 tab., 46 refs. (author)

  19. Safety criteria for spent-fuel transport. Final report

    International Nuclear Information System (INIS)

    Goldmann, K.; Gekler, W.C.

    1986-10-01

    The focus of this study is on the question, ''Do current regulations provide reasonable assurance of safety for a transport scenario of spent fuel, as presently anticipated by the Department of Energy, under the Nuclear Waste Policy Act.'' This question has been addressed by developing a methodology for identifying the expected frequency of Accidents Which Exceed Regulatory Conditions in Severity (AWERCS) for spent fuel transport casks and then assessing the health effects resulting from that frequency. By applying the methodology to an illustrative case of road transports, it was found that the accidental release of radioactive material from impact AWERCS would make negligible contributions to health effects associated with spent fuel transports by road. It is also concluded that the current regulatory drop test requirements in 10 CFR 71.51 which form the basis for cask design and were used to establish AWERCS screening criteria for this study are adequate, and that no basis was found to conclude that cask performance under expected road accident conditions represents an undue risk to the public

  20. Safety approach to the selection of design criteria for the CRBRP reactor refueling system

    International Nuclear Information System (INIS)

    Meisl, C.J.; Berg, G.E.; Sharkey, N.F.

    1979-01-01

    The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents. The process steps are illustrated by examples

  1. International comparison of safety criteria applied to radwaste repositories. Safety aspects of the post-operational phase

    International Nuclear Information System (INIS)

    Baltes, B.

    1994-01-01

    There is a generally accepted system of framework safety conditions governing the construction, operation, and post-operational monitoring of radwaste repositories. Although the development of these framework conditions may vary from country to country, the resulting criteria are based on the commonly accepted system of priciples and purposes established for ultimate radioactive waste disposal. The experience accumulated by GRS in the course of the plan approval procedure for the Konrad mine site and the safety-relevant studies performed for the planned Morsleben repository clearly show demand for further development of the safety criteria. In Germany, it is especially the safety criteria and detailed requirements filling the framework safety conditions that need revision and in-depth definition, as well as comparison and harmonisation with internationally applied criteria. These activities will particularly consider the international convention on radioactive waste management currently in preparation under the auspieces of the IAEA. (orig.) [de

  2. Safety assessment of indoor live fire range, May 1989

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-05-01

    The following Safety Assessment (SA) pertains to the indoor live fire range (LFR) at EG&G Mound Applied Technology plant. The purpose of the indoor LFR is to conduct training with live ammunition for all designated personnel. The SA examines the risks that are attendant to the operation of an indoor LFR for this purpose.

  3. Human factors engineering design review acceptance criteria for the safety parameter display

    International Nuclear Information System (INIS)

    McGevna, V.; Peterson, L.R.

    1981-01-01

    This report contains human factors engineering design review acceptance criteria developed by the Human Factors Engineering Branch (HFEB) of the Nuclear Regulatory Commission (NRC) to use in evaluating designs of the Safety Parameter Display System (SPDS). These criteria were developed in response to the functional design criteria for the SPDS defined in NUREG-0696, Functional Criteria for Emergency Response Facilities. The purpose of this report is to identify design review acceptance criteria for the SPDS installed in the control room of a nuclear power plant. Use of computer driven cathode ray tube (CRT) displays is anticipated. General acceptance criteria for displays of plant safety status information by the SPDS are developed. In addition, specific SPDS review criteria corresponding to the SPDS functional criteria specified in NUREG-0696 are established

  4. Safety objectives and design criteria for the NHR-200

    International Nuclear Information System (INIS)

    Xue Dazhi; Zheng Wenxiang

    1997-01-01

    The construction of a nuclear district heating reactor (NHR) demonstration plant with a thermal power of 200 MW has been decided for the northeast of China. To facilitate the design and licensability a set of design criteria were developed for the NHR, based on existing general criteria for NPP but amended with regard to the unique features of NHR-200. Some key points are discussed in this paper. (author). 7 refs

  5. Safety objectives and design criteria for the NHR-200

    Energy Technology Data Exchange (ETDEWEB)

    Dazhi, Xue; Wenxiang, Zheng [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The construction of a nuclear district heating reactor (NHR) demonstration plant with a thermal power of 200 MW has been decided for the northeast of China. To facilitate the design and licensability a set of design criteria were developed for the NHR, based on existing general criteria for NPP but amended with regard to the unique features of NHR-200. Some key points are discussed in this paper. (author). 7 refs.

  6. Safety criteria for design of nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

    In Finland the general safety requirements for nuclear power plants are presented in the Council of State Decision (395/91). In this guide, safety principles which supplement the Council of State Decision and which are to be used in the design of nuclear power plants are defined

  7. Improving safety margin of LWRs by rethinking the emergency core cooling system criteria and safety system capacity

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Bokyung, E-mail: bkkim2@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2016-10-15

    Highlights: • Zircaloy embrittlement criteria can increase to 1370 °C for CP-ECR lower than 13%. • The draft ECCS criteria of U.S. NRC allow less than 5% in power margin. • The Japanese fracture-based criteria allow around 5% in power margin. • Increasing SIT inventory is effective in assuring safety margin for power uprates. - Abstract: This study investigates the engineering compatibility between emergency core cooling system criteria and safety water injection systems, in the pursuit of safety margin increase of light water reactors. This study proposes an acceptable temperature increase to 1370 °C as long as equivalent cladding reacted calculated by the Cathcart–Pawel equation is below 13%, after an extensive literature review. The influence of different ECCS criteria on the safety margin during large break loss of coolant accident is investigated for OPR-1000 by the system code MARS-KS, implemented with the KINS-REM method. The fracture-based emergency core cooling system (ECCS) criteria proposed in this study are shown to enable power margins up to 10%. In the meantime, the draft U.S. NRC’s embrittlement criteria (burnup-sensitive) and Japanese fracture-based criteria are shown to allow less than 5%, and around 5% of power margins, respectively. Increasing safety injection tank (SIT) water inventory is the key, yet convenient, way of assuring safety margin for power increase. More than 20% increase in the SIT water inventory is required to allow 15% power margins, for the U.S. NRC’s burnup-dependent embrittlement criteria. Controlling SIT water inventory would be a useful option that could allow the industrial desire to pursue power margins even under the recent atmosphere of imposing stricter ECCS criteria for the considerable burnup effects.

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

    International Nuclear Information System (INIS)

    Reitsma, Frederik; Silady, Fred; Kunitomi, Kazuhiko

    2014-01-01

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

  9. Fuel safety criteria in NEA member countries - Compilation of responses received from member countries

    International Nuclear Information System (INIS)

    2003-03-01

    In 2001 the Committee on the Safety of Nuclear Installations (CSNI) issued a report on Fuel Safety Criteria Technical Review. The objective was to review the present fuel safety criteria and judge to which extent they are affected by the 'new' design elements, such as different cladding materials, higher burnup, the use of MOX fuels, etc. The report stated that the current framework of fuel safety criteria remains generally applicable, being largely unaffected by the 'new' or modern design elements. The levels (numbers) in the individual safety criteria may, however, change in accordance with the particular fuel and core design features. Some of these levels have already been - or are continuously being - adjusted. The level adjustments of several other criteria (RIA, LOCA) also appears to be needed, on the basis of experimental data and the analysis thereof. As a follow-up, among its first tasks, the CSNI Special Expert Group on Fuel Safety Margins (SEG FSM) initiated the collection of information on the present fuel safety criteria used in NEA member states with the objective to solicit national practices in the use of fuel safety criteria, in particular to get information on their specific national levels/values, including their recent adjustments, and to identify the differences and commonalties between the different countries. Two sources of information were used to produce this report: a compilation of responses to a questionnaire prepared for the June 2000 CNRA meeting, and individual responses from the SEGFSM members to the new revised questionnaire issued by the task Force preparing this report. In accordance with the latter, the fuel safety criteria discussed in this report were divided into three categories: (A) safety criteria - criteria imposed by the regulator; (B) operational criteria - specific to the fuel design and provided by the fuel vendor as part of the licensing basis; (C) design criteria - limits employed by vendors and/or utilities for fuel

  10. Safety criteria for siting a nuclear power plant

    International Nuclear Information System (INIS)

    2001-01-01

    The guide sets forth requirements for safety of the population and the environment in nuclear power plant siting. It also sets out the general basis for procedures employed by other competent authorities when they issue regulations or grant licences. On request STUK (Radiation and Nuclear Safety Authority of Finland) issues case-specific statements about matters relating to planning and about other matters relating to land use in the environment of nuclear power plants

  11. Safety Design Criteria and Approaches to Safety Substantiation of the BN-1200

    International Nuclear Information System (INIS)

    Ashurko, I.

    2013-01-01

    Russian experience in SFR area: Activities on development of safety design criteria for SFRs of the 4th generation is carried out within the GIF framework. Although this reactor technology is considered as innovative that is relevant to the 4th generation, however, it has already a certain history. In this relation, it seems to be useful to analyze the corresponding experience that is available in various countries. 4 SFRs have been successfully operated in the USSR and in the Russian Federation: • Experimental reactor BR-5/10; • Research reactor BOR-60; • Prototype BN-350 power reactor; • Commercial BN-600 power unit at the Beloyarsk NPP. Thus, Russia gained a considerable experience of design, construction and operation of SFRs. In particular, a certain experience has been acquired on safety substantiation of reactors of this type and their licensing. Now BOR-60 and BN-600 continue their operation, BN-800 power unit is under construction, development of the commercial BN-1200 power unit, that is considered as the 4th generation reactor, has been started. Due to limited number of operating SFRs in the world, successful Russian experience in this area should be taken into account for further development and improvement of SFR SDC developed by the GIF Task Force. In particular, participation of SFR designers in this activities would be fruitful and useful

  12. Analysis of existing work-zone devices with MASH safety performance criteria.

    Science.gov (United States)

    2009-02-01

    Crashworthy, work-zone, portable sign support systems accepted under NCHRP Report No. 350 were analyzed to : predict their safety peformance according to the TL-3 MASH evaluation criteria. An analysis was conducted to determine : which hardware param...

  13. A utility theoretic view on probabilistic safety criteria

    International Nuclear Information System (INIS)

    Holmberg, J.E.

    1997-03-01

    A probabilistic safety criterion specifies the maximum acceptable hazard rates of various accidental consequences. Assuming that the criterion depends also on the benefit of the process to society and on the licensing time applied, we can regard such statements as preference relations. In this paper, a probabilistic safety criterion is interpreted to mean that if the accident hazard rate is higher than the accident hazard rate criterion, then the optimal stopping time of a hazardous process is shorter than the licensing time. This interpretation yields a condition for a feasible utility function. In particular, we derive such a condition for the parameters of a linear plus exponential utility function. (orig.) (12 refs.)

  14. Criteria Document for B-plant's Surveillance and Maintenance Phase Safety Basis Document

    International Nuclear Information System (INIS)

    SCHWEHR, B.A.

    1999-01-01

    This document is required by the Project Hanford Managing Contractor (PHMC) procedure, HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval. This document specifies the criteria that shall be in the B Plant surveillance and maintenance phase safety basis in order to obtain approval of the DOE-RL. This CD describes the criteria to be addressed in the S and M Phase safety basis for the deactivated Waste Fractionization Facility (B Plant) on the Hanford Site in Washington state. This criteria document describes: the document type and format that will be used for the S and M Phase safety basis, the requirements documents that will be invoked for the document development, the deactivated condition of the B Plant facility, and the scope of issues to be addressed in the S and M Phase safety basis document

  15. A critical overview of safety-related and technological criteria for nuclear fuel

    International Nuclear Information System (INIS)

    Lahodova, M.; Valach, M.

    2000-10-01

    A detailed overview of the safety criteria, methods of analysis and computer codes used in OECD countries is presented. A critical analysis of the validity of criteria in the high burnup domain was performed, and recommendations for testing their validity based on available experimental data are put forth. (author)

  16. Licensing procedures and safety criteria for core conversion in Japan

    International Nuclear Information System (INIS)

    Kanda, K.; Nakagome, Y.; Hayashi, M.

    1983-01-01

    Procedures relating to the construction and operation of reactor facilities are discussed. Specifically, the Safety Analysis Report on the Kyoto University Critical Assembly (KUCA) core conversion (93% to 45% enrichment) is noted. The results of critical experiments in the KUCA and of burnup tests in the Oak Ridge Research (ORR) Reactor will be used in the final determination of the feasibility of the conversion of the Kyoto University High Flux Reactor (KUHFR) to the use of 45% enrichment

  17. Advanced Range Safety System for High Energy Vehicles

    Science.gov (United States)

    Claxton, Jeffrey S.; Linton, Donald F.

    2002-01-01

    The advanced range safety system project is a collaboration between the National Aeronautics and Space Administration and the United States Air Force to develop systems that would reduce costs and schedule for safety approval for new classes of unmanned high-energy vehicles. The mission-planning feature for this system would yield flight profiles that satisfy the mission requirements for the user while providing an increased quality of risk assessment, enhancing public safety. By improving the speed and accuracy of predicting risks to the public, mission planners would be able to expand flight envelopes significantly. Once in place, this system is expected to offer the flexibility of handling real-time risk management for the high-energy capabilities of hypersonic vehicles including autonomous return-from-orbit vehicles and extended flight profiles over land. Users of this system would include mission planners of Space Launch Initiative vehicles, space planes, and other high-energy vehicles. The real-time features of the system could make extended flight of a malfunctioning vehicle possible, in lieu of an immediate terminate decision. With this improved capability, the user would have more time for anomaly resolution and potential recovery of a malfunctioning vehicle.

  18. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core

  19. Safety Design Criteria (SDC) for Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Nakai, Ryodai

    2013-01-01

    SDC Development Background & Objectives: • Safety Design Criteria (SDC) Development for Gen-IV SFR: – Proposed at the GIF Policy Group (PG) meeting in October 2010 –SDC “harmonization” is increasingly important for: • Realization of enhanced safety designs meeting to Gen-IV safety goals and safety approach common to SFR systems; • Preparation for the forthcoming licensing in the near future; • Because Gen-IV SFR are progressing into conceptual design stage. • The SDC is the Reference criteria: – Of the designs of safety-related Structures, Systems & Components that are specific to the SFR system; – For clarifying the requisites systematically & comprehensively; – When the technology developers apply the basic safety approach and use the codes & standards for conceptual design of the Gen-IV SFR system

  20. Licensing procedures and safety criteria for core conversion in Japan

    International Nuclear Information System (INIS)

    Kanda, K.; Nakagome, Y.; Hayashi, M.

    1983-01-01

    In Japan, the establishment and operation of nuclear installations are governed mainly by the Law for Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors. This law lays down the regulations and conditions for licensing of the various installations involved in the nuclear fuel cycle, namely licensing of installations for refining, fabricating and reprocessing; and reactors, as well as licensing of the use of nuclear fuels in research facilities. Although procedures for the installations listed above vary depending on the installation concerned, only those relating to construction and operation of reactor facilities will be analysed in this study, as the conditions and principles applying to licensing and control of other installations are, to a large extent, similar to those concerning reactor facilities. The second part of this presentation describes the safety review of the KUCA reactor core conversion form HEU to MEU. For the safety review of the core conversion, the Committee on Examination of Reactor Safety of Japanese Government examined mainly the the nuclear characteristics and the integrity of aluminide fuel plates, which was very severe because we had no experience to use aluminide fuel plates in Japan. The integrity of fuel plates and the results of the worst accident analysis for the MEU core are shown with the comparison between the HEU and MEU cores. The significant difference was not observed between them. All the regulatory procedures were completed in September 1980. Fabrication of MEU fuel elements for the KUCA experiments by CERCA in France was started in September 1980, and will be completed in March 1981. The critical experiments in the KUCA with MEU fuel will be started on a single-core in May 1981 as a first step. Those on a coupled-core will follow

  1. Exemption, exception and other criteria for transport criticality safety

    International Nuclear Information System (INIS)

    Mennerdahl, D.

    2004-01-01

    Many strange concepts, requirements and specifications related to criticality safety are present in the Regulations. Some earlier problems have been corrected but, going back to 1961 and the first edition of the Regulations, it seems as many changes have been to the worse. Fissile material was defined correctly as a material that could consist of or contain fissile nuclides. Materials consisting of pure fissile nuclides don't exist but are important in package designs. 238 Pu was included as a fissile nuclide only as an emergency, because there was no alternative, but this caused some people to think that all nuclides supporting criticality are fissile. Neutron interaction between different (non-identical) packages had to be evaluated, making the transport index or allowable number of packages a credible safety control. That is not true anymore. The 15 gram exception limit for fissile nuclides was combined with a transport mode limit, similar to but more restrictive than the current consignment limit. The confinement system was introduced to help with formulation of a single requirement for safety of the containment system but is becoming something very different. Controls before the first use of a packaging have become controls of the first use of a package, supporting multiple shipments of the same package. The lack of exemption limits for fissile material essentially makes all radioactive materials fissile (all radioactive material contains some fissile atoms). Radioactive material seems to be defined without consideration of the criticality hazard of the material. LSA materials are defined with consideration of criticality, but only relates to quantities in fissile exceptions when other properties can be equally or more important. In July 2004, a number of proposals to IAEA have been submitted by Sweden to improve and expand the criticality safety control of the Regulations. Essential is the introduction of the fissionable nuclide and material concepts in

  2. Exemption, exception and other criteria for transport criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Mennerdahl, D. [E Mennerdahl Systems, Taeby (Sweden)

    2004-07-01

    Many strange concepts, requirements and specifications related to criticality safety are present in the Regulations. Some earlier problems have been corrected but, going back to 1961 and the first edition of the Regulations, it seems as many changes have been to the worse. Fissile material was defined correctly as a material that could consist of or contain fissile nuclides. Materials consisting of pure fissile nuclides don't exist but are important in package designs. {sup 238}Pu was included as a fissile nuclide only as an emergency, because there was no alternative, but this caused some people to think that all nuclides supporting criticality are fissile. Neutron interaction between different (non-identical) packages had to be evaluated, making the transport index or allowable number of packages a credible safety control. That is not true anymore. The 15 gram exception limit for fissile nuclides was combined with a transport mode limit, similar to but more restrictive than the current consignment limit. The confinement system was introduced to help with formulation of a single requirement for safety of the containment system but is becoming something very different. Controls before the first use of a packaging have become controls of the first use of a package, supporting multiple shipments of the same package. The lack of exemption limits for fissile material essentially makes all radioactive materials fissile (all radioactive material contains some fissile atoms). Radioactive material seems to be defined without consideration of the criticality hazard of the material. LSA materials are defined with consideration of criticality, but only relates to quantities in fissile exceptions when other properties can be equally or more important. In July 2004, a number of proposals to IAEA have been submitted by Sweden to improve and expand the criticality safety control of the Regulations. Essential is the introduction of the fissionable nuclide and material

  3. Real-life effectiveness of omalizumab in severe allergic asthma above the recommended dosing range criteria.

    Science.gov (United States)

    Hew, M; Gillman, A; Sutherland, M; Wark, P; Bowden, J; Guo, M; Reddel, H K; Jenkins, C; Marks, G B; Thien, F; Rimmer, J; Katsoulotos, G P; Cook, M; Yang, I; Katelaris, C; Bowler, S; Langton, D; Wright, C; Bint, M; Yozghatlian, V; Burgess, S; Sivakumaran, P; Yan, K Y; Kritikos, V; Peters, M; Baraket, M; Aminazad, A; Robinson, P; Jaffe, A; Powell, H; Upham, J W; McDonald, V M; Gibson, P G

    2016-11-01

    Omalizumab (Xolair) dosing in severe allergic asthma is based on serum IgE and bodyweight. In Australia, patients eligible for omalizumab but exceeding recommended ranges for IgE (30-1500 IU/mL) and bodyweight (30-150 kg) may still receive a ceiling dose of 750 mg/4 weeks. About 62% of patients receiving government-subsidized omalizumab are enrolled in the Australian Xolair Registry (AXR). To determine whether AXR participants above the recommended dosing ranges benefit from omalizumab and to compare their response to within-range participants. Data were stratified according to dose range status (above-range or within-range). Further sub-analyses were conducted according to the reason for being above the dosing range (IgE only vs. IgE and weight). Data for 179 participants were analysed. About 55 (31%) were above recommended dosing criteria; other characteristics were similar to within-range participants. Above-range participants had higher baseline IgE [812 (IQR 632, 1747) IU/mL vs. 209 (IQR 134, 306) IU/mL] and received higher doses of omalizumab [750 (IQR 650, 750) mg] compared to within-range participants [450 (IQR, 300, 600) mg]. At 6 months, improvements in Juniper 5-item Asthma Control Questionnaire (ACQ-5, 3.61 down to 2.01 for above-range, 3.47 down to 1.93 for within-range, P omalizumab have significantly improved symptom control, quality of life and lung function to a similar degree to within-range participants, achieved without dose escalation above 750 mg. © 2016 John Wiley & Sons Ltd.

  4. Compilation of nuclear safety criteria potential application to DOE nonreactor facilities

    International Nuclear Information System (INIS)

    1992-03-01

    This bibliographic document compiles nuclear safety criteria applied to the various areas of nuclear safety addressed in a Safety Analysis Report for a nonreactor nuclear facility (NNF). The criteria listed are derived from federal regulations, Nuclear Regulatory Commission (NRC) guides and publications, DOE and DOE contractor publications, and industry codes and standards. The titles of the chapters and sections of Regulatory Guide 3.26, ''Standard Format and Content of Safety Analysis Reports for Fuel Reprocessing Plants'' were used to format the chapters and sections of this compilation. In each section the criteria are compiled in four groups, namely: (1) Code of Federal Regulations, (2) USNRC Regulatory Guides, (3) Codes and Standards, and (4) Supplementary Information

  5. Evaluation of criteria for developing traffic safety materials for Latinos.

    Science.gov (United States)

    Streit-Kaplan, Erica L; Miara, Christine; Formica, Scott W; Gallagher, Susan Scavo

    2011-03-01

    This quantitative study assessed the validity of guidelines that identified four key characteristics of culturally appropriate Spanish-language traffic safety materials: language, translation, formative evaluation, and credible source material. From a sample of 190, the authors randomly selected 12 Spanish-language educational materials for analysis by 15 experts. Hypotheses included that the experts would rate materials with more of the key characteristics as more effective (likely to affect behavioral change) and rate materials originally developed in Spanish and those that utilized formative evaluation (e.g., pilot tests, focus groups) as more culturally appropriate. Although results revealed a weak association between the number of key characteristics in a material and the rating of its effectiveness, reviewers rated materials originally created in Spanish and those utilizing formative evaluation as significantly more culturally appropriate. The findings and methodology demonstrated important implications for developers and evaluators of any health-related materials for Spanish speakers and other population groups.

  6. What do implementers need in terms of regulatory safety criteria for the post-closure phase?

    International Nuclear Information System (INIS)

    Cahen, B.

    2010-01-01

    Bruno Cahen, Director Safety Division (ANDRA) presented the point of view of the NEA Integration Group for the Safety Case (IGSC) on 'What do implementers need in terms of regulatory safety criteria for the post-closure phase?' B. Cahen acknowledged that the national experience in siting and developing conceptual designs of geological disposal is growing rapidly. It implies increasing opportunities for interactions between implementers and regulators. There has been large development of international guidance in the recent years. Many regulators have already developed a regulatory framework. The implementers need practical, transparent and deliverable regulations. These regulations should draw on experiences gained from development of geological disposal projects. The IGSC has identified five key questions that the RF may focus on: 1. Over what time frame are the waste deemed to present a hazard? 2. Over what time frames are regulatory criteria applied and do they change over time? 3. Over what time frame(s) are safety assessments required to be conducted? 4. How do implementers have to address uncertainties in the long time frames? 5. What happens after cut-offs: are additional analyses needed? What types of arguments are to be used? Stable, understandable and practical criteria mean, namely, that they need to be developed on a strong scientific and societal basis, that there is consistency of safety options and requirements for different types of waste, that, in the longer time frames, the emphasis is given to robust systems, passive safety and multiple safety functions and that the criteria should fit the various phases of the project (siting, designing, operating, closure and post-closure). Experience feedback from safety cases shows that safety priorities depend very much on time frames. The derived safety criteria for the individual components should lead to measurable, verifiable specifications. The assessment of geological repository post-closure safety

  7. Squale: evaluation criteria of functioning safety; Squale: criteres d`evaluation de la surete de fonctionnement

    Energy Technology Data Exchange (ETDEWEB)

    Deswarte, Y; Kaaniche, M [Centre National de la Recherche Scientifique (CNRS), 31 - Toulouse (France). Laboratoire d` Analyse et d` Architecture des Systemes; Corneillie, P [CE2A-DI, 92 - Courbevoie (France); Benoit, P [Matra Transport International, 92 - Montrouge (France)

    1998-05-01

    The SQUALE (security, safety and quality evaluation for dependable systems) project is part of the ACTS (advanced communications, technologies and services) European program. Its aim is to develop confidence evaluation criteria to test the functioning safety of systems. All industrial sectors that use critical applications (nuclear, railway, aerospace..) are concerned. SQUALE evaluation criteria differ from the classical evaluation methods: they are independent of the application domains and industrial sectors, they take into account the overall functioning safety attributes, and they can progressively change according to the level of severity required. In order to validate the approach and to refine the criteria, a first experiment is in progress with the METEOR automatic underground railway and another will be carried out on a telecommunication system developed by Bouygues company. (J.S.) 15 refs.

  8. Ares-I-X Vehicle Preliminary Range Safety Malfunction Turn Analysis

    Science.gov (United States)

    Beaty, James R.; Starr, Brett R.; Gowan, John W., Jr.

    2008-01-01

    Ares-I-X is the designation given to the flight test version of the Ares-I rocket (also known as the Crew Launch Vehicle - CLV) being developed by NASA. As part of the preliminary flight plan approval process for the test vehicle, a range safety malfunction turn analysis was performed to support the launch area risk assessment and vehicle destruct criteria development processes. Several vehicle failure scenarios were identified which could cause the vehicle trajectory to deviate from its normal flight path, and the effects of these failures were evaluated with an Ares-I-X 6 degrees-of-freedom (6-DOF) digital simulation, using the Program to Optimize Simulated Trajectories Version 2 (POST2) simulation framework. The Ares-I-X simulation analysis provides output files containing vehicle state information, which are used by other risk assessment and vehicle debris trajectory simulation tools to determine the risk to personnel and facilities in the vicinity of the launch area at Kennedy Space Center (KSC), and to develop the vehicle destruct criteria used by the flight test range safety officer. The simulation analysis approach used for this study is described, including descriptions of the failure modes which were considered and the underlying assumptions and ground rules of the study, and preliminary results are presented, determined by analysis of the trajectory deviation of the failure cases, compared with the expected vehicle trajectory.

  9. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  10. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    International Nuclear Information System (INIS)

    Kastenberg, William E.; Blandford, Edward; Kim, Lance

    2009-01-01

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public

  11. Analysis of differences in fuel safety criteria for WWER and western PWR nuclear power plants

    International Nuclear Information System (INIS)

    2003-11-01

    In 2001 the OECD issued a report of the NEA/CSNI (Committee on the Safety of Nuclear Installations) Task Force on the existing safety criteria for reactor fuel for western LWR nuclear power plants (both for PWRs and BWRs) under new design elements. Likewise in 2001, the IAEA released a report by a Working Group on the existing safety criteria for reactor fuel for WWER nuclear power plants under new design requirements. However, it was found that it was not possible to compare the two sets of criteria on the basis upon which they had been established. Therefore, the IAEA initiated an assessment of the common features and differences in fuel safety criteria between plants of eastern and western design, focusing on western PWRs and eastern WWER reactors. Between October 2000 and November 2001, the IAEA organized several workshops with representatives from eastern and western European countries in which the current fuel safety related criteria for PWR and WWER reactors were reviewed and compared. The workshops brought together expert representatives from the Russian Federation, from the Ukraine and from western countries that operate PWRs. The first workshop focused on a general overview of the fuel safety criteria in order for all representatives to appreciate the various criteria and their respective bases. The second workshop (which involved one western and one eastern expert) concentrated on addressing and explaining the differences observed, and documenting all these results in preparation for a panel discussion. This panel discussion took place during the third workshop, where the previously obtained results were reviewed in detail and final recommendations were made. This report documents the findings of the workshops. It highlights the common features and differences between PWR and WWER fuel, and may serve as a general basis for the safety evaluation of these fuels. Therefore, it will be very beneficial for licensing activities for PWR and WWER plants, as it

  12. 30 Years of NRWG activities towards harmonization of nuclear safety criteria and requirements

    International Nuclear Information System (INIS)

    2002-11-01

    This report describes the work performed and the results achieved by the NRWG since its creation in 1972 to advise the Commission on nuclear safety matters (safety methodologies, criteria, standards, postulated accidents inside the nuclear installations, natural hazards, man-made hazards, training of personnel and use of simulator, ALARA policy to reduce the doses to the personnel and the public, emergency planning, defence in depth and integrity of the successive barriers between the radioactive products and the environment, radiological consequences of postulated accidents, probabilistic safety analysis, severe accidents analysis and management. The report also lists a number of technical subjects where NRWG has played a leading role. (author)

  13. 77 FR 58607 - Office of Commercial Space Transportation Safety Approval Performance Criteria

    Science.gov (United States)

    2012-09-21

    ...: Notification of criteria used to evaluate the National Aerospace Training and Research (NASTAR) Center safety... approval for the ability of its Falcon 12/4 Altitude Chamber to replicate pressures experienced at altitude...). NASTAR's Falcon 12/4 Altitude Chamber is capable of replicating any pressure experienced at altitudes...

  14. 78 FR 28275 - Office of Commercial Space Transportation; Safety Approval Performance Criteria

    Science.gov (United States)

    2013-05-14

    ... provide as a service, scenario based physiology training, which includes hypobaric chamber training. BST may offer its scenario based physiology altitude training as a service to a prospective launch and...: Notification of criteria used to evaluate the Black Sky Training, Inc. (BST) safety approval application...

  15. Food safety assurance systems: Microbiological testing, sampling plans, and microbiological criteria

    NARCIS (Netherlands)

    Zwietering, M.H.; Ross, T.; Gorris, L.G.M.

    2014-01-01

    Microbiological criteria give information about the quality or safety of foods. A key component of a microbiological criterion is the sampling plan. Considering: (1) the generally low level of pathogens that are deemed tolerable in foods, (2) large batch sizes, and (3) potentially substantial

  16. Generalized Safety and Efficacy of Simplified Intravenous Thrombolysis Treatment (SMART) Criteria in Acute Ischemic Stroke

    DEFF Research Database (Denmark)

    Sørensen, Sigrid B; Barazangi, Nobl; Chen, Charlene

    2016-01-01

    BACKGROUND: Common intravenous recombinant tissue plasminogen activator (IV rt-PA) exclusion criteria may substantially limit the use of thrombolysis. Preliminary data have shown that the SMART (Simplified Management of Acute stroke using Revised Treatment) criteria greatly expand patient...... eligibility by reducing thrombolysis exclusions, but they have not been assessed on a large scale. We evaluated the safety and efficacy of general adoption of SMART thrombolysis criteria to a large regional stroke network. METHODS: Retrospective analysis of consecutive patients who received IV thrombolysis...... within a regional stroke network was performed. Patients were divided into those receiving thrombolysis locally versus at an outside hospital. The primary outcome was modified Rankin Scale score (≤1) at discharge and the main safety outcome was symptomatic intracranial hemorrhage (sICH) rate. RESULTS...

  17. Radiological Protection Criteria for the Safety of LILW Repository in Croatia

    International Nuclear Information System (INIS)

    Levanat, I.; Lokner, V.; Subasic, D.

    2000-01-01

    Preparations for a LILW repository development in Croatia, conducted by APO Hazardous Waste Management Agency, have reached a point where the first safety assessment of the prospective facility is being attempted. For evaluation of the calculated radiological impact in the assessed option of repository development, a set of radiological protection criteria should be included in the definition of the assessment context. The Croatian regulations do not explicitly require that the repository development be supported by such safety assessment process, and do not provide a specific set of radiological criteria intended for the repository assessment which would be suitable for the constrained optimization of protection. For the initial safety assessment iterations of the prospective repository, which will address long term performance of the facility for various design and other safety options, we propose to use relatively simple radiological protection criteria, consisting only of individual dose and risk constraints for the general population. The numerical values for these constraints are established in accordance with the recognized international recommendations and in compliance with all possibly relevant Croatian safety requirements. (author)

  18. Overview criteria for the environmental, safety and health evaluation of remedial action project planning

    International Nuclear Information System (INIS)

    Stenner, R.D.; Denham, D.H.

    1984-10-01

    Overview criteria (i.e., subject areas requiring review) for evaluating remedial action project plans with respect to environmental, safety and health issues were developed as part of a Department of Energy, Office of Operational Safety, technical support project. Nineteen elements were identified as criteria that should be addressed during the planning process of a remedial action (decontamination and decommissioning) project. The scope was interpreted broadly enough to include such environmental, safety and health issues as public image, legal obligation and quality assurance, as well as more obvious concerns such as those involving the direct protection of public and worker health. The nineteen elements are discussed along with suggested ways to use a data management software system to organize and report results

  19. LMFBR safety criteria and guidelines for consideration in the design of future plants

    International Nuclear Information System (INIS)

    1990-01-01

    For many years the Commission of the European Communities has been conducting activities aimed at the progressive harmonization of safety requirements and criteria applied to nuclear installations in the Community. These activities cover thermal and fast reactors. This publication represents a major achievement in reaching this goal. The document, which has been prepared in the framework of activities of the CEC fast-reactor safety working group (SWG), consists of safety criteria and guidelines for fast reactors. It represents the common view of all EC Member States which have a fast-reactor programme or are interested in fast-reactor development. The criteria and guidelines are structured according to different types of possible faults, such as core reactivity faults, general cooling faults, subassembly faults, faults outside the core and causes external to the station. Only those events are considered which are in the design basis of current fast-reactor projects. Proposed measures or guidelines to satisfy the criteria are based on the present knowledge and proven technology

  20. LMFBR safety criteria: cost-benefit considerations under the constraint of an a priori risk criterion

    International Nuclear Information System (INIS)

    Hartung, J.

    1979-01-01

    The role of cost-benefit considerations and a priori risk criteria as determinants of Core Disruptive Accident (CDA)-related safety criteria for large LMFBR's is explored with the aid of quantitative risk and probabilistic analysis methods. A methodology is described which allows a large number of design and siting alternatives to be traded off against each other with the goal of minimizing energy generation costs subject to the constraint of both an a priori risk criterion and a cost-benefit criterion. Application of this methodology to a specific LMFBR design project is described and the results are discussed. 5 refs

  1. Developing glovebox robotics to meet the national robot safety standard and nuclear safety criteria

    International Nuclear Information System (INIS)

    McMahon, T.T.; Sievers, R.H.

    1991-09-01

    Development of a glove box based robotic system by the Lawrence Livermore National Laboratory (LLNL) is reported. Safety issues addressed include planning to meet the special constraints of operations within a hazardous material glove box and with hostile environments, compliance with the current and draft national robotic system safety standards, and eventual satisfaction of nuclear material handling requirements. Special attention has been required for the revision to the robot and control system models which antedate adoption of the present national safety standard. A robotic test bed, using non-radioactive surrogates is being activated at the Lawrence Livermore National Laboratory to develop the material handling system and the process interfaces for future special nuclear material processing applications. Part of this effort is to define, test, and revise adequate safety controls to ensure success when the system is eventually deployed at a DOE site. The current system is primarily for demonstration and testing, but will evolve into the baseline configuration from which the production system is to be derived. This results in special hazards associated with research activities which may not be present on a production line. Nuclear safety is of paramount importance and has been successfully addressed for 50 years in the DOE weapons production complex. It carries its particular requirements for robot systems and manual operations, as summarized below: Criticality must be avoided (materials cannot consolidate or accumulate to approach a critical mass). Radioactive materials must be confined. The public and workers must be protected from accountable radiation exposure. Nuclear material must be readily retrievable. Nuclear safety must be conclusively demonstrated through hazards analysis. 7 refs

  2. Criteria adopted by the Argentine Nuclear Regulatory Authority for assessing digital systems related to safety

    International Nuclear Information System (INIS)

    Terrado, Carlos A.; Chiossi, Carlos E.; Felizia, Eduardo R.; Roca, Jose L.; Sajaroff, Pedro M.

    2004-01-01

    Following the technological evolution in Instrumentation and Control (I and C) design, analog components are replaced by digital in almost every industry. Due to growing challenges of obsolescence and increasing maintenance costs, licensees of nuclear and radioactive installations are increasingly upgrading or replacing their existing I and C analog systems and components. In existing installations, this involves analog to digital replacements. In new installations design, the use of digital I and C systems is being considered from the very beginning, becoming a good alternative, even in safety applications. Up to now, in Argentina, there is no specific rules for safety-related digital systems, every safety system, analog or digital, must comply with the same generic regulations. The Nuclear Regulatory Authority is now developing criteria to assess digital systems related to safety in nuclear and radioactive installations. In this paper some of those criteria, based on local research and the recognized state of the art, are explained. From a regulatory point of view, the use of digital technology often raises new technical and licensing issues, particularly for safety-related applications. Examples include new failure modes, the potential for common-cause failure of redundant components, electromagnetic interference (EMI), software verification and validation, configuration management and a more exhaustive quality assurance system. The mentioned criteria comprehend the design, operation, maintenance and acquisition of digital systems and components important to safety. The main topics covered are: requirements specifications for digital systems, planning and documentation for digital system development, effectiveness of a digital system, commercial off the shelf (COTS) treatment and considerations involving tools for software development. (author)

  3. Study on safety evaluation for unrestricted recycling criteria of radioactive waste from dismantling operation

    International Nuclear Information System (INIS)

    Yoshimori, Michiro; Ohkoshi, Minoru; Abe, Masayoshi

    1995-01-01

    The study on safety evaluation was done, under contracting with the Science and Technology Agency, for recycling scrap metal arising from dismantling of reactor facilities. An object of this study is to contribute to the examination of establishing criteria and safety regulation for unrestricted recycling steel scrap. To define amount of market flow of iron material in Japan and the amount of radioactive waste generated from dismantling of reactor facilities, investigation had been carried out. On basis of these investigation results and data in several literature, individual doses to workers and to the members of the public have been calculated as well as collective doses. (author)

  4. Attitude of the Korean dentists towards radiation safety and selection criteria

    International Nuclear Information System (INIS)

    Lee, Byung Do; Ludlow, John B.

    2013-01-01

    X-ray exposure should be clinically justified and each exposure should be expected to give patients benefits. Since dental radiographic examination is one of the most frequent radiological procedures, radiation hazard becomes an important public health concern. The purpose of this study was to investigate the attitude of Korean dentists about radiation safety and use of criteria for selecting the frequency and type of radiographic examinations. The study included 267 Korean dentists. Five questions related to radiation safety were asked of each of them. These questions were about factors associated with radiation protection of patients and operators including the use of radiographic selection criteria for intraoral radiographic procedures. The frequency of prescription of routine radiographic examination (an example is a panoramic radiograph for screening process for occult disease) was 34.1%, while that of selective radiography was 64.0%. Dentists' discussion of radiation risk and benefit with patients was infrequent. More than half of the operators held the image receptor by themselves during intraoral radiographic examinations. Lead apron/thyroid collars for patient protection were used by fewer than 22% of dental offices. Rectangular collimation was utilized by fewer than 15% of dental offices. The majority of Korean dentists in the study did not practice radiation protection procedures which would be required to minimize exposure to unnecessary radiation for patients and dental professionals. Mandatory continuing professional education in radiation safety and development of Korean radiographic selection criteria is recommended.

  5. Attitude of the Korean dentists towards radiation safety and selection criteria

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Do [Dept. of Oral and Maxillofacial Radiology and Wonkwang Dental Research Institute, College of Dentistry, Wonkwang University, Iksan (Korea, Republic of); Ludlow, John B. [Graduate Program in Oral and Maxillofacial Radiology, School of Dentistry, University of North Carolina, Chapel Hill (United States)

    2013-09-15

    X-ray exposure should be clinically justified and each exposure should be expected to give patients benefits. Since dental radiographic examination is one of the most frequent radiological procedures, radiation hazard becomes an important public health concern. The purpose of this study was to investigate the attitude of Korean dentists about radiation safety and use of criteria for selecting the frequency and type of radiographic examinations. The study included 267 Korean dentists. Five questions related to radiation safety were asked of each of them. These questions were about factors associated with radiation protection of patients and operators including the use of radiographic selection criteria for intraoral radiographic procedures. The frequency of prescription of routine radiographic examination (an example is a panoramic radiograph for screening process for occult disease) was 34.1%, while that of selective radiography was 64.0%. Dentists' discussion of radiation risk and benefit with patients was infrequent. More than half of the operators held the image receptor by themselves during intraoral radiographic examinations. Lead apron/thyroid collars for patient protection were used by fewer than 22% of dental offices. Rectangular collimation was utilized by fewer than 15% of dental offices. The majority of Korean dentists in the study did not practice radiation protection procedures which would be required to minimize exposure to unnecessary radiation for patients and dental professionals. Mandatory continuing professional education in radiation safety and development of Korean radiographic selection criteria is recommended.

  6. Long-Term Safety Analysis of Baldone Radioactive Waste Repository and Updating of Waste Acceptance Criteria

    International Nuclear Information System (INIS)

    2001-12-01

    The main objective of the project was to provide advice to the Latvian authorities on the safety enhancements and waste acceptance criteria for near surface radioactive waste disposal facilities of the Baldone repository. The project included the following main activities: Analysis of the current status of the management of radioactive waste in Latvia in general and, at the Baldone repository in particular Development of the short and long-term safety analysis of the Baldone repository, including: the planned increasing of capacity for disposal and long term storage, the radiological analysis for the post-closure period Development of the Environment Impact Statement, for the new foreseen installations, considering the non radiological components Proposal of recommendations for future updating of radioactive waste acceptance criteria Proposal of recommendations for safety upgrades to the facility. The work programme has been developed in phases and main tasks as follows. Phase 0: Project inception, Phase 1: Establishment of current status, plans and practices (Legislation, regulation and standards, Radioactive waste management, Waste acceptance criteria), Phase 2: Development of future strategies for long-term safety management and recommendations for safety enhancements. The project team found the general approach use at the installation, the basic design and the operating practices appropriate to international standards. Nevertheless, a number of items subject to potential improvements were also identified. These upgrading recommendations deal with general aspects of the management (mainly storage versus disposal of long-lived sources), site and environmental surveillance, packaging (qualification of containers, waste characterization requirements), the design of an engineered cap and strategies for capping. (author)

  7. NEA perspectives on timescales and criteria in post-closure safety of geological disposal

    International Nuclear Information System (INIS)

    Preter, P. de; Smith, P.; Pescatore, C.; Forinash, B.

    2006-01-01

    A key challenge in the development of safety cases for geological repositories is associated with the long periods of time over which radioactive wastes that are disposed of in repositories remain hazardous. The OECD Nuclear Energy Agency (NEA) has recently examined issues related to timescales in the context of two projects under the auspices of the Radioactive Waste Management Committee (RWMC): the Timescales Initiative and the Long-Term Safety Criteria (LTSC) Initiative. These projects examine, respectively, the treatment of timescales in actual safety cases and in the development of radiological protection criteria for geological disposal. They treat different aspects of timescales but have some overlap and have shown some convergence of the results achieved to date. Based on these projects, this paper examines general considerations in the handling of timescales, including ethical principles, evolution of the hazards of radioactive waste over time, and uncertainty in the evolution of repository systems (including geological features). The implications of these considerations are examined in terms of repository siting; levels of protection in regulations; planning for pre-closure and post-closure actions; and development and presentation of safety cases. A comparison is made with previous NEA work related to timescales, in order to show evolutions in current understanding. (authors)

  8. NEA perspectives on timescales and criteria in post-closure safety of geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    Preter, P. de [ONDRAF/NIRAS, Brussels (Belgium); Smith, P. [Safety Assessment Management Ltd, SAM Ltd. (United Kingdom); Pescatore, C.; Forinash, B. [OECD/NEA, Nuclear Energy Agency, 92 - Issy les Moulineaux (France)

    2006-07-01

    A key challenge in the development of safety cases for geological repositories is associated with the long periods of time over which radioactive wastes that are disposed of in repositories remain hazardous. The OECD Nuclear Energy Agency (NEA) has recently examined issues related to timescales in the context of two projects under the auspices of the Radioactive Waste Management Committee (RWMC): the Timescales Initiative and the Long-Term Safety Criteria (LTSC) Initiative. These projects examine, respectively, the treatment of timescales in actual safety cases and in the development of radiological protection criteria for geological disposal. They treat different aspects of timescales but have some overlap and have shown some convergence of the results achieved to date. Based on these projects, this paper examines general considerations in the handling of timescales, including ethical principles, evolution of the hazards of radioactive waste over time, and uncertainty in the evolution of repository systems (including geological features). The implications of these considerations are examined in terms of repository siting; levels of protection in regulations; planning for pre-closure and post-closure actions; and development and presentation of safety cases. A comparison is made with previous NEA work related to timescales, in order to show evolutions in current understanding. (authors)

  9. Description of present practice concerning the safety criteria for nuclear power plants

    International Nuclear Information System (INIS)

    1977-01-01

    In the description at hand, the authors portray how the aims defined in the safety criteria are reached, and they make proposals for improvement. Basic principles, acceptances and requirements, with which the experts of TUeV and GRS involved in licensing procedures work at the moment, are compiled. This description of present practice has to be adapted perhaps to the existing scientific knowledge at the time. In order that an optimal behaviour as regards safety is reached by the employees in nuclear power plants, criterion 2.5 requires the following measures: the places of work and the work routine in nuclear power plants are to be organized in such a way, that they offer the conditions for the optimal behaviour of employees as regards safety. (orig./HP) [de

  10. Safety criteria for the future LMFBR's in France and main safety issues for the rapide 1500 project

    International Nuclear Information System (INIS)

    Justin, F.; Natta, M.; Orzoni, G.

    1985-04-01

    The main safety criteria for future LMFBR in France and the related issues for the RAPIDE 1500 project are presented and discussed. The evolutions with respect to SUPERPHENIX options and requirements are emphasized, in particular for the concerns of the prevention of core melt accidents, fuel damage limits and related required performances of the protection system, since one main option is not to consider whole core melt accidents in the containment design. One shall also point out the advantages of some mitigating features which were nevertheless added in the containment design, although without any explicit consideration for core melt accidents

  11. Evaluation of hygiene and safety criteria in the production of a traditional Piedmont cheese

    Directory of Open Access Journals (Sweden)

    Sara Astegiano

    2014-08-01

    Full Text Available Traditional products and related processes must be safe to protect consumers’ health. The aim of this study was to evaluate microbiological criteria of a traditional Piedmont cheese, made by two different cheese producers (A and B. Three batches of each cheese were considered. The following seven samples of each batch were collected: raw milk, milk at 38°C, curd, cheese at 7, 30, 60, 90 days of ripening. During cheese making process, training activities dealing with food safety were conducted. Analyses regarding food safety and process hygiene criteria were set up according to the EC Regulation 2073/2005. Other microbiological and chemical-physical analyses [lactic streptococci, lactobacilli, pH and water activity (Aw] were performed as well. Shiga-toxin Escherichia coli, aflatoxin M1 and antimicrobial substances were considered only for raw milk. All samples resulted negative for food safety criteria; Enterobacteriaceae, E.coli and coagulase-positive staphylococci (CPS were high in the first phase of cheese production, however they decreased at the end of ripening. A high level of CPS in milk was found in producer A, likewise in some cheese samples a count of >5 Log CFU/g was reached; staphylococcal enterotoxins resulted negative. The pH and Aw values decreased during the cheese ripening period. The competition between lactic flora and potential pathogen microorganisms and decreasing of pH and Aw are considered positive factors in order to ensure safety of dairy products. Moreover, training activities play a crucial role to manage critical points and perform corrective action. Responsible application of good manufacturing practices are considered key factors to obtain a high hygienic level in dairy products.

  12. Evaluation of Hygiene and Safety Criteria in the Production of a Traditional Piedmont Cheese.

    Science.gov (United States)

    Astegiano, Sara; Bellio, Alberto; Adriano, Daniela; Bianchi, Daniela Manila; Gallina, Silvia; Gorlier, Alessandra; Gramaglia, Monica; Lombardi, Giampiero; Macori, Guerrino; Zuccon, Fabio; Decastelli, Lucia

    2014-08-28

    Traditional products and related processes must be safe to protect consumers' health. The aim of this study was to evaluate microbiological criteria of a traditional Piedmont cheese, made by two different cheese producers (A and B). Three batches of each cheese were considered. The following seven samples of each batch were collected: raw milk, milk at 38°C, curd, cheese at 7, 30, 60, 90 days of ripening. During cheese making process, training activities dealing with food safety were conducted. Analyses regarding food safety and process hygiene criteria were set up according to the EC Regulation 2073/2005. Other microbiological and chemical-physical analyses [lactic streptococci, lactobacilli, pH and water activity (A w )] were performed as well. Shiga-toxin Escherichia coli , aflatoxin M1 and antimicrobial substances were considered only for raw milk. All samples resulted negative for food safety criteria; Enterobacteriaceae , E.coli and coagulase-positive staphylococci (CPS) were high in the first phase of cheese production, however they decreased at the end of ripening. A high level of CPS in milk was found in producer A, likewise in some cheese samples a count of >5 Log CFU/g was reached; staphylococcal enterotoxins resulted negative. The pH and A w values decreased during the cheese ripening period. The competition between lactic flora and potential pathogen microorganisms and decreasing of pH and A w are considered positive factors in order to ensure safety of dairy products. Moreover, training activities play a crucial role to manage critical points and perform corrective action. Responsible application of good manufacturing practices are considered key factors to obtain a high hygienic level in dairy products.

  13. Oak Ridge National Laboratory Health and Safety Long-Range Plan: Fiscal years 1989--1995

    Energy Technology Data Exchange (ETDEWEB)

    1989-06-01

    The health and safety of its personnel is the first concern of ORNL and its management. The ORNL Health and Safety Program has the responsibility for ensuring the health and safety of all individuals assigned to ORNL activities. This document outlines the principal aspects of the ORNL Health and Safety Long-Range Plan and provides a framework for management use in the future development of the health and safety program. Each section of this document is dedicated to one of the health and safety functions (i.e., health physics, industrial hygiene, occupational medicine, industrial safety, nuclear criticality safety, nuclear facility safety, transportation safety, fire protection, and emergency preparedness). Each section includes functional mission and objectives, program requirements and status, a summary of program needs, and program data and funding summary. Highlights of FY 1988 are included.

  14. Oak Ridge National Laboratory Health and Safety Long-Range Plan: Fiscal years 1989--1995

    International Nuclear Information System (INIS)

    1989-06-01

    The health and safety of its personnel is the first concern of ORNL and its management. The ORNL Health and Safety Program has the responsibility for ensuring the health and safety of all individuals assigned to ORNL activities. This document outlines the principal aspects of the ORNL Health and Safety Long-Range Plan and provides a framework for management use in the future development of the health and safety program. Each section of this document is dedicated to one of the health and safety functions (i.e., health physics, industrial hygiene, occupational medicine, industrial safety, nuclear criticality safety, nuclear facility safety, transportation safety, fire protection, and emergency preparedness). Each section includes functional mission and objectives, program requirements and status, a summary of program needs, and program data and funding summary. Highlights of FY 1988 are included

  15. Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belal, Al Momani [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

  16. Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

    International Nuclear Information System (INIS)

    Belal, Al Momani; Jo, Jong Chull

    2013-01-01

    This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

  17. The maternal early warning criteria: a proposal from the national partnership for maternal safety.

    Science.gov (United States)

    Mhyre, Jill M; D'Oria, Robyn; Hameed, Afshan B; Lappen, Justin R; Holley, Sharon L; Hunter, Stephen K; Jones, Robin L; King, Jeffrey C; D'Alton, Mary E

    2014-01-01

    Case reviews of maternal death have revealed a concerning pattern of delay in recognition of hemorrhage, hypertensive crisis, sepsis, venous thromboembolism, and heart failure. Early-warning systems have been proposed to facilitate timely recognition, diagnosis, and treatment for women developing critical illness. A multidisciplinary working group convened by the National Partnership for Maternal Safety used a consensus-based approach to define The Maternal Early Warning Criteria, a list of abnormal parameters that indicate the need for urgent bedside evaluation by a clinician with the capacity to escalate care as necessary in order to pursue diagnostic and therapeutic interventions. This commentary reviews the evidence supporting the use of early-warning systems, describes The Maternal Early Warning Criteria, and provides considerations for local implementation. © 2014 AWHONN, the Association of Women's Health, Obstetric and Neonatal Nurses.

  18. Determination of performance criteria of safety systems in a nuclear power plant via simulated annealing optimization method

    International Nuclear Information System (INIS)

    Jung, Woo Sik

    1993-02-01

    This study presents and efficient methodology that derives design alternatives and performance criteria of safety functions/systems in commercial nuclear power plants. Determination of design alternatives and intermediate-level performance criteria is posed as a reliability allocation problem. The reliability allocation is performed for determination of reliabilities of safety functions/systems from top-level performance criteria. The reliability allocation is a very difficult multi objective optimization problem (MOP) as well as a global optimization problem with many local minima. The weighted Chebyshev norm (WCN) approach in combination with an improved Metropolis algorithm of simulated annealing is developed and applied to the reliability allocation problem. The hierarchy of probabilistic safety criteria (PSC) may consist of three levels, which ranges from the overall top level (e.g., core damage frequency, acute fatality and latent cancer fatality) through the interlnediate level (e.g., unavailiability of safety system/function) to the low level (e.g., unavailability of components, component specifications or human error). In order to determine design alternatives of safety functions/systems and the intermediate-level PSC, the reliability allocation is performed from the top-level PSC. The intermediated level corresponds to an objective space and the top level is related to a risk space. The reliability allocation is performed by means of a concept of two-tier noninferior solutions in the objective and risk spaces within the top-level PSC. In this study, two kinds of towtier noninferior solutions are defined: intolerable intermediate-level PSC and desirable design alternatives of safety functions/systems that are determined from Sets 1 and 2, respectively. Set 1 is obtained by maximizing simultaneously not only safety function/system unavailabilities but also risks. Set 1 reflects safety function/system unavailabilities in the worst case. Hence, the

  19. Nuclear energy generation and the safety criteria for Brazilian power plants

    International Nuclear Information System (INIS)

    Silva, Gustavo Brandão e

    2016-01-01

    The purpose of this paper is to show how the use of nuclear technology can help to diversify the national electricity matrix in a sustainable and efficient way. For this, an analysis of the current situation of the Brazilian electric sector will be made, exposing its fragilities and highlighting the advantages of the nuclear source as an alternative to integrate the necessary thermoelectric base to the reliable supply of electricity in the country. In addition, the objective of the work is to detail the process of exploiting atomic energy in Brazil from raw material mining, through the stages involving the manufacture of nuclear fuel, to the current operation and situation of Brazilian power plants. By taking the Angra 2 Nuclear Power Plant as a case study, the safety criteria adopted in its design and operation will be highlighted. Particular attention will also be given to the electric supply alternatives and to the active safety systems of the plant

  20. Hinkley Point 'C' power station public inquiry: proof of evidence on safety criteria

    International Nuclear Information System (INIS)

    Taylor, R.H.

    1988-09-01

    A public inquiry has been set up to examine the planning application made by the Central Electricity Generating Board (CEGB) for the construction of a 1200 MW Pressurized Water Reactor power station at Hinkley Point (Hinkley Point ''C'') in the United Kingdom. The policy is to replicate the Sizewell ''B'' PWR design which was accepted as safe by an earlier enquiry. In this evidence to the Inquiry, subsequent developments are examined with a view to determining whether these would reverse the Sizewell decision. They are: the possible revision of radiation risk estimates upwards; whether cases of leukaemia occur with greater frequency around nuclear sites than elsewhere; publication of the Health and Safety Executive's consultative document ''The Tolerability of Risk from Nuclear Power Stations''. The overall conclusion is that these developments do not undermine the findings of the Sizewell ''B'' inquiry or the validity of the CEGB's safety criteria. (author)

  1. A consistent approach to assess safety criteria for reactivity initiated accidents

    International Nuclear Information System (INIS)

    Sartoris, C.; Taisne, A.; Petit, M.; Barre, F.; Marchand, O.

    2010-01-01

    In the context of more and more demanding reactor managements, the fuel assembly discharge burn-up increases and raises the question of the current safety criteria relevance. In order to assess new safety criteria for reactivity initiated accidents, the IRSN is developing a consistent and original approach to assess safety. This approach is based on: -A thorough understanding of the physical mechanisms involved in each phase (PCMI and post-boiling phases) of the RIA, supported by the interpretation of the experimental database. This experimental data is constituted of global test outcomes, such as CABRI or Nuclear Safety Research Reactor (NSRR) experiments, and analytical program outcomes, such as PATRICIA tests, intending to understand some particular physical phenomena; -The development of computing codes, modelling the physical phenomena. The physical phenomena observed during the tests mentioned above were modelled in the SCANAIR code. SCANAIR is a thermal-mechanical code calculating fuel and clad temperatures and strains during RIA. The CLARIS module is used as a post-calculation tool to evaluate the clad failure risk based on critical flaw depth. These computing codes were validated by global and analytical tests results; -The development of a methodology. The first step of this methodology is the identification of all the parameters affecting the hydride rim depth. Besides, an envelope curve resulting from burst tests giving the hydride rim depth versus oxidation thickness is defined. After that, the critical flaw depth for a given energy pulse is calculated then compared to the hydride rim depth. This methodology results in an energy or enthalpy limit versus burn-up. This approach is planned to be followed for each phase of the RIA. An example of application is presented to evaluate a PCMI limit for a zircaloy-4 cladding UO 2 rod at Hot Zero Power.

  2. Simplified probabilistic approach to determine safety factors in deterministic flaw acceptance criteria

    International Nuclear Information System (INIS)

    Barthelet, B.; Ardillon, E.

    1997-01-01

    The flaw acceptance rules in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a reliable method of evaluating the safety margins and the integrity of components led Electricite de France to launch a study to link safety factors with requested reliability. A simplified analytical probabilistic approach is developed to analyse the failure risk in Fracture Mechanics. Assuming lognormal distributions of the main random variables, it is possible considering a simple Linear Elastic Fracture Mechanics model, to determine the failure probability as a function of mean values and logarithmic standard deviations. The 'design' failure point can be analytically calculated. Partial safety factors on the main variables (stress, crack size, material toughness) are obtained in relation with reliability target values. The approach is generalized to elastic plastic Fracture Mechanics (piping) by fitting J as a power law function of stress, crack size and yield strength. The simplified approach is validated by detailed probabilistic computations with PROBAN computer program. Assuming reasonable coefficients of variations (logarithmic standard deviations), the method helps to calibrate safety factors for different components taking into account reliability target values in normal, emergency and faulted conditions. Statistical data for the mechanical properties of the main basic materials complement the study. The work involves laboratory results and manufacture data. The results of this study are discussed within a working group of the French in service inspection code RSE-M. (authors)

  3. Relative hazard potential: the basis for definition of safety criteria for fast reactors

    International Nuclear Information System (INIS)

    Cave, L.; Ilberg, D.

    1977-02-01

    One of the main safety criteria to be met for larger thermal reactors is that the probability of exceeding the dose limits imposed by 10 CRF 100 should not be greater than 10 per reactor year. The potential hazard presented by a fast reactor could be substantially greater than that due to an LWR. The potential for harm of a reactor system may be judged by the effects which would arise from a severe accident. Several different types of effects may be considered: number of latent fatal cancers; number of deaths due to acute effects; number of thyroid tumors or nodules; extent of property damage; and genetic effects. Analytical methods for comparison are employed in this paper. A second important parameter reviewed in this report is the radio-toxicity attributed to the various isotopes. It was found that the worst conceivable accident to a 1000 MW(e) fast reactor would lead to effects on health greater by an order of magnitude than the worst accident usually considered for an LWR. Therefore, some reconsideration of the need for additional safety criteria for LMFBRs, as a guide to designers in relation to the control of the effects of very severe accidents, is desirable

  4. Preliminary safety criteria for organic watch list tanks at the Hanford site

    International Nuclear Information System (INIS)

    Webb, A.B.; Stewart, J.L.; Turner, O.A.; Plys, M.G.; Malinovic, B.; Grigsby, J.M.; Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J.

    1995-11-01

    Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended

  5. Preliminary safety criteria for organic watch list tanks at the Hanford site

    Energy Technology Data Exchange (ETDEWEB)

    Webb, A.B.; Stewart, J.L.; Turner, O.A. [Westinghouse Hanford Co., Richland, WA (United States); Plys, M.G.; Malinovic, B. [Fauske and Associates, Inc., Burr Ridge, IL (United States); Grigsby, J.M. [G & P Consulting, Inc. (United States); Camaioni, D.M.; Heasler, P.G.; Samuels, W.O.; Toth, J.J. [Pacific Northwest Lab., Portland, OR (United States)

    1995-11-01

    Condensed-phase, rapid reactions of organic salts with nitrates/nitrites in Hanford High Level Radioactive Waste single-shell tanks could lead to structural failure of the tanks resulting in significant releases of radionuclides and toxic materials. This report establishes appropriate preliminary safety criteria to ensure that tank wastes will be maintained safe. These criteria show that if actual dry wastes contain less than 1.2 MJ/kg of reactants reaction energy or less 4.5 wt % of total organic carbon, then the waste will be safe and will not propagate if ignited. Waste moisture helps to retard reactions; when waste moisture exceeds 20 wt %, rapid reactions are prevented, regardless of organic carbon concentrations. Aging and degradation of waste materials has been considered to predict the types and amounts to organic compounds present in the waste. Using measurements of 3 waste phases (liquid, salt cake, and sludge) obtained from tank waste samples analyzed in the laboratory, analysis of variance (ANOVA) models were used to estimate waste states for unmeasured tanks. The preliminary safety criteria are based upon calorimetry and propagation testing of likely organic compounds which represent actual tank wastes. These included sodium salts of citrate, formate, acetate and hydroxyethylethylenediaminetricetate (HEDTA). Hot cell tests of actual tank wastes are planned for the future to confirm propagation tests performed in the laboratory. The effects of draining liquids from the tanks which would remove liquids and moisture were considered because reactive waste which is too dry may propagate. Evaporation effects which could remove moisture from the tanks were also calculated. The various ways that the waste could be heated or ignited by equipment failures or tank operations activities were considered and appropriate monitoring and controls were recommended.

  6. Criteria for development of a database for safety evaluation of fragrance ingredients.

    Science.gov (United States)

    Ford, R A; Domeyer, B; Easterday, O; Maier, K; Middleton, J

    2000-04-01

    Over 2000 different ingredients are used in the manufacture of fragrances. The majority of these ingredients have been used for many decades. Despite this long history of use, all of these ingredients need continued monitoring to ensure that each ingredient meets acceptable safety standards. As with other large databases of existing chemicals, fulfilling this need requires an organized approach to identify the most important potential hazards. One such approach, specifically considering the dermal route of exposure as the most relevant one for fragrance ingredients, has been developed. This approach provides a rational selection of materials for review and gives guidance for determining the test data that would normally be considered necessary for the elevation of safety under intended conditions of use. As a first step, the process takes into account the following criteria: quantity of use, consumer exposure, and chemical structure. These are then used for the orderly selection of materials for review with higher quantity, higher exposure, and the presence of defined structural alerts all contributing to a higher priority for review. These structural alerts along with certain exposure and volume limits are then used to develop guidelines for determining the quality and quantity of data considered necessary to support an adequate safety evaluation of the chosen materials, taking into account existing data on the substance itself as well as on closely related analogs. This approach can be considered an alternative to testing; therefore, it is designed to be conservative but not so much so as to require excessive effort when not justified.

  7. Nuclear safety criteria applied in site selection - the practice in France

    International Nuclear Information System (INIS)

    Candes, P.; Aussourd, Ph.

    1975-01-01

    In France, the safety of nuclear facilities is the responsibility of the Ministry for Industry and Research (Central Department for the Safety of Nuclear Facilities). The first part of the paper deals with the conception and contents of the site studies which are included in a safety report with the object of obtaining authorization to go ahead with work on the establishment of a facility. The conception is governed by the following two considerations: (a) the site is a place where the natural elements and living organisms occur and which is characterized by the permanent presence of the human factor, while the proposed nuclear facility will - like any industrial facility - present risks and have an impact on the site, particularly through the discharge of radioactive effluent and potentially in consequence of a nuclear accident; (b) the site exercises an influence - in fact, it even imposes constraints - on the nuclear facility. The site study as submitted by the operators to the authorities responsible for the safety evaluation traditionally consists of six sections, covering: (I) description and history of the site; (II) meteorological conditions; (III) hydrology of the area; (IV) geological and seismological conditions; (V) ecological factors; (VI) natural and/or previous radioactivity at the site. These six sections contain the data which serve as a basis for applying the two considerations spelled out above. However, the two corresponding directions of study and analysis do not settle the fundamental problem of the distribution of the population around the site. Methods for dealing with this problem are suggested in the second part of the paper; they take into account the efforts made so far at the international level. The authors consider that limiting criteria should not be based solely on the radioactive effluent discharges associated with normal operation but on the radioactivity releases associated with accidents. The methods proposed by them constitute

  8. Analysis of Driving Safety Criteria Based on National Regulations for the Suspension Systems of NGVs

    Directory of Open Access Journals (Sweden)

    Ronald Mauricio Martinod

    2015-01-01

    Full Text Available The work analyses the technical evaluation process of the suspension system for vehicles that have been adapted to natural-gas-fuelled engines from power light-duty gasoline, and diesel vehicles; this evaluation is done through a mechanical review established by national regulations. The development of this analysis is focused on establishing the relationship between the natural-gas-fuelled equipment and the dynamic effect caused by the extra-weight, according to two measuring criteria that determine the safety and driving comfort, these are: (i tire-road adhesion index; and (ii tire excitation phase angle. The paper also proposes new elements that can be added to the current national regulations and that are currently applied to assess the suspension of natural gas vehicles, recorded using a test standard benchmark for the evaluation of the suspension.

  9. Status, experience and future prospects for the development of probabilistic safety criteria

    International Nuclear Information System (INIS)

    1989-09-01

    During 27-31 January 1986 the IAEA held a Technical Committee Meeting on ''Status, Experience, and Future Prospects for the Development of Probabilistic Safety Criteria''. Participation included representation of essentially all countries with major developments in the area as well as the Nuclear Energy Agency of the OECD and CEC. Though it has to be recognized that in such a short time period it is impossible to resolve or even analyse all aspects of this complex issue, the present situation, the main problems and the directions for future work clearly emerged. This report was prepared by the members of the Technical Committee based on the opinions expressed and on the information available at the time of the meeting. The report also contains 20 papers presented at the meeting by participants. A separate abstract was prepared for each of these 20 papers. Refs, figs and tabs

  10. A Criteria Standard for Conflict Resolution: A Vision for Guaranteeing the Safety of Self-Separation in NextGen

    Science.gov (United States)

    Munoz, Cesar; Butler, Ricky; Narkawicz, Anthony; Maddalon, Jeffrey; Hagen, George

    2010-01-01

    Distributed approaches for conflict resolution rely on analyzing the behavior of each aircraft to ensure that system-wide safety properties are maintained. This paper presents the criteria method, which increases the quality and efficiency of a safety assurance analysis for distributed air traffic concepts. The criteria standard is shown to provide two key safety properties: safe separation when only one aircraft maneuvers and safe separation when both aircraft maneuver at the same time. This approach is complemented with strong guarantees of correct operation through formal verification. To show that an algorithm is correct, i.e., that it always meets its specified safety property, one must only show that the algorithm satisfies the criteria. Once this is done, then the algorithm inherits the safety properties of the criteria. An important consequence of this approach is that there is no requirement that both aircraft execute the same conflict resolution algorithm. Therefore, the criteria approach allows different avionics manufacturers or even different airlines to use different algorithms, each optimized according to their own proprietary concerns.

  11. Safety criteria for the acquisition of meat in Brazilian University restaurants

    Directory of Open Access Journals (Sweden)

    Marizete Oliveira de Mesquita

    2014-03-01

    Full Text Available The present study's objective was to analyze the procedures aimed at guaranteeing sanitary conditions when acquiring meat. The study was conducted with university restaurants of the Federal Institutions of Higher Education (IFES located in the five regions of Brazil. Data were collected using a questionnaire and an evaluation list, which was available online to restaurant professionals. The results showed that restaurants chose one or two types of meat, the most frequent of which were beef and chicken. In restaurants managed by the IFES, the acquisition of raw material occurred by bidding. For vendor selection, the restaurants required product registration with the Inspection Service and requested regulation of the supplier by the Health Surveillance. Monitoring was carried out through a technical visit to the supplier and a review of the past records of the supplier. A higher percentage of restaurants in the Southeast region met appropriate sanitary and hygienic criteria for the receipt of meat, followed by the South, Midwest, Northeast and North. We conclude that restaurants meet most of the safety criteria set in the legislation. However, some weaknesses are evident in the physical and functional structure, the system of transportation of raw material and the records of control measures at the place of reception.

  12. Health-safety and environmental risk assessment of power plants using multi criteria decision making method

    Directory of Open Access Journals (Sweden)

    Jozi Ali Seyed

    2011-01-01

    Full Text Available Growing importance of environmental issues at global and regional levels including pollution of water, air etc. as well as the outcomes such as global warming and climate change has led to being considered environmental aspects as effective factors for power generation. Study ahead, aims at examination of risks resulting from activities of Yazd Combined Cycle Power Plant located in Iran. Method applied in the research is analytical hierarchy process. After identification of factors causing risk, the analytical hierarchy structure of the power plant risks were designed and weight of the criteria and sub-criteria were calculated by intensity probability product using Eigenvector Method and EXPERT CHOICE Software as well. Results indicate that in technological, health-safety, biophysical and socio economic sections of the power plant, factors influenced by the power plant activities like fire and explosion, hearing loss, quantity of groundwater, power generation are among the most important factors causing risk in the power plant. The drop in underground water levels is the most important natural consequence influenced on Yazd Combined Cycle Power Plant.

  13. Criteria of reference radionuclides for safety analysis of spent fuel waste disposal

    International Nuclear Information System (INIS)

    Suryanto

    1998-01-01

    Study on the criteria for reference radionuclides selection for assessment on spent fuel disposal have done. The reference radionuclides in this study means radionuclides are predicted to contribute of the most radiological effect for man if spent fuel waste are discharged on deep geology formation. The research was done by investigate critically of parameters were used on evaluation a kind of radionuclide. Especially, this research study of parameter which relevant disposal case and or spent fuel waste on deep geology formation . The research assumed that spent fuel discharged on deep geology by depth 500-1000 meters from surface of the land. The migration scenario Radionuclides from waste form to man was assumed particularly for normal release in which Radionuclides discharge from waste form in a series thorough container, buffer, geological, rock, to fracture(fault) and move together with ground water go to biosphere and than go into human body. On this scenario, the parameter such as radionuclides inventory, half life, heat generation, hazard index based on maximum permissible concentration (MPC) or annual limit on intake (ALI) was developed as criteria of reference radionuclides selection. The research concluded that radionuclides inventory, half live, heat generated, hazard index base on MPC or ALI can be used as criteria for selection of reference Radionuclide. The research obtained that the main radionuclides are predicted give the most radiological effect to human are as Cs-137, Sr-90, I-129, Am-243, Cm-244, Pu-238, Pu-239, Pu-240. The radionuclides reasonable to be used as reference radionuclides in safety analysis at spent fuel disposal. (author)

  14. Development of the Human Error Management Criteria and the Job Aptitude Evaluation Criteria for Rail Safety Personnel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, In Soo; Seo, Sang Mun; Park, Geun Ok (and others)

    2008-08-15

    It has been estimated that up to 90% of all workplace accidents have human error as a cause. Human error has been widely recognized as a key factor in almost all the highly publicized accidents, including Daegu subway fire of February 18, 2003 killed 198 people and injured 147. Because most human behavior is 'unintentional', carried out automatically, root causes of human error should be carefully investigated and regulated by a legal authority. The final goal of this study is to set up some regulatory guidance that are supposed to be used by the korean rail organizations related to safety managements and the contents are : - to develop the regulatory guidance for managing human error, - to develop the regulatory guidance for managing qualifications of rail drivers - to develop the regulatory guidance for evaluating the aptitude of the safety-related personnel.

  15. Development of the Human Error Management Criteria and the Job Aptitude Evaluation Criteria for Rail Safety Personnel

    International Nuclear Information System (INIS)

    Koo, In Soo; Seo, Sang Mun; Park, Geun Ok

    2008-08-01

    It has been estimated that up to 90% of all workplace accidents have human error as a cause. Human error has been widely recognized as a key factor in almost all the highly publicized accidents, including Daegu subway fire of February 18, 2003 killed 198 people and injured 147. Because most human behavior is 'unintentional', carried out automatically, root causes of human error should be carefully investigated and regulated by a legal authority. The final goal of this study is to set up some regulatory guidance that are supposed to be used by the korean rail organizations related to safety managements and the contents are : - to develop the regulatory guidance for managing human error, - to develop the regulatory guidance for managing qualifications of rail drivers - to develop the regulatory guidance for evaluating the aptitude of the safety-related personnel

  16. Criteria for Use in Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Guide (Arabic Edition)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-11-01

    This Safety Guide presents a coherent set of generic criteria (expressed numerically in terms of radiation dose) that form a basis for developing the operational levels needed for decision making concerning protective and response actions. The set of generic criteria addresses the requirements established in IAEA Safety Standards Series No. GS-R-2 for emergency preparedness and response, including lessons learned from responses to past emergencies, and provides an internally consistent foundation for the application of principles of radiation protection. The publication also provides a basis for a plain language explanation of the criteria for the public and for public officials. Contents: 1. Introduction; 2. Basic considerations; 3. Framework for emergency response criteria; 4. Guidance values for emergency workers; 5. Operational criteria; Appendix I: Dose concepts and dosimetric quantities; Appendix II: Examples of default OILs for deposition, individual contamination and contamination of food, milk and water; Appendix III: Development of EALs and example EALs for light water reactors; Appendix IV: Observables on the scene of a radiological emergency.

  17. Criteria for Use in Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2013-01-01

    This Safety Guide presents a coherent set of generic criteria (expressed numerically in terms of radiation dose) that form a basis for developing the operational levels needed for decision making concerning protective and response actions. The set of generic criteria addresses the requirements established in IAEA Safety Standards Series No. GS-R-2 for emergency preparedness and response, including lessons learned from responses to past emergencies, and provides an internally consistent foundation for the application of radiation protection. The publication also proposes a basis for a plain language explanation of the criteria for the public and for public officials. Contents: 1. Introduction; 2. Basic considerations; 3. Framework for emergency response criteria; 4. Guidance values for emergency workers; 5. Operational criteria; Appendix I: Dose concepts and dosimetric quantities; Appendix II: Examples of default oils for deposition, individual monitoring and contamination of food, milk and water; Appendix III: Development of EALs and example EALs for light water reactors; Appendix IV: Observables at the scene of a nuclear or radiological emergency

  18. Criteria for Use in Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Guide (Russian Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide presents a coherent set of generic criteria (expressed numerically in terms of radiation dose) that form a basis for developing the operational levels needed for decision making concerning protective and response actions. The set of generic criteria addresses the requirements established in IAEA Safety Standards Series No. GS-R-2 for emergency preparedness and response, including lessons learned from responses to past emergencies, and provides an internally consistent foundation for the application of radiation protection. The publication also proposes a basis for a plain language explanation of the criteria for the public and for public officials. Contents: 1. Introduction; 2. Basic considerations; 3. Framework for emergency response criteria; 4. Guidance values for emergency workers; 5. Operational criteria; Appendix I: Dose concepts and dosimetric quantities; Appendix II: Examples of default oils for deposition, individual monitoring and contamination of food, milk and water; Appendix III: Development of EALs and example EALs for light water reactors; Appendix IV: Observables at the scene of a nuclear or radiological emergency.

  19. Safety design criteria for the next generation Sodium-cooled fast reactors based on lessons learned from the Fukushima NPS accident

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2012-01-01

    In this presentation, architecture of the safety design criteria as requirements for SFR system and the activities on safety research works to establish safety evaluation methods for the next generation SFRs are summarized with the basis on lessons learned from the Fukushima NPS accident. Nuclear safety is a grovel issue which should be achieved by the international cooperation. In respect of the development for the next generation reactor, it is necessary to build the harmonized safety criteria and evaluation methods to establish the next level of safety

  20. Strain-based plastic instability acceptance criteria for ferritic steel safety class 1 nuclear components under level D

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji Su; Lee, Han Sang; Kim, Yun Jae [Dept. of Mechanical Engineering, Korea University, Seoul (Korea, Republic of); Kim, Jong Sung [Dept. of Mechanical Engineering, Sunchon National University, Suncheon (Korea, Republic of); Kim, Jin Won [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-04-15

    This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1) a section average approach and (2) a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA) for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV) outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the over conservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

  1. Strain-based plastic instability acceptance criteria for ferritic steel safety class 1 nuclear components under level D

    International Nuclear Information System (INIS)

    Kim, Ji Su; Lee, Han Sang; Kim, Yun Jae; Kim, Jong Sung; Kim, Jin Won

    2015-01-01

    This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1) a section average approach and (2) a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA) for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV) outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the over conservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

  2. Spent nuclear fuel project-criteria document Cold Vacuum Drying Facility phase 2 safety analysis report

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1998-01-01

    The criteria document provides the criteria and guidance for developing the SNF CVDF Phase 2 SAR. This SAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, and installation acceptance testing of the CVDF systems

  3. Soil criteria to protect terrestrial wildlife and open-range livestock from metal toxicity at mining sites.

    Science.gov (United States)

    Ford, Karl L; Beyer, W Nelson

    2014-03-01

    Thousands of hard rock mines exist in the western USA and in other parts of the world as a result of historic and current gold, silver, lead, and mercury mining. Many of these sites in the USA are on public lands. Typical mine waste associated with these sites are tailings and waste rock dumps that may be used by wildlife and open-range livestock. This report provides wildlife screening criteria levels for metals in soil and mine waste to evaluate risk and to determine the need for site-specific risk assessment, remediation, or a change in management practices. The screening levels are calculated from toxicity reference values based on maximum tolerable levels of metals in feed, on soil and plant ingestion rates, and on soil to plant uptake factors for a variety of receptors. The metals chosen for this report are common toxic metals found at mining sites: arsenic, cadmium, copper, lead, mercury, and zinc. The resulting soil screening values are well above those developed by the US Environmental Protection Agency. The difference in values was mainly a result of using toxicity reference values that were more specific to the receptors addressed rather than the most sensitive receptor.

  4. Conformationally selective multidimensional chemical shift ranges in proteins from a PACSY database purged using intrinsic quality criteria

    International Nuclear Information System (INIS)

    Fritzsching, Keith J.; Hong, Mei; Schmidt-Rohr, Klaus

    2016-01-01

    We have determined refined multidimensional chemical shift ranges for intra-residue correlations ( 13 C– 13 C, 15 N– 13 C, etc.) in proteins, which can be used to gain type-assignment and/or secondary-structure information from experimental NMR spectra. The chemical-shift ranges are the result of a statistical analysis of the PACSY database of >3000 proteins with 3D structures (1,200,207 13 C chemical shifts and >3 million chemical shifts in total); these data were originally derived from the Biological Magnetic Resonance Data Bank. Using relatively simple non-parametric statistics to find peak maxima in the distributions of helix, sheet, coil and turn chemical shifts, and without the use of limited “hand-picked” data sets, we show that ∼94 % of the 13 C NMR data and almost all 15 N data are quite accurately referenced and assigned, with smaller standard deviations (0.2 and 0.8 ppm, respectively) than recognized previously. On the other hand, approximately 6 % of the 13 C chemical shift data in the PACSY database are shown to be clearly misreferenced, mostly by ca. −2.4 ppm. The removal of the misreferenced data and other outliers by this purging by intrinsic quality criteria (PIQC) allows for reliable identification of secondary maxima in the two-dimensional chemical-shift distributions already pre-separated by secondary structure. We demonstrate that some of these correspond to specific regions in the Ramachandran plot, including left-handed helix dihedral angles, reflect unusual hydrogen bonding, or are due to the influence of a following proline residue. With appropriate smoothing, significantly more tightly defined chemical shift ranges are obtained for each amino acid type in the different secondary structures. These chemical shift ranges, which may be defined at any statistical threshold, can be used for amino-acid type assignment and secondary-structure analysis of chemical shifts from intra-residue cross peaks by inspection or by using a

  5. Conformationally selective multidimensional chemical shift ranges in proteins from a PACSY database purged using intrinsic quality criteria

    Energy Technology Data Exchange (ETDEWEB)

    Fritzsching, Keith J., E-mail: kfritzsc@brandeis.edu [Brandeis University, Department of Chemistry (United States); Hong, Mei [Massachusetts Institute of Technology, Department of Chemistry (United States); Schmidt-Rohr, Klaus, E-mail: srohr@brandeis.edu [Brandeis University, Department of Chemistry (United States)

    2016-02-15

    We have determined refined multidimensional chemical shift ranges for intra-residue correlations ({sup 13}C–{sup 13}C, {sup 15}N–{sup 13}C, etc.) in proteins, which can be used to gain type-assignment and/or secondary-structure information from experimental NMR spectra. The chemical-shift ranges are the result of a statistical analysis of the PACSY database of >3000 proteins with 3D structures (1,200,207 {sup 13}C chemical shifts and >3 million chemical shifts in total); these data were originally derived from the Biological Magnetic Resonance Data Bank. Using relatively simple non-parametric statistics to find peak maxima in the distributions of helix, sheet, coil and turn chemical shifts, and without the use of limited “hand-picked” data sets, we show that ∼94 % of the {sup 13}C NMR data and almost all {sup 15}N data are quite accurately referenced and assigned, with smaller standard deviations (0.2 and 0.8 ppm, respectively) than recognized previously. On the other hand, approximately 6 % of the {sup 13}C chemical shift data in the PACSY database are shown to be clearly misreferenced, mostly by ca. −2.4 ppm. The removal of the misreferenced data and other outliers by this purging by intrinsic quality criteria (PIQC) allows for reliable identification of secondary maxima in the two-dimensional chemical-shift distributions already pre-separated by secondary structure. We demonstrate that some of these correspond to specific regions in the Ramachandran plot, including left-handed helix dihedral angles, reflect unusual hydrogen bonding, or are due to the influence of a following proline residue. With appropriate smoothing, significantly more tightly defined chemical shift ranges are obtained for each amino acid type in the different secondary structures. These chemical shift ranges, which may be defined at any statistical threshold, can be used for amino-acid type assignment and secondary-structure analysis of chemical shifts from intra

  6. Third Joint GIF–IAEA Workshop on Safety Design Criteria for Sodium-Cooled Fast Reactors, 26-27 February 2013, Vienna, Austria. Summary Report

    International Nuclear Information System (INIS)

    2013-01-01

    The main objectives of the meeting were to: • Present and share information on the work carried out by GIF, the IAEA and the Member States on the definition of safety design criteria for SFR, including safety approach and requirements on general plant design; • Present the document prepared by the GIF-SFR Task Force on Safety Design Criteria; • Present and discuss safety design concepts of SFRs under development in Member States, with particular emphasis on design measures against Design Basis Accidents and Design Extended Conditions, as well as the associated safety evaluations and supporting R&D; • Draft a room document which should be the basis of the discussion for the Panel on Safety Design Criteria of the FR13 Conference in Paris. • Discuss the results and agree on the future actions of the 3rd Joint GIF-IAEA Workshop on Safety of Sodium-Cooled Fast Reactors

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

    International Nuclear Information System (INIS)

    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

  8. PROBLEMS OF APPLYING FIXED FORMULAE TO SAFETY CRITERIA AND SITE SELECTION

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. K.

    1963-10-15

    The problem of developing a formula or calculation procedure for that could more-or-less automatically indicate whether or not a nuclear plant would be considered safe at a particular location is discussed. The difficulties and impossibilities of any sach formula for making decisions on siting and safety involving large amounts of money and public safety are considered. (P.C.H.)

  9. Safety indicators for the safety assessment of radioactive waste disposal. Sixth report of the Working Group on Principles and Criteria for Radioactive Waste Disposal

    International Nuclear Information System (INIS)

    2003-09-01

    The report describes a few indicators that are considered to be the most promising for assessing the long term safety of disposal systems. The safety indicators that are discussed here may be applicable to a range of disposal systems for different waste types, including near surface disposal facilities for low level waste. The appropriateness of the different indicators may, however, vary depending on the characteristics of the waste, the facility and the assessment context. The focus of the report is thus on the use of time-scales of containment and transport, and radionuclide concentrations and fluxes, as indicators of disposal system safety, that may complement the more usual safety indicators of dose and risk. Summarised are the broad elements that a safety case for an underground radioactive waste disposal facility should possess and the role and use of performance and safety indicators within these elements. An overview of performance and safety indicators is given. The use is discussed of dose and risk as safety indicators and, in particular, problems that can arise in their use. Also presented are some specific indicators that have the potential to be used as complementary safety indicators. Discussed is also how fluxes of naturally occurring elements and radionuclides due to the operation of natural processes such as erosion and groundwater discharge may be quantified for comparison with fluxes of waste derived contaminants

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  11. The use of criteria in the regulatory safety analysis in France

    International Nuclear Information System (INIS)

    Queniart, D.

    1988-12-01

    This paper describes the framework set up in France to allow continuous technical dialogue between operators and safety organizations. The operators, who have primary responsibility for the safety of their installations, propose the measures implemented, or to be implemented, in their installations. Each of these measures is then subjected to a detailed technical examination carried out by the Institute for Nuclear Safety and Protection, without reference to any technical regulations defined a priori. This approach has resulted, particularly in the case of pressurized water reactors (PWRs), in significant progress in the field of safety. This has been achieved by progressively completing the initial approach, derived from American practice for PWR plants, by probabilistic considerations, by a specific approach to severe accidents and by constant use of experience feedback. This last method seems particularly fruitful, and there would appear to be a need also for an indepth study of containment

  12. 15 CFR 970.801 - Criteria for safety of life and property at sea.

    Science.gov (United States)

    2010-01-01

    ... REGULATIONS OF THE ENVIRONMENTAL DATA SERVICE DEEP SEABED MINING REGULATIONS FOR EXPLORATION LICENSES Safety... jurisdiction on the high seas and subject to domestic enforcement procedures. With respect to foreign flag...

  13. 25 Years of Community Activities towards Harmonization of Nuclear Safety Criteria and Requirements - Achievements and Prospects

    International Nuclear Information System (INIS)

    Lillington, J.N.; Turland, B.D.; Haste, T.J.; Seiler, J.M.; Carretero, A.; Perez, T.; Geutges, A.; Van Hienen, J.F.A.; Jehee, J.N.T.; Sehgal, B.R.; Mattila, L.; Holmstrom, H.; Karwat, H.; Maroti, L.; Toth, I.; Husarcek, J.

    2001-10-01

    The main objective was to advise the EC on future challenges and opportunities in terms of enhanced co-operation in the area of nuclear safety and harmonization of safety requirements and practices in an enlarged European Union. The activities were divided into 3 sub-tasks as follows: part A, to prepare an analysis, synthesis and assessment of the main achievements from Community activities related to the Resolutions on the technological problems of nuclear safety of 1975 and 1992, with due consideration for related research activities; part B, to prepare an overview of safety philosophies and practices in EU Member States, taking account of their specific national practices in terms of legal framework, type and age of operating nuclear reactors; part C, to provide elements of a strategy for future activities in the frame of the Council Resolutions, with particular attention to the context of enlargement of the EU. (author)

  14. Safety Requirements / Design Criteria for SFR. Lessons Learned from the Fukushima Dai-ichi Accident

    International Nuclear Information System (INIS)

    Yllera, Javier

    2013-01-01

    After the Fukushima event (March 2011) the IAEA has started an action to review and revise, if necessary, all Safety Standards to take into consideration the lessons learned from the accident. The Safety Standards that need to be revised have been identified. A Prioritization Approach has been established: The first priority is to review safety guides applicable for NPPs and spent fuel storage with focus on the measures for the prevention and mitigation of severe accident due to external hazards - ● Regulatory framework, Safety assessment, Management system, Radiation protection and Emergency Preparedness and response; ● Sitting, Design, Operation of NPPs ● Decommissioning and Waste Management. Original sources for lessons learned: IAE fact Finding Mission, Japan´s report to the Ministerial Conference, INSAG Report, etc. Later, other lesson sources considered

  15. Multi-criteria analysis for evaluating the radiological and ecological safety measures in radioactive waste management

    International Nuclear Information System (INIS)

    Sazykina, T.G.; Kryshev, I.I.

    2006-01-01

    A methodological approach is presented for multicriterial evaluating the effectiveness of radiation ecological safety measures during radioactive waste management. The approach is based on multicriterial analysis with consideration of radiological, ecological, social, economical consequences of various safety measures. The application of the multicriterial approach is demonstrated taking as an example of decision-making on the most effective actions for rehabilitation of a water subject, contaminated with radionuclides [ru

  16. Safety principles and technical criteria for the underground disposal of high level radioactive wastes

    International Nuclear Information System (INIS)

    1989-01-01

    The main objective of this book is to set out an internationally agreed set of principles and criteria for the design of deep underground repositories for the disposal of high level radioactive wastes. This book is concerned with the post-closure period. Consideration of the operational requirements which must be met when wastes are being handled, stored and emplaced are not therefore included

  17. Use of decision criteria based on expected values to support decision-making in a production assurance and safety setting

    International Nuclear Information System (INIS)

    Aven, T.; Flage, R.

    2009-01-01

    We consider decision problems related to production assurance and safety. The issue is to what extent we should use decision criteria based on expected values, such as the expected net present value (E[NPV]) and the expected cost per expected number of saved lives (ICAF), to guide the decision. Such criteria are recognised as practical tools for supporting decision-making under uncertainty, but is uncertainty adequately taken into account by these criteria? Based on the prevailing practice and the existing literature, we conclude that there is a need for a clarification of the rationale of these criteria. Adjustments of the standard approaches have been suggested to reflect risks and uncertainties, but can cautionary and precautionary concerns be replaced by formulae and mechanical procedures? These issues are discussed in the present paper, particularly addressing the company level. We argue that the search for such formulae and procedures should be replaced by a more balanced perspective acknowledging that there will always be a need for management review and judgment beyond the realm of the analyses. Most of the suggested adjustments of the E[NPV] and ICAF approaches should be avoided. They add more confusion than value.

  18. Technical regulations on the general design and safety criteria for design and construction of nuclear reactors of May 1975

    International Nuclear Information System (INIS)

    1975-05-01

    These Technical Regulations published on 5th September 1975 were made in implementation of Section 33 of Decree No 7/9141 on the procedure for the licensing of nuclear installations. They serve as a guide to licensing authorities, project designers and operators in the nuclear field and therefore provide general criteria for safety standards, engineering codes, siting considerations, design bases for overall environmental radiation protection, and also deal with reactor core design, instrumentation, control, alarm systems, including an emergency core cooling system. Finally, the safe design of fuel elements must be ensured and fuel storage and handling techniques complied with. (NEA) [fr

  19. Proposal of criteria for evaluation of engineering safety factors of VVER core parameters

    International Nuclear Information System (INIS)

    Shishkov, L.; Tsyganov, S.; Dementiev, V.

    2009-01-01

    The paper states that the regulatory documentation, as a rule, do not give explicit recommendations on formation techniques of engineering safety factors for design limited parameters of normal operation (K eng ). The AER countries use different approaches to K eng evaluation (sometimes even one country in relation of various power units). The paper suggests the development of uniform rules to be used in calculation of engineering safety factor for all VVER reactors. The paper presents principal problems that must be solved in the course of the discussion, and in the form of an exercise suggests the way of their solution. (authors)

  20. Probabilistic safety criteria for improvement of Nuclear Power Plant design and operation

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Chung, Woo Sick; Park, Moon Kyu [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The procedure of this study is to : research on the status of IAEA(International Atomic Energy Agency) member states about the policy of safety goals, study figures of merit and demerit that inherently exist in the existing methodology for reliability allocation, develop an efficient methodology for allocating reliability from top-level safety goals to intermediate and low-level PSC, write a computer code on the basis of the methodology proposed in the study, and apply the methodology to Surry Unit 1 that is the type of PWR.

  1. Proposal of criteria for evaluation of engineering safety factors of WWER core parameters

    International Nuclear Information System (INIS)

    Shishkov, L.; Tsyganov, S.; Dementiev, V.

    2009-01-01

    The paper states that the regulatory documentation, as a rule, do not give explicit recommendations on formation techniques of engineering safety factors for design limited parameters of normal operation. The AER countries use different approaches to evaluation (sometimes even one country in relation of various power units). The paper suggests the development of uniform rules to be used in calculation of engineering safety factor for all WWER reactors. The paper presents principal problems that must be solved in the course of the discussion, and in the form of an exercise suggests the way of their solution. (Authors)

  2. K Basin sludge packaging design criteria (PDC) and safety analysis report for packaging (SARP) approval plan

    International Nuclear Information System (INIS)

    Brisbin, S.A.

    1996-01-01

    This document delineates the plan for preparation, review, and approval of the Packaging Design Crieteria for the K Basin Sludge Transportation System and the Associated on-site Safety Analysis Report for Packaging. The transportation system addressed in the subject documents will be used to transport sludge from the K Basins using bulk packaging

  3. Designing sustainable concrete on the basis of equivalence performance: assessment criteria for safety

    NARCIS (Netherlands)

    Visser, J.H.M.; Bigaj, A.J.

    2014-01-01

    In order not to hampers innovations, the Dutch National Building Regulations (NBR), allow an alternative approval route for new building materials. It is based on the principles of equivalent performance which states that if the solution proposed can be proven to have the same level of safety,

  4. Understanding the differences amongst national regulatory criteria for the long-term safety of radioactive waste disposal

    International Nuclear Information System (INIS)

    Larsson, C.M.; Ferch, R.; Pescatore, C.

    2008-01-01

    Carl-Magnus Larsson detailed then the work of the Regulators' Forum and the origin of the LTSC initiative. He explained that one of the objectives of the LTSC was to identify a set of issues on long-term protection criteria and collate findings in a report. He explained why the idea of a 'collective opinion' was abandoned and why it should be replaced by a common understanding where differences between countries ought to be explained and understood. C.-M. Larsson detailed the different types of approaches to regulating long-term safety and the different approaches for numerical targets. He gave some explanations of the reasons for the differences in regulatory targets between countries (level of conservatism, progress in the safety case methodology, etc.). The regulatory function takes into account the nature of the demonstration (illustrations and societal demands). C.-M. Larsson referred to the evolution of IAEA safety fundamentals and stressed that the 'sustainability' concept, introduced by the Joint Convention, is not mentioned in the new safety standard. The term 'adequately protected' is now preferred in relation to future generations. The ICRP recommends that less emphasis be placed on assessment of doses in the long term. C.-M. Larsson concluded that one of the challenges for the regulator is not to promise nor require the impossible. (authors)

  5. Human Factors engineering criteria and design for the Hanford Waste Vitrification Plant preliminary safety analysis report

    International Nuclear Information System (INIS)

    Wise, J.A.; Schur, A.; Stitzel, J.C.L.

    1993-09-01

    This report provides a rationale and systematic methodology for bringing Human Factors into the safety design and operations of the Hanford Waste Vitrification Plant (HWVP). Human Factors focuses on how people perform work with tools and machine systems in designed settings. When the design of machine systems and settings take into account the capabilities and limitations of the individuals who use them, human performance can be enhanced while protecting against susceptibility to human error. The inclusion of Human Factors in the safety design of the HWVP is an essential ingredient to safe operation of the facility. The HWVP is a new construction, nonreactor nuclear facility designed to process radioactive wastes held in underground storage tanks into glass logs for permanent disposal. Its design and mission offer new opposites for implementing Human Factors while requiring some means for ensuring that the Human Factors assessments are sound, comprehensive, and appropriately directed

  6. Multi-criteria Generation-Expansion Planning with Carbon dioxide emissions and Nuclear Safety considerations

    International Nuclear Information System (INIS)

    Lee, Hun Gyu; Kim, Young Chang

    2010-01-01

    A multiple criteria decision making (MCDM) method is developed to aid decision makers in Generation Expansion planning and management. Traditionally, the prime objective of an electric utility's generation-expansion planning has been to determine the minimum cost supply plans that meet expected demands over a planning horizon (typically 10 to 30 years). Today, however, the nature of decision environments has changed substantially. Increased policy attention is given to solve the multiple tradeoff function including environmental and social factors as well as economic one related to nuclear power expansion. In order to deal with this MCDM problem, the Analytic Hierarchy Process (AHP) Model is applied

  7. Determined analysis of safety, viability and residual service life on criteria of crack mechanics

    International Nuclear Information System (INIS)

    Matvienko, Yu.G.

    1997-01-01

    Unified methods used in analysis of reliability, vulnerability, and residual lifetime of equipment with crack damage are considered, an increase in the desired lifetime is proven in the framework of vulnerability concept that allows crack developing with regard to the given level of reliability. The problem of reliability, vulnerability, and the lifetime is shown to be an interrelated problem. Optimal combination of the strength value, plasticity and resistance to crack developing results from the criteria of reliability and vulnerability based, in turn, on the principles of the mechanics of cracks. Structural features of technical systems can hinder the crack developing and prevent drastic damages of the equipment thus increasing the lifetime

  8. The application of modern safety criteria to restarting and operating the USDOE K-Reactor

    International Nuclear Information System (INIS)

    Dimenna, R.A.; Taylor, G.A.; Brandyberry, M.D.

    1993-01-01

    The United States Department of Energy's (USDOE's) K-reactor, a defense production reactor located at the Savannah River Site in Aiken, South Carolina, was shut down in the summer of 1988 for safety upgrades to bring it into conformance with modern safety standards prior to restart. Over the course of the succeeding four years, all aspects of the 35-year old reactor, including hardware, operations, and analysis, were upgraded to ensure that the reactor could operate safely according to standards similar to those applied to modern nuclear reactors. This paper describes the decision making processes by which issues were identified, priorities assigned, and analysis improved to enhance reactor safety. Special emphasis is given to the probabilistic risk assessment (PRA) decision making processes used to quantify the risks and consequences of operating the K-reactor, the analytical hierarchy process (AHP) used to identify key phenomena, and modifications made to the RELAP5 computer code to make it applicable to K-reactor analysis. The success of the project was demonstrated when the K-reactor was restarted in the summer of 1992

  9. Criteria for the Research Institute for Fragrance Materials, Inc. (RIFM) safety evaluation process for fragrance ingredients.

    Science.gov (United States)

    Api, A M; Belsito, D; Bruze, M; Cadby, P; Calow, P; Dagli, M L; Dekant, W; Ellis, G; Fryer, A D; Fukayama, M; Griem, P; Hickey, C; Kromidas, L; Lalko, J F; Liebler, D C; Miyachi, Y; Politano, V T; Renskers, K; Ritacco, G; Salvito, D; Schultz, T W; Sipes, I G; Smith, B; Vitale, D; Wilcox, D K

    2015-08-01

    The Research Institute for Fragrance Materials, Inc. (RIFM) has been engaged in the generation and evaluation of safety data for fragrance materials since its inception over 45 years ago. Over time, RIFM's approach to gathering data, estimating exposure and assessing safety has evolved as the tools for risk assessment evolved. This publication is designed to update the RIFM safety assessment process, which follows a series of decision trees, reflecting advances in approaches in risk assessment and new and classical toxicological methodologies employed by RIFM over the past ten years. These changes include incorporating 1) new scientific information including a framework for choosing structural analogs, 2) consideration of the Threshold of Toxicological Concern (TTC), 3) the Quantitative Risk Assessment (QRA) for dermal sensitization, 4) the respiratory route of exposure, 5) aggregate exposure assessment methodology, 6) the latest methodology and approaches to risk assessments, 7) the latest alternatives to animal testing methodology and 8) environmental risk assessment. The assessment begins with a thorough analysis of existing data followed by in silico analysis, identification of 'read across' analogs, generation of additional data through in vitro testing as well as consideration of the TTC approach. If necessary, risk management may be considered. Copyright © 2014 Elsevier Ltd. All rights reserved.

  10. The development of safety criteria for use in the nuclear industry

    International Nuclear Information System (INIS)

    Higson, D.J.

    1978-01-01

    Limits to routine radiation exposure have been laid down in the health regulations of industrial nations and provide a basis for the safe operation of nuclear power stations, uranium mines and other nuclear installations. However, these limits do not take account of the possibility of accidents, which may also be a major concern in the sitting and design of plants. In this paper specific limits to fatal accident frequencies are recommended. An indication of the required level of safety has been derived from the records of other industries and human activities which are already regarded as safe

  11. Rethinking the Zircaloy Embrittlement Criteria and Its Impact on Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Lee, You Ho; Kim, Bo Kyung; No, Hee Cheon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    These fuel rod failure modes include integral thermal shock fracture, and impact tests. It is quite remarkable to see that the proposed Zircaloy embrittlemt criteria attained from ring compression tests, in general, successfully assure structural integrity of fuel rods subject to relevant failure modes in accidents. This fact demonstrates that ductility of Zircaloy is the key metric to structural integrity of fuel rods. However, the Zircaloy embrittlement criteria set in 1970s inevitably pose limitations that have become increasingly important for today's nuclear fuel and reactor operations. In particular, the criteria do not take into account the steady-state hydrogen embrittlement with burnup. This may be understandable considering the markedly lower discharge burnup in 1970s compared to that of today. The revision of the rule has been already conducted by the U.S NRC to account for high burnup effects on ECR while the temperature limit remains unchanged. The newly proposed rule of the U.S NRC stick to the similar ring compression tests conducted in the early 1970s. In the monumental experimental investigation of Hobson and Rittenhouse in 1972 and 1973, the experimental evidence for the current 1204oC was first addressed. The study found a reasonably accurate correlation between zero ductility temperature and the sum of alpha and oxide layer thickness for the specimens oxidized below 2200oF (1204 .deg. C). However, in spite of the similar oxidation degree, specimens oxidized at 2400 .deg. F (1315 deg. C) were markedly more brittle than specimens oxidized at 2200 .deg. F (1204 .deg. C). The study explained this by the increase in solid-solution hardening due to a higher oxygen solubility at a higher temperature. Such a nice experimental correlation attained between the nil ductility temperature and the remaining beta layer thickness fraction below 1204 .deg. C has become a critical basis for the current temperature limit; at 1315 .deg. C- thecorrelation

  12. The historical development of criteria on the safety of nuclear power plants

    International Nuclear Information System (INIS)

    Talarek, H.D.

    1976-01-01

    Starting from the lump-sum criterion of distance, the criterion of MCA has developed based on the idea of limiting accident consequences with the object in mind of guaranteeing the protection of the population in case of the maximum credible accident in its absolute sense. This claim has proved to be indefensible for the utilization and design of the MCA criterion on nuclear power plants of larger capacities on sites near to conurbation centres. Within the concept of design basis accidents this is not claimed, an imperative formulation was found instead concerning the sort of accident to be considered, including possible demands on its safety design. Using probability methods has lead to a blurring of the relationship with the technological system of nuclear power plants for formulating a criterion for the advantage of risk quantification. If the probability method is to be applied in the licensing procedure concerning nuclear power plants it will be one of the tasks of future safety research to elaborate this relationship clearly. (orig.) [de

  13. Radiation safety and culture of prevention in the use of radioactive materials in industry : criteria and trends

    International Nuclear Information System (INIS)

    Truppa, Walter Adrian

    2008-01-01

    As time goes by and experience is gained, modernization and technological development show the need to implement more complex programs and procedures to ensure a high level of compliance with radiation safety, particularly in those activities in which radioactive material is used in industry. A relevant aspect of present technology is the concern to introduce mechanisms to prevent radiological accidents or incidents, to ensure early detection of failures. This includes systems that either individually or as a whole, increase the level of responsibility of the different disciplines involved, so as to avoid a situation that could lead to loss of control of the facility or part of it. The prevention of an abnormal situation, overexposure of workers or unwanted risks, should be considered in the level of vulnerability of the facility, a concept drawn from international protection systems and which is applied directly in radiation safety. Preventive management, risk communication and proposals for change or improvement along with the detection of risks and training, constitute all the factors contained within prevention policies. Dose limitation, optimization and justification, old tools used for decades, could not be replaced by other modern concepts and criteria. ALARA culture (including performance indicators) should be considered. The atmosphere at work, working under pressure as well as other factors such as quality issues, ethics of prevention, etc. align with this idea of prevention and safety, besides changes in attitude, towards risk prevention (methods, reports, intervention guides, working instructions, and any other helpful tool), are followed by preventive, as well as predictive and corrective maintenance, applied to minimize the dose absorbed by workers. A clear policy of prevention is needed as well as an appropriate level of radiation safety which should be taken into account since the very beginning of the development of a given practice. All these

  14. Radiation Safety and Culture of Prevention in the Use of Radioactive Materials in Industry. Criteria and Trends

    International Nuclear Information System (INIS)

    Truppa, W.A.

    2011-01-01

    As time goes by and experience is gained, modernization and technological development show the need to implement more complex programs and procedures to ensure a high level of compliance with radiation safety, particularly in those activities in which radioactive material is used in industry. A relevant aspect of present technology is the concern to introduce mechanisms to prevent radiological accidents or incidents, to ensure early detection of failures. This includes systems that either individually or as a whole, increase the level of responsibility of the different disciplines involved, so as to avoid a situation that could lead to loss of control of the facility or part of it. The prevention of an abnormal situation, overexposure of workers or unwanted risks, should be considered in the level of vulnerability of the facility, a concept drawn from international protection systems and which is applied directly in radiation safety. Preventive management, risk communication and proposals for change or improvement along with the detection of risks and training, constitute all the factors contained within prevention policies. Dose limitation, optimization and justification, old tools used for decades, could not be replaced by other modern concepts and criteria. ALARA culture (including performance indicators) should be considered. The atmosphere at work, working under pressure as well as other factors such as quality issues, ethics of prevention, etc. align with this idea of prevention and safety, besides changes in attitude, towards risk prevention (methods, reports, intervention guides, working instructions, and any other helpful tool), are followed by preventive, as well as predictive and corrective maintenance, applied to minimize the dose absorbed by workers. A clear policy of prevention is needed as well as an appropriate level of radiation safety which should be taken into account since the very beginning of the development of a given practice. All these

  15. JSFR design progress related to development of safety design criteria for Generation IV sodium-cooled fast reactors. (1) Overview

    International Nuclear Information System (INIS)

    Kamide, Hideki; Ando, Masato; Ito, Takaya

    2015-01-01

    JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R and D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a looptype next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500 MWe for the commercial version and 750 MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR. (author)

  16. Increasing NASA SSC Range Safety by Developing the Framework to Monitor Airspace and Enforce Restrictions

    Data.gov (United States)

    National Aeronautics and Space Administration — The NASA John C. Stennis Space Center (SSC) Office of Safety and Mission Assurance (SMA) has a safety concern associated with unauthorized aircraft entering...

  17. Enhancing swimming pool safety by the use of range-imaging cameras

    Science.gov (United States)

    Geerardyn, D.; Boulanger, S.; Kuijk, M.

    2015-05-01

    Drowning is the cause of death of 372.000 people, each year worldwide, according to the report of November 2014 of the World Health Organization.1 Currently, most swimming pools only use lifeguards to detect drowning people. In some modern swimming pools, camera-based detection systems are nowadays being integrated. However, these systems have to be mounted underwater, mostly as a replacement of the underwater lighting. In contrast, we are interested in range imaging cameras mounted on the ceiling of the swimming pool, allowing to distinguish swimmers at the surface from drowning people underwater, while keeping the large field-of-view and minimizing occlusions. However, we have to take into account that the water surface of a swimming pool is not a flat, but mostly rippled surface, and that the water is transparent for visible light, but less transparent for infrared or ultraviolet light. We investigated the use of different types of 3D cameras to detect objects underwater at different depths and with different amplitudes of surface perturbations. Specifically, we performed measurements with a commercial Time-of-Flight camera, a commercial structured-light depth camera and our own Time-of-Flight system. Our own system uses pulsed Time-of-Flight and emits light of 785 nm. The measured distances between the camera and the object are influenced through the perturbations on the water surface. Due to the timing of our Time-of-Flight camera, our system is theoretically able to minimize the influence of the reflections of a partially-reflecting surface. The combination of a post image-acquisition filter compensating for the perturbations and the use of a light source with shorter wavelengths to enlarge the depth range can improve the current commercial cameras. As a result, we can conclude that low-cost range imagers can increase swimming pool safety, by inserting a post-processing filter and the use of another light source.

  18. Solving the Problem of Multiple-Criteria Building Design Decisions with respect to the Fire Safety of Occupants: An Approach Based on Probabilistic Modelling

    Directory of Open Access Journals (Sweden)

    Egidijus Rytas Vaidogas

    2015-01-01

    Full Text Available The design of buildings may include a comparison of alternative architectural and structural solutions. They can be developed at different levels of design process. The alternative design solutions are compared and ranked by applying methods of multiple-criteria decision-making (MCDM. Each design is characterised by a number of criteria used in a MCDM problem. The paper discusses how to choose MCDM criteria expressing fire safety related to alternative designs. Probability of a successful evacuation of occupants from a building fire and difference between evacuation time and time to untenable conditions are suggested as the most important criteria related to fire safety. These two criteria are treated as uncertain quantities expressed by probability distributions. Monte Carlo simulation of fire and evacuation processes is natural means for an estimation of these distributions. The presence of uncertain criteria requires applying stochastic MCDM methods for ranking alternative designs. An application of the safety-related criteria is illustrated by an example which analyses three alternative architectural floor plans prepared for a reconstruction of a medical building. A MCDM method based on stochastic simulation is used to solve the example problem.

  19. A stress-based fracture criteria validated on mixed microstructures of ferrite and bainite over a range of stress triaxialities

    Energy Technology Data Exchange (ETDEWEB)

    Golling, Stefan, E-mail: stefan.golling@ltu.se [Luleå University of Technology, SE 971 87 Luleå (Sweden); Östlund, Rickad [Gestamp HardTech, Ektjärnsvägen 5, SE 973 45 Luleå (Sweden); Oldenburg, Mats [Luleå University of Technology, SE 971 87 Luleå (Sweden)

    2016-09-30

    Hot stamping is a sequential process for formation and heat-treatment of sheet metal components with superior mechanical properties. By applying different cooling rates, the microstructural composition and thus the material properties of steel can be designed. By controlling the cooling rate in different sections of a blank, the material properties can be tailored depending on the desired toughness. Under continuous cooling, various volume fractions of ferrite and bainite are formed depending on the rate of cooling. This paper focuses on the ductile fracture behavior of a thin sheet metal made of low-alloyed boron steel with varying amounts of ferrite and bainite. An experimental setup was applied in order to produce microstructures with different volume fractions of ferrite and bainite. In total, five different test specimen geometries, representing different stress triaxialities, were heat treated and tensile tested. Through full-field measurements, flow curves extending beyond necking and the equivalent plastic strain to fracture were determined. Experimental results were further investigated using a mean-field homogenization scheme combined with local fracture criteria. The mean-field homogenization scheme comprises the influence of microstructure composition and stress triaxiality with usable accuracy, connoting auspicious possibilities for constitutive modeling of hot-stamped components.

  20. A study on safety concept and criteria of site release of nuclear installation proposed by international organizations and adopted in decommissioning practices

    International Nuclear Information System (INIS)

    Enokido, Yuji; Miyasaka, Yasuhiko; Ishikawa, Hironori

    2008-01-01

    Regulatory systems and safety criteria of site release of nuclear installation proposed by international organizations such as IAEA and applied in decommissioning in domestic and foreign countries have been studied, in order to avail them to deliberate the relevant domestic regulation and guides. In addition, the applicability of the proposal and practices to domestic legislation have been discussed. Regarding the national safety criteria, the annual individual dose constraint is optimized between 10 μSv and 300 μSv after recommendation and/or guides of IAEA etc. Unconditional release should be achieved, but the conditional and/or partial site release are possible under the same safety criteria to make the selection flexible for licensees. (author)

  1. Nuclear Safety: Our Overriding Priority. EDF Group Report 2015 in response to FTSE4Good Nuclear Criteria

    International Nuclear Information System (INIS)

    Maillart, H.

    2015-01-01

    contractors enforce that requirement and employ fully-trained, rigorously professional staff. The Group is convinced that excellence in everything it does, backed by reliable equipment, human performance and efficient work management, is the driver of nuclear safety, which in turn enhances performance in other areas (Professional excellence is the overriding theme of EDF Nuclear Generation Division's Generation 2020). The Group recognises the importance of instilling a good nuclear safety culture in staff and contractors. The Group's companies maintain an efficient crisis system in a state of constant readiness. This is tested and improved via regular emergency drills with local and national authorities. The crisis management organisation in France has been strengthened, with the deployment of new guidelines at the end of 2012, supplemented by feedback from experience with the Fukushima accident Continuous improvement is fostered and organised using the full range of knowledge and services within the Group, enriched by international experience. Dialogue and transparency are essential to building trust through clear and timely communication on events and their impact. The EDF Group's nuclear safety policy has been discussed with the directors of nuclear power plants and engineering departments to determine how best to deploy it. The aim is to ensure that all EDF personnel and contractors working on the site understand and implement the main aims and points of this policy. The new nuclear safety policy has been incorporated into training programmes for EDF personnel and contractors

  2. Boundary conditions for pathways, safety analysis and basic criteria for low-level radiation waste site selection

    International Nuclear Information System (INIS)

    Saverot, P.

    1994-01-01

    There are three successive periods in the life of a disposal facility: the operating period, the institutional control period and the unrestricted site access period. The purpose of safety analysis of the disposal facility is to ensure that the radiological impacts for each period in the life of the facility are acceptable under all circumstances. Founded on a deterministic approach, this analysis leads to a determination of the maximum quantity of each radionuclide present in the facility at the beginning of the institutional control period in order for the impacts to be considered acceptable. Safety analysis involves the calculation of the radiological impacts of a given radiological inventory under a selected scenario, from all plausible scenarios of radionuclide migration to the environment in both normal and accident conditions, and taking into account other specified variables. The calculation itself involves an assessment of the quantities of radionuclides that could be released to the environment under the specific scenario selected and following identified pathways, and a determination of the resultant exposure, both internal and external, to the public. An iterative approach is used in the performance of pathways analyses. If the pathways analyses result in unacceptable radiological impacts, either the radiological inventory of the site is reduced or barrier characteristics not previously factored into the analysis are taken into account. New pathways analyses are then performed until the results are within the acceptable range. Once accepted by the safety authorities, the radiological inventory becomes the radiological capacity, which is the approved quantities of specific radionuclides that may be disposed of at the site. The following elaborates on the boundary conditions used in safety analyses and describes the types of pathways analyses performed for a LLW disposal facility

  3. 33 CFR 165.1406 - Safety Zone: Pacific Missile Range Facility (PMRF), Barking Sands, Island of Kauai, Hawaii.

    Science.gov (United States)

    2010-07-01

    ... Range Facility (PMRF), Barking Sands, Island of Kauai, Hawaii. 165.1406 Section 165.1406 Navigation and...), Barking Sands, Island of Kauai, Hawaii. (a) Location. The following area is established as a safety zone during launch operations at PMRF, Kauai, Hawaii: The waters bounded by the following coordinates: (22°01...

  4. 14 CFR 414.35 - Public notification of the criteria by which a safety approval was issued.

    Science.gov (United States)

    2010-01-01

    ... issued. For each grant of a safety approval, the FAA will publish in the Federal Register a notice of the... which a safety approval was issued. 414.35 Section 414.35 Aeronautics and Space COMMERCIAL SPACE TRANSPORTATION, FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION LICENSING SAFETY APPROVALS Safety...

  5. Basic design criteria for an impact test frame for safety glazing; Criterios basicos de diseno de banco de ensayos para impactos de vidrios de seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Postigo, S.; Pacios, A.; Huerta, C.

    2011-07-01

    The Spanish Building Code establishes the essential requirements of safety and habitability that buildings must satisfy. The Basic Document of Safety in Use and Accessibility identifies some critical areas where falling through brittle elements may cause a risk to the user. The document also establishes the minimum performance of glasses located in such areas, according to the impact procedure described in UNE-EN 12600:2003. However, this standard does not provide detailed information about the characteristics of the test equipment, but indicates a final calibration as validation test. The general criteria and conditions of this calibration are also incorporated in the UNE-EN 12600. To better achieve a successful manufacture of a pendulum complying with calibration limits, a proposal of the basic design criteria of a test frame for impacts of safety glazing is presented in this paper. Prototypes and results have been evaluated using dynamic design criteria of the impact phenomenon. Three criteria proposed and applied in the design and manufacture of a real test frame have helped to achieve the calibration required by the UNE-EN 12600:2003. The repeatability and reproducibility of the tests presented in this paper also guaranty the robustness of the set-up. (Author)

  6. Safety indicators in different time frames for the safety assessment of underground radioactive waste repositories. First report of the INWAC subgroup on principles and criteria for radioactive waste disposal

    International Nuclear Information System (INIS)

    1994-10-01

    Principles and criteria for the disposal of long lived radioactive waste involve issues which go beyond those normally considered in the basic system of radiation protection. Safety criteria based on radiation risk an dose limitation are commonly accepted as the principal basis for judging the acceptability of radioactive waste repositories. However, the long time-scales of interest mean that risks or doses to future individuals cannot be predicted with any certainty as they depend, amongst other things, on assumptions made about the integrity of the waste matrix, the man-made barriers, the geology, the dispersion of groundwater, etc. and future biospheric conditions and human lifestyles. This document discusses various safety indicators and their applicability in the context of the future time-scales which have to be considered in safety assessments of deep geologic repositories. Quantitative assessment are based on numerical estimates of consequences (e.g. risk or dose) and the assessment is made against numerical criteria. Qualitative assessments are based on estimates of hazard potential which are not exact or absolute and the assessment is made against criteria which may not be numerically defined. Examples of such criteria are the convenient reference values provided by levels of radionuclides in the natural environment. Refs, figs and tabs

  7. Challenges in the Acceptance/Licensing of a Mobile Ballistic Missile Range Safety Technology (BMRST) System

    National Research Council Canada - National Science Library

    Bartone, Chris

    2001-01-01

    ...), Space Vehicle Directorate, Ballistic Missile Technology program. The BMRST Program is to develop and to demonstrate a "certifiable" mobile launch range tracking and control system based upon the Global Positioning System (GPS...

  8. Toxic chemical risk acceptance criteria

    International Nuclear Information System (INIS)

    Craig, D.K.; Davis, J.; Lee, L.; Lein, P.; Omberg, S.

    1992-01-01

    This paper presents recommendations of a subcommittee of the Westinghouse M ampersand 0 Nuclear Facility Safety Committee concerning toxic chemical risk acceptance criteria. Two sets of criteria have been developed, one for use in the hazard classification of facilities, and the second for use in comparing risks in DOE non-reactor nuclear facility Safety Analysis Reports. The Emergency Response Planning Guideline (ERPG) values are intended to provide estimates of concentration ranges for specific chemicals above which exposure would be expected to lead to adverse heath effects of increasing severity for ERPG-1, -2, and -3s. The subcommittee recommends that criteria for hazard class or risk range be based on ERPGs for all chemicals. Probability-based Incremental Cancer Risk (ICR) criteria are recommended for additional analyses of risks from all known or suspected human carcinogens. Criteria are given for both on-site and off-site exposure. The subcommittee also recommends that the 5-minute peak concentration be compared with the relevant criterion with no adjustment for exposure time. Since ERPGs are available for only a limited number of chemicals, the subcommittee has developed a proposed hierarchy of concentration limit parameters for the different criteria

  9. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (4) Balance of plant

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Katoh, Atsushi; Nabeshima, Kunihiko; Ohtaka, Masahiko; Uzawa, Masayuki; Ikari, Risako; Iwasaki, Mikinori

    2015-01-01

    In this paper, design study and evaluation related with safety design criteria (SDC) and safety design guideline (SDG) on the balance of plant (BOP) of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system, requirements and relation with safety grade components such investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG. (author)

  10. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  11. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  12. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  13. Post-disposal safety assessment of toxic and radioactive waste: waste types, disposal practices, disposal criteria, assessment methods and post-disposal impacts

    International Nuclear Information System (INIS)

    Torres, C.; Simon, I.; Little, R.H.; Charles, D.; Grogan, H.A.; Smith, G.M.; Sumerling, T.J.; Watkins, B.M.

    1993-01-01

    The need for safety assessments of waste disposal stems not only from the implementation of regulations requiring the assessment of environmental effects, but also from the more general need to justify decisions on protection requirements. As waste-disposal methods have become more technologically based, through the application of more highly engineered design concepts and through more rigorous and specific limitations on the types and quantities of the waste disposed, it follows that assessment procedures also must become more sophisticated. It is the overall aim of this study to improve the predictive modelling capacity for post-disposal safety assessments of land-based disposal facilities through the development and testing of a comprehensive, yet practicable, assessment framework. This report records all the work which has been undertaken during Phase 1 of the study. Waste types, disposal practices, disposal criteria and assessment methods for both toxic and radioactive waste are reviewed with the purpose of identifying those features relevant to assessment methodology development. Difference and similarities in waste types, disposal practices, criteria and assessment methods between countries, and between toxic and radioactive wastes are highlighted and discussed. Finally, an approach to identify post-disposal impacts, how they arise and their effects on humans and the environment is described

  14. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR and GEN IV

    International Nuclear Information System (INIS)

    O'Donnell, William J.; Griffin, Donald S.

    2007-01-01

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  15. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

    Energy Technology Data Exchange (ETDEWEB)

    William J. O’Donnell; Donald S. Griffin

    2007-05-07

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  16. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (3) Progress of component design

    International Nuclear Information System (INIS)

    Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi; Eto, Masao; Miyagawa, Takayuki

    2015-01-01

    In the frame work of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors have been developing in the circumstance of worldwide deployment of SFRs. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC in the feasibility study of SDG for Sodium-cooled Fast Reactor (SFR). In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX (pump/IHX) was modified for the primary heat exchanger (PHX), which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator (SG), protective wall tube type design is under investigation as an option with less R and D risks. (author)

  17. Guide to the declaration procedure and coding system for criteria concerning significant events related to safety, radiation protection or the environment, applicable to basic nuclear installations and the transport of radioactive materials

    International Nuclear Information System (INIS)

    Lacoste, Andre-Claude

    2005-01-01

    This guide notably contains various forms associated with the declaration of significant events, and explanations to fill them in: significant event declaration form for a basic nuclear installation, significant event declaration form for radioactive material transport, significant event report for a basic nuclear installation, significant event report for radioactive material transport, declaration criteria for significant events related to the safety of non-PWR basic nuclear installations, declaration criteria for significant events related to PWR safety, significant events declared further to events resulting in group 1 unavailability and non-compliance with technical operating specifications, declaration criteria for significant events concerning radiation protection for basic nuclear installations, declaration criteria for significant events concerning environmental protection, applicable to basic nuclear installations, and declaration criteria for significant events concerning radioactive material transport

  18. Application of the MERIT survey in the multi-criteria quality assessment of occupational health and safety management.

    Science.gov (United States)

    Korban, Zygmunt

    2015-01-01

    Occupational health and safety management systems apply audit examinations as an integral element of these systems. The examinations are used to verify whether the undertaken actions are in compliance with the accepted regulations, whether they are implemented in a suitable way and whether they are effective. One of the earliest solutions of that type applied in the mining industry in Poland involved the application of audit research based on the MERIT survey (Management Evaluation Regarding Itemized Tendencies). A mathematical model applied in the survey facilitates the determination of assessment indexes WOPi for each of the assessed problem areas, which, among other things, can be used to set up problem area rankings and to determine an aggregate (synthetic) assessment. In the paper presented here, the assessment indexes WOPi were used to calculate a development measure, and the calculation process itself was supplemented with sensitivity analysis.

  19. Application range affected by software failures in safety relevant instrumentation and control systems of nuclear power plants

    International Nuclear Information System (INIS)

    Jopen, Manuela; Mbonjo, Herve; Sommer, Dagmar; Ulrich, Birte

    2017-03-01

    This report presents results that have been developed within a BMUB-funded research project (Promotion Code 3614R01304). The overall objective of this project was to broaden the knowledge base of GRS regarding software failures and their impact in software-based instrumentation and control (I and C) systems. To this end, relevant definitions and terms in standards and publications (DIN, IEEE standards, IAEA standards, NUREG publications) as well as in the German safety requirements for nuclear power plants were analyzed first. In particular, it was found that the term ''software fault'' is defined differently and partly contradictory in the considered literature sources. For this reason, a definition of software fault was developed on the basis of the software life cycle of software-based I and C systems within the framework of this project, which takes into account the various aspects relevant to software faults and their related effects. It turns out that software failures result from latent faults in a software-based control system, which can lead to a non-compliant behavior of a software-based I and C system. Hereby a distinction should be made between programming faults and specification faults. In a further step, operational experience with software failures in software-based I and C systems in nuclear facilities and in nonnuclear sector was investigated. The identified events were analyzed with regard to their cause and impacts and the analysis results were summarized. Based on the developed definition of software failure and on the COMPSIS-classification scheme for events related to software based I and C systems, the COCS-classification scheme was developed to classify events from operating experience with software failures, in which the events are classified according to the criteria ''cause'', ''affected system'', ''impact'' and ''CCF potential''. This classification scheme was applied to evaluate the events identified in the framework of this project

  20. Efficacy and Safety of OnabotulinumtoxinA Treatment of Forehead Lines: A Multicenter, Randomized, Dose-Ranging Controlled Trial.

    Science.gov (United States)

    Solish, Nowell; Rivers, Jason K; Humphrey, Shannon; Muhn, Channy; Somogyi, Chris; Lei, Xiaofang; Bhogal, Meetu; Caulkins, Carrie

    2016-03-01

    Various onabotulinumtoxinA doses are effective in treating forehead lines (FHL), with a trend toward lower doses. To evaluate efficacy and safety of onabotulinumtoxinA dose-ranging treatment of FHL when the frontalis area and glabellar complex are treated together. Adults with moderate-to-severe FHL received onabotulinumtoxinA 40 U (FHL, 20 U; glabellar lines [GL], 20 U), 30 U (FHL, 10 U; GL, 20 U), or placebo. Response was assessed at weeks 1, 2, day 30, and monthly to day 180. Coprimary efficacy end points were investigator- and subject-assessed Facial Wrinkle Scale scores of none or mild (day 30). Patient-reported outcomes, onset/duration of effect, and adverse events (AEs) were evaluated. Responder rates (investigator/subject, respectively) were 40-U group, 91.2%/89.5%; 30-U group, 86.4%/81.4%; placebo, 1.7%/5.1%. OnabotulinumtoxinA resulted in significantly greater responder rates than placebo (p < .001). Adverse events were mild to moderate and similar between groups (most common AEs: nasopharyngitis [4.6%] and headache [4.0%]). Treatment of FHL with onabotulinumtoxinA 40 and 30 U (in frontalis and glabellar complex muscles) was tolerable, effective, and sustained. Both doses significantly reduced FHL severity; however, the 40-U dose demonstrated a trend toward greater sustained benefit and longer duration of effect versus the 30-U dose, with similar AE rates.

  1. Multimegawatt Space Reactor Safety

    International Nuclear Information System (INIS)

    Stanley, M.L.

    1989-01-01

    The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

  2. Key Performance Criteria Affecting the Most the Safety of a Nuclear Waste Long Term Storage : A Case Study Commissioned by CEA

    International Nuclear Information System (INIS)

    Marvy, A.; Lioure, A; Heriard-Dubreuil, G.; Gadbois, S.; Schneider, T.; Schieber, C.

    2003-01-01

    As part of the work scope set in the French law on high level long lived waste R and D passed in 1991, CEA is conducting a research program to establish the scientific basis and assess the feasibility of long term storage as an option for the safe management of nuclear waste for periods as long as centuries. This goal is a significant departure from the current industrial practice where storage facilities are usually built to last only a few decades. From a technical viewpoint such an extension in time seems feasible provided care and maintenance is exercised. Considering such long periods of time, the risk for Society of loosing oversight and control of such a facility is real, which triggers the question of whether and how long term storage safety can be actually achieved. Therefore CEA commissioned a study (1) in which MUTADIS Consultants (2) and CEPN (3) were both involved. The case study looks into several past and actual human enterprises conducted over significant periods o f time, one of them dating back to the end of the 18th century, and all identified out of the nuclear field. Then-prevailing societal behavior and organizational structures are screened out to show how they were or are still able to cope with similar oversight and control goals. As a result, the study group formulated a set of performance criteria relating to issues like responsibility, securing funds, legal and legislative implications, economic sustainable development, all being areas which are not traditionally considered as far as technical studies are concerned. These criteria can be most useful from the design stage onward, first in an attempt to define the facility construction and operating guiding principles, and thereafter to substantiate the safety case for long term storage and get geared to the public dialogue on that undertaking should it become a reality

  3. Key Performance Criteria Affecting the Most the Safety of a Nuclear Waste Long Term Storage : A Case Study Commissioned by CEA

    Energy Technology Data Exchange (ETDEWEB)

    Marvy, A.; Lioure, A; Heriard-Dubreuil, G.; Gadbois, S.; Schneider, T.; Schieber, C.

    2003-02-24

    As part of the work scope set in the French law on high level long lived waste R&D passed in 1991, CEA is conducting a research program to establish the scientific basis and assess the feasibility of long term storage as an option for the safe management of nuclear waste for periods as long as centuries. This goal is a significant departure from the current industrial practice where storage facilities are usually built to last only a few decades. From a technical viewpoint such an extension in time seems feasible provided care and maintenance is exercised. Considering such long periods of time, the risk for Society of loosing oversight and control of such a facility is real, which triggers the question of whether and how long term storage safety can be actually achieved. Therefore CEA commissioned a study (1) in which MUTADIS Consultants (2) and CEPN (3) were both involved. The case study looks into several past and actual human enterprises conducted over significant periods o f time, one of them dating back to the end of the 18th century, and all identified out of the nuclear field. Then-prevailing societal behavior and organizational structures are screened out to show how they were or are still able to cope with similar oversight and control goals. As a result, the study group formulated a set of performance criteria relating to issues like responsibility, securing funds, legal and legislative implications, economic sustainable development, all being areas which are not traditionally considered as far as technical studies are concerned. These criteria can be most useful from the design stage onward, first in an attempt to define the facility construction and operating guiding principles, and thereafter to substantiate the safety case for long term storage and get geared to the public dialogue on that undertaking should it become a reality.

  4. A look at new key performance criteria that could most affect the safety of long term storage of nuclear waste. A case study commissioned by CEA

    International Nuclear Information System (INIS)

    Marvy, A.; Lioure, A.; Heriard-Dubreuil, G.; Gadbois, S.; Schneider, T.; Schieber, C

    2002-01-01

    As part of the work scope set in the French law on high level long lived waste R and D passed in 1991, CEA is conducting research work to establish the scientific basis and assess the feasibility of long term storage as an option for the safe management of nuclear waste for periods as long as a few centuries. This goal is a significant departure from current industrial practice where storage facilities are usually built to last only a few decades. From a technical viewpoint such an extension in time seems feasible provided care and maintenance is exercised. Considering such long periods of time risk for Society of loosing oversight and control of such a facility is real which triggers the question of whether and how long term storage safety can be actually achieved. Therefore CEA commissioned a study in which MUTADIS Consultants and CEPN were both involved. The case study looks into several past and actual human enterprises conducted over significant periods of time - one dating back to the end of the 18th century - and identified off the nuclear field. Then-prevailing societal behaviour and organizational structures are screened out to show how they were and are still able to cope with similar oversight and control goals. As a result the study group obtained a set of performance criteria relating to issues like responsibility, securing funds, legal and legislative implications, economic sustainable development, all being areas which are not traditionally considered when technical studies are conducted. These criteria can be most useful from the design stage onward, first in an attempt to define the facility construction and operating guiding principles, and thereafter to substantiate the safety case for long term storage and get geared to the public dialogue on that undertaking should it become a reality. (author)

  5. Recommendations from the workshop on Comparative Approaches to Safety Assessment of GM Plant Materials: A road toward harmonized criteria?

    Science.gov (United States)

    Bartholomaeus, Andrew; Batista, Juan Carlos; Burachik, Moisés; Parrott, Wayne

    2015-01-01

    An international meeting of genetically modified (GM) food safety assessors from the main importing and exporting countries from Asia and the Americas was held in Buenos Aires, Argentina, between June 26(th) and 28(th), 2013. Participants shared their evaluation approaches, identified similarities and challenges, and used their experience to propose areas for future work. Recommendations for improving risk assessment procedures and avenues for future collaboration were also discussed. The deliberations of the meeting were also supported by a survey of participants which canvassed risk assessment approaches across the regions from which participants came. This project was initiated by Argentine Agri-Food Health and Quality National Service (SENASA, Ministry of Agriculture, Argentina), with the support of the International Life Sciences Institute (ILSI) and other partner institutions. The importance of making all possible efforts toward more integrated and harmonized regulatory oversight for GM organisms (GMOs) was strongly emphasized. This exercise showed that such harmonization is a feasible goal that would contribute to sustain a fluid trade of commodities and ultimately enhance food security. Before this can be achieved, key issues identified in this meeting will have to be addressed in the near future to enable regulatory collaboration or joint work. The authors propose that the recommendations coming out of the meeting should be used as a basis for continuing work, follow up discussions and concrete actions.

  6. New decision criteria for selecting delta check methods based on the ratio of the delta difference to the width of the reference range can be generally applicable for each clinical chemistry test item.

    Science.gov (United States)

    Park, Sang Hyuk; Kim, So-Young; Lee, Woochang; Chun, Sail; Min, Won-Ki

    2012-09-01

    Many laboratories use 4 delta check methods: delta difference, delta percent change, rate difference, and rate percent change. However, guidelines regarding decision criteria for selecting delta check methods have not yet been provided. We present new decision criteria for selecting delta check methods for each clinical chemistry test item. We collected 811,920 and 669,750 paired (present and previous) test results for 27 clinical chemistry test items from inpatients and outpatients, respectively. We devised new decision criteria for the selection of delta check methods based on the ratio of the delta difference to the width of the reference range (DD/RR). Delta check methods based on these criteria were compared with those based on the CV% of the absolute delta difference (ADD) as well as those reported in 2 previous studies. The delta check methods suggested by new decision criteria based on the DD/RR ratio corresponded well with those based on the CV% of the ADD except for only 2 items each in inpatients and outpatients. Delta check methods based on the DD/RR ratio also corresponded with those suggested in the 2 previous studies, except for 1 and 7 items in inpatients and outpatients, respectively. The DD/RR method appears to yield more feasible and intuitive selection criteria and can easily explain changes in the results by reflecting both the biological variation of the test item and the clinical characteristics of patients in each laboratory. We suggest this as a measure to determine delta check methods.

  7. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  8. Safety

    International Nuclear Information System (INIS)

    2001-01-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  9. Analysis of static safety of power systems: a study about contingencies selection criteria in the reactive subproblem; Analise de seguranca estatica de sistemas de potencia: um estudo sobre criterios de selecao de contingencias no subproblema reativo

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Jose Vicente Canto dos

    1993-12-01

    The main objective of static safety's analysis in power systems is the determination of the level of gravity of the different contingencies that can occur in a system. Habitually, static safety's analysis is divided in two parts: selection and analysis of contingencies. In this work, they are studied several criteria of selection of applicable contingencies to the sub-problem reactive and are introduced comparisons among results provided by different criteria. They are also studied several forms of evaluation of the impact caused by contingencies on the power systems reactive profile.

  10. Analysis of static safety of power systems: a study about contingencies selection criteria in the reactive subproblem; Analise de seguranca estatica de sistemas de potencia: um estudo sobre criterios de selecao de contingencias no subproblema reativo

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Jose Vicente Canto dos

    1993-12-01

    The main objective of static safety's analysis in power systems is the determination of the level of gravity of the different contingencies that can occur in a system. Habitually, static safety's analysis is divided in two parts: selection and analysis of contingencies. In this work, they are studied several criteria of selection of applicable contingencies to the sub-problem reactive and are introduced comparisons among results provided by different criteria. They are also studied several forms of evaluation of the impact caused by contingencies on the power systems reactive profile.

  11. Health and safety plan for the removal action at the former YS-860 Firing Ranges, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-24

    This health and safety plan sets forth the requirements and procedures to protect the personnel involved in the removal action project at the former YS-860 Firing Ranges. This project will be conducted in a manner that ensures the protection of the safety and health of workers, the public, and the environment. The purpose of this removal action is to address lead-contaminated soil and reduce a potential risk to human health and the environment. This site is an operable unit within the Upper East Fork Poplar Creek watershed. The removal action will contribute to early source actions within the watershed. The project will accomplish this through the removal of lead-contaminated soil in the target areas of the two small arms firing ranges. The primary hazards include temperature extremes, equipment operation, noise, potential lead exposure, uneven and slippery working surfaces, and insects.

  12. Health and safety plan for the removal action at the former YS-860 Firing Ranges, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1998-01-01

    This health and safety plan sets forth the requirements and procedures to protect the personnel involved in the removal action project at the former YS-860 Firing Ranges. This project will be conducted in a manner that ensures the protection of the safety and health of workers, the public, and the environment. The purpose of this removal action is to address lead-contaminated soil and reduce a potential risk to human health and the environment. This site is an operable unit within the Upper East Fork Poplar Creek watershed. The removal action will contribute to early source actions within the watershed. The project will accomplish this through the removal of lead-contaminated soil in the target areas of the two small arms firing ranges. The primary hazards include temperature extremes, equipment operation, noise, potential lead exposure, uneven and slippery working surfaces, and insects

  13. Plutonium storage criteria

    Energy Technology Data Exchange (ETDEWEB)

    Chung, D. [Scientech, Inc., Germantown, MD (United States); Ascanio, X. [Dept. of Energy, Germantown, MD (United States)

    1996-05-01

    The Department of Energy has issued a technical standard for long-term (>50 years) storage and will soon issue a criteria document for interim (<20 years) storage of plutonium materials. The long-term technical standard, {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides,{close_quotes} addresses the requirements for storing metals and oxides with greater than 50 wt % plutonium. It calls for a standardized package that meets both off-site transportation requirements, as well as remote handling requirements from future storage facilities. The interim criteria document, {open_quotes}Criteria for Interim Safe Storage of Plutonium-Bearing Solid Materials{close_quotes}, addresses requirements for storing materials with less than 50 wt% plutonium. The interim criteria document assumes the materials will be stored on existing sites, and existing facilities and equipment will be used for repackaging to improve the margin of safety.

  14. Repository operational criteria analysis

    International Nuclear Information System (INIS)

    Hageman, J.P.; Chowdhury, A.H.

    1992-08-01

    The objective of the ''Repository Operational Criteria (ROC) Feasibility Studies'' (or ROC task) was to conduct comprehensive and integrated analyses of repository design, construction, and operations criteria in 10 CFR Part 60 regulations, considering the interfaces and impacts of any potential changes to those regulations. The study addresses regulatory criteria related to the preclosure aspects of the geologic repository. The study task developed regulatory concepts or potential repository operational criteria (PROC) based on analysis of a repository's safety functions and other regulations for similar facilities. These regulatory concepts or PROC were used as a basis to assess the sufficiency and adequacy of the current criteria in 10 CFR Part 60. Where the regulatory concepts were same as current operational criteria, these criteria were referenced. The operations criteria referenced or the PROC developed are given in this report. Detailed analyses used to develop the regulatory concepts and any necessary PROC for those regulations that may require a minor change are also presented. The results of the ROC task showed a need for further analysis and possible major rule change related to the design bases of a geologic repository operations area, siting, and radiological emergency planning

  15. Design criteria and pressure vessel codes - an American view

    International Nuclear Information System (INIS)

    Tuppeny, W.H.

    1975-01-01

    To the pressure vessel designer, codes and criteria represent the common ground where the stress analyst and the metallurgist must interact and evolve rules and procedures which will ensure safety and open-ended responsiveness to technological, economic, and environmental change. The paper briefly discusses the evolution and rationale behind the current ASME code sections -emphasizing those portions applicable to designs operating in the creep range. The author then proposes a plan of action so that the analysts and materials people can make optimum use of time and resources, and evolve data and design criteria which will be responsive to changing technology and the economic and safety requirements of the future. (author)

  16. Rating of environmental criteria

    Energy Technology Data Exchange (ETDEWEB)

    Glueck, K; Krasser, G

    1980-01-01

    After a general theoretical discussion on the question of rating within a framework of cost-benefit studies, first trials as to the quantification and standardisation of twelve selected environmental criteria by means of an indicator system are worked out and compiled. The selection includes the criteria exhaust gas, dust, micro climate, water pollution, water regime, land requirement, vibrations, traffic noise, landscape scene, urban scene, effect of separation and safety risks. An insight is given of the rating practice using an evaluation of the available literature, of a household interview and of an interview of experts. The interviewing of 156 experts as to their rating conception of ten criteria in the second round has provided contributions to the general problem of the evaluation estimate based on multi criteria analysis as well as differentiation of the twelve or ten environmental criteria. The following criteria ratings given by the experts and which are averaged and smoothed are: traffic noise 20,0% +- 8,5; air pollution 15,0% +- 7,0; safety risk 13,0% +- 7,0; soil and water pollution 8,5% +- 5,0; landscape scene 8,0% +- 4,5; urban scene 8,0% +- 4,5; water regime 6,5% +- 3,5 and vibrations 4,5% +- 2,5.

  17. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  18. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  19. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  20. Health and safety plan for the removal action at the former YS-860 Firing Ranges, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-28

    This health and safety plan sets forth the requirements and procedures to protect the personnel involved in the Lead Source Removal Project at the Former YS-86O Firing Ranges. This project will be conducted in a manner that ensures the protection of the safety and health of workers, the public, and the environment. The purpose of this removal action is to address lead contaminated soil and reduce a potential risk to human health and the environment. This site is an operable unit within the Upper East Fork Poplar Creek watershed. The removal action will contribute to early source actions within the watershed. The project will accomplish this through the removal of lead-contaminated soil in the target areas of the two small arms firing ranges. This plan covers the removal actions at the Former YS-86O Firing Ranges. These actions involve the excavation of lead-contaminated soils, the removal of the concrete trench and macadam (asphalt) paths, verification/confirmation sampling, grading and revegetation. The primary hazards include temperature extremes, equipment operation, noise, potential lead exposure, uneven and slippery working surfaces, and insects.

  1. Health and safety plan for the removal action at the former YS-860 Firing Ranges, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1998-01-01

    This health and safety plan sets forth the requirements and procedures to protect the personnel involved in the Lead Source Removal Project at the Former YS-86O Firing Ranges. This project will be conducted in a manner that ensures the protection of the safety and health of workers, the public, and the environment. The purpose of this removal action is to address lead contaminated soil and reduce a potential risk to human health and the environment. This site is an operable unit within the Upper East Fork Poplar Creek watershed. The removal action will contribute to early source actions within the watershed. The project will accomplish this through the removal of lead-contaminated soil in the target areas of the two small arms firing ranges. This plan covers the removal actions at the Former YS-86O Firing Ranges. These actions involve the excavation of lead-contaminated soils, the removal of the concrete trench and macadam (asphalt) paths, verification/confirmation sampling, grading and revegetation. The primary hazards include temperature extremes, equipment operation, noise, potential lead exposure, uneven and slippery working surfaces, and insects

  2. 49 CFR Appendix to Subpart H of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Non-North America-Domiciled...

    Science.gov (United States)

    2010-10-01

    ... safety audit will include: (1) Verification of available performance data and safety management programs; (2) Verification of a controlled substances and alcohol testing program consistent with part 40 of... Regulations, parts 382 through 399 of this subchapter, and the Federal Hazardous Material Regulations, parts...

  3. Criteria CSR

    OpenAIRE

    Vovk, V.; Zateyshikova, O.

    2014-01-01

    In the article the theoretical aspects regarding criteria for assessing CSR proposed by A. Carroll, including: economic, legal, ethical, philanthropic. Based on this, it is proposed to characterize these criteria with respect to the interested parties (stakeholders), including: investors, shareholders suppliers, customers, employees, society and the state. This will make a qualitative assessment of the presence and depth using social responsibility in the company, as well as determine the ext...

  4. Rationales for the Lightning Launch Commit Criteria

    Science.gov (United States)

    Willett, John C. (Editor); Merceret, Francis J. (Editor); Krider, E. Philip; O'Brien, T. Paul; Dye, James E.; Walterscheid, Richard L.; Stolzenburg, Maribeth; Cummins, Kenneth; Christian, Hugh J.; Madura, John T.

    2016-01-01

    Since natural and triggered lightning are demonstrated hazards to launch vehicles, payloads, and spacecraft, NASA and the Department of Defense (DoD) follow the Lightning Launch Commit Criteria (LLCC) for launches from Federal Ranges. The LLCC were developed to prevent future instances of a rocket intercepting natural lightning or triggering a lightning flash during launch from a Federal Range. NASA and DoD utilize the Lightning Advisory Panel (LAP) to establish and develop robust rationale from which the criteria originate. The rationale document also contains appendices that provide additional scientific background, including detailed descriptions of the theory and observations behind the rationales. The LLCC in whole or part are used across the globe due to the rigor of the documented criteria and associated rationale. The Federal Aviation Administration (FAA) adopted the LLCC in 2006 for commercial space transportation and the criteria were codified in the FAA's Code of Federal Regulations (CFR) for Safety of an Expendable Launch Vehicle (Appendix G to 14 CFR Part 417, (G417)) and renamed Lightning Flight Commit Criteria in G417.

  5. Principles and Criteria for Design

    DEFF Research Database (Denmark)

    Beghin, D.; Cervetto, D.; Hansen, Peter Friis

    1997-01-01

    The mandate of ISSC Committee IV.1 on principles and Criteria for Design is to report on the following:The ongoing concern for quantification of general economic and safety criteria for marine structures and for the development of appropriate principles for rational life cycle design using...

  6. Criteria for controlled atmosphere chambers

    International Nuclear Information System (INIS)

    Robinson, J.N.

    1980-03-01

    The criteria for design, construction, and operation of controlled atmosphere chambers intended for service at ORNL are presented. Classification of chambers, materials for construction, design criteria, design, controlled atmosphere chamber systems, and operating procedures are presented. ORNL Safety Manual Procedure 2.1; ORNL Health Physics Procedure Manual Appendix A-7; and Design of Viewing Windows are included in 3 appendices

  7. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  8. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    International Nuclear Information System (INIS)

    1980-01-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases

  9. FHR Generic Design Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, George F [ORNL; Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL

    2012-06-01

    The purpose of this document is to provide an initial, focused reference to the safety characteristics of and a licensing approach for Fluoride-Salt-Cooled High-Temperature Reactors (FHRs). The document does not contain details of particular reactor designs nor does it attempt to identify or classify either design basis or beyond design basis accidents. Further, this document is an initial attempt by a small set of subject matter experts to document the safety and licensing characteristics of FHRs for a larger audience. The document is intended to help in setting the safety and licensing research, development, and demonstration path forward. Input from a wider audience, further technical developments, and additional study will be required to develop a consensus position on the safety and licensing characteristics of FHRs. This document begins with a brief overview of the attributes of FHRs and then a general description of their anticipated safety performance. Following this, an overview of the US nuclear power plant approval process is provided that includes both test and power reactors, as well as the role of safety standards in the approval process. The document next describes a General Design Criteria (GDC) - based approach to licensing an FHR and provides an initial draft set of FHR GDCs. The document concludes with a description of a path forward toward developing an FHR safety standard that can support both a test and power reactor licensing process.

  10. 'Nuclear safety: our absolute priority'. File 2014 of the EDF Group in response to the FTSE4Good criteria

    International Nuclear Information System (INIS)

    Maillart, H.

    2014-01-01

    After a brief presentation of the EDF Group activity, this report presents its different nuclear assets which belong to EDF SA, EDF Energy, CENG or TNPJVC (these companies are also briefly presented). The next part addresses and describes the various aspects related to safety and to radiation protection: the safety policy and its implementation within the group, incidents and events, unplanned outages, assessments of nuclear safety, risk analysis, preparation to emergency situations, adaptation to climate change, lessons learned from the Fukushima accident, public and workers exposure to radiations, site protection. The fourth part addresses issues related to wastes: general presentation, legal and regulatory context, policy, management of radioactive wastes, management of used fuels, dismantling and wastes. The next part addresses issues related to training: overview of human resources and training policy, implementation. The last part presents reporting actions and results

  11. Efficacy and safety of belimumab in patients with rheumatoid arthritis: a phase II, randomized, double-blind, placebo-controlled, dose-ranging Study.

    Science.gov (United States)

    Stohl, William; Merrill, Joan T; McKay, James D; Lisse, Jeffrey R; Zhong, Z John; Freimuth, William W; Genovese, Mark C

    2013-05-01

    To evaluate the efficacy/safety of belimumab in patients with rheumatoid arthritis (RA). Patients fulfilling American College of Rheumatology (ACR) criteria for RA for ≥ 1 year who had at least moderate disease activity while receiving stable disease-modifying antirheumatic drug (DMARD) therapy and failed ≥ 1 DMARD were randomly assigned to placebo or belimumab 1, 4, or 10 mg/kg, administered intravenously on Days 1, 14, and 28, and then every 4 weeks for 24 weeks (n = 283). This was followed by an optional 24-week extension (n = 237) in which all patients received belimumab. Primary efficacy endpoint was the Week 24 ACR20 response. Week 24 ACR20 responses with placebo and belimumab 1, 4, and 10 mg/kg were 15.9%, 34.7% (p = 0.010), 25.4% (p = 0.168), and 28.2% (p = 0.080), respectively. Patients taking any belimumab dose who continued with belimumab in the open-label extension had an ACR20 response of 41% at 48 weeks. A similar ACR20 response (42%) at 48 weeks was seen in patients taking placebo who switched in the extension to belimumab 10 mg/kg. Greater response rates were observed in patients who at baseline were rheumatoid factor-positive, anticitrullinated protein antibody-positive, or tumor necrosis factor inhibitor-naive, or had elevated C-reactive protein levels, Disease Activity Score 28 > 5.1, or low B lymphocyte stimulator levels (< 0.858 ng/ml). Adverse event rates were similar across treatment groups. In this phase II trial, belimumab demonstrated efficacy and was generally well tolerated in patients with RA who had failed previous therapies. [ClinicalTrials.gov identifier NCT00071812].

  12. 49 CFR Appendix A to Subpart E of... - Explanation of Pre-Authorization Safety Audit Evaluation Criteria for Mexico-Domiciled Motor...

    Science.gov (United States)

    2010-10-01

    ...) Verification of available performance data and safety management programs; (2) Verification of a controlled substances and alcohol testing program consistent with part 40 of this title; (3) Verification of the carrier... subchapter, and the Federal Hazardous Material Regulations, parts 171 through 180 of this title; (6...

  13. Determination of performance criteria for high-level solidified nuclear waste from the commercial nuclear fuel cycle: a probabilistic safety analysis

    International Nuclear Information System (INIS)

    Heckman, R.A.

    1978-01-01

    To minimize the radiological risk from the operation of a waste management system for the safe disposal of high-level waste, performance characteristics of the solidified waste form must be specified. The minimum waste form characteristics that must be specified are the radionuclide volatilization fraction, airborne particulate dispersion fraction, and the aqueous dissolution characteristics. The results indicate that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. The actual values of expected risk are sensitive to modeling assumptions and data base uncertainties. The transportation step appears to be the most limiting in determining the required performance characteristics

  14. An open-label multicenter study to assess the safety of dextromethorphan/quinidine in patients with pseudobulbar affect associated with a range of underlying neurological conditions.

    Science.gov (United States)

    Pattee, Gary L; Wymer, James P; Lomen-Hoerth, Catherine; Appel, Stanley H; Formella, Andrea E; Pope, Laura E

    2014-11-01

    Pseudobulbar affect (PBA) is associated with neurological disorders or injury affecting the brain, and characterized by frequent, uncontrollable episodes of crying and/or laughing that are exaggerated or unrelated to the patient's emotional state. Clinical trials establishing dextromethorphan and quinidine (DM/Q) as PBA treatment were conducted in patients with amyotrophic lateral sclerosis (ALS) or multiple sclerosis (MS). This trial evaluated DM/Q safety in patients with PBA secondary to any neurological condition affecting the brain. To evaluate the safety and tolerability of DM/Q during long-term administration to patients with PBA associated with multiple neurological conditions. Fifty-two-week open-label study of DM/Q 30/30 mg twice daily. Safety measures included adverse events (AEs), laboratory tests, electrocardiograms (ECGs), vital signs, and physical examinations. #NCT00056524. A total of 553 PBA patients with >30 different neurological conditions enrolled; 296 (53.5%) completed. The most frequently reported treatment-related AEs (TRAEs) were nausea (11.8%), dizziness (10.5%), headache (9.9%), somnolence (7.2%), fatigue (7.1%), diarrhea (6.5%), and dry mouth (5.1%). TRAEs were mostly mild/moderate, generally transient, and consistent with previous controlled trials. Serious AEs (SAEs) were reported in 126 patients (22.8%), including 47 deaths, mostly due to ALS progression and respiratory failure. No SAEs were deemed related to DM/Q treatment by investigators. ECG results suggested no clinically meaningful effect of DM/Q on myocardial repolarization. Differences in AEs across neurological disease groups appeared consistent with the known morbidity of the primary neurological conditions. Study interpretation is limited by the small size of some disease groups, the lack of a specific efficacy measure and the use of a DM/Q dose higher than the eventually approved dose. DM/Q was generally well tolerated over this 52 week trial in patients with PBA

  15. Safety of dual kidney transplantation compared to single kidney transplantation from expanded criteria donors: a single center cohort study of 39 recipients.

    Science.gov (United States)

    Mendel, Lionel; Albano, Laetitia; Bentellis, Imad; Yandza, Thierry; Bernardi, Caroline; Quintens, Herve; Tibi, Brannwel; Jourdan, Jacques; Durand, Matthieu; Amiel, Jean; Chevallier, Daniel

    2018-05-17

    Our objective was to compare the outcomes of dual kidney transplanataion (DKT) to single kidney transplantation (SKT) performed with grafts from expanded criteria donors (ECD) in recipients ≥65 years, focusing on surgical complications. All kidney transplantations (KT) performed between 2006 and 2014 in our institution were analysed. DKT was indicated according to the criteria of the French national Agence de la Biomedecine. Thirty-nine DKT and 155 SKT were included, with a median follow-up of 36 and 26.5 months, respectively. The rate of early surgical revisions was not significantly higher after DKT (23.1% vs 15.5% (P = 0.2593)) but more venous graft thromboses (12.8% vs 3.2% (P = 0.02)) were reported. The glomerular filtration rate (GFR) 24 months after KT was significantly higher after DKT (45.0 ± 16.3 vs 39.8 ± 13.8 ml/min/1.73m 2 ; P = 0.04) and allowed shorter waiting time without a significant increased risk of surgical revision, excepted for venous graft thrombosis, more frequent after DKT. Graft survivals were not significantly different and GFR was higher after DKT. DKT seems to remain an appropriate strategy to address the growing graft shortage in elderly patients. © 2018 Steunstichting ESOT.

  16. Application range affected by software failures in safety relevant instrumentation and control systems of nuclear power plants; Auswirkungsbereiche von Softwarefehlern in sicherheitstechnisch wichtigen Einrichtungen von Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Jopen, Manuela; Mbonjo, Herve; Sommer, Dagmar; Ulrich, Birte

    2017-03-15

    This report presents results that have been developed within a BMUB-funded research project (Promotion Code 3614R01304). The overall objective of this project was to broaden the knowledge base of GRS regarding software failures and their impact in software-based instrumentation and control (I and C) systems. To this end, relevant definitions and terms in standards and publications (DIN, IEEE standards, IAEA standards, NUREG publications) as well as in the German safety requirements for nuclear power plants were analyzed first. In particular, it was found that the term ''software fault'' is defined differently and partly contradictory in the considered literature sources. For this reason, a definition of software fault was developed on the basis of the software life cycle of software-based I and C systems within the framework of this project, which takes into account the various aspects relevant to software faults and their related effects. It turns out that software failures result from latent faults in a software-based control system, which can lead to a non-compliant behavior of a software-based I and C system. Hereby a distinction should be made between programming faults and specification faults. In a further step, operational experience with software failures in software-based I and C systems in nuclear facilities and in nonnuclear sector was investigated. The identified events were analyzed with regard to their cause and impacts and the analysis results were summarized. Based on the developed definition of software failure and on the COMPSIS-classification scheme for events related to software based I and C systems, the COCS-classification scheme was developed to classify events from operating experience with software failures, in which the events are classified according to the criteria ''cause'', ''affected system'', ''impact'' and ''CCF potential''. This

  17. Safety, immunogenicity and dose ranging of a new Vi-CRM₁₉₇ conjugate vaccine against typhoid fever: randomized clinical testing in healthy adults.

    Science.gov (United States)

    van Damme, Pierre; Kafeja, Froukje; Anemona, Alessandra; Basile, Venere; Hilbert, Anne Katrin; De Coster, Ilse; Rondini, Simona; Micoli, Francesca; Qasim Khan, Rana M; Marchetti, Elisa; Di Cioccio, Vito; Saul, Allan; Martin, Laura B; Podda, Audino

    2011-01-01

    Typhoid fever causes more than 21 million cases of disease and 200,000 deaths yearly worldwide, with more than 90% of the disease burden being reported from Asia. Epidemiological data show high disease incidence in young children and suggest that immunization programs should target children below two years of age: this is not possible with available vaccines. The Novartis Vaccines Institute for Global Health developed a conjugate vaccine (Vi-CRM₁₉₇) for infant vaccination concomitantly with EPI vaccines, either starting at 6 weeks with DTP or at 9 months with measles vaccine. We report the results from a Phase 1 and a Phase 2 dose ranging trial with Vi-CRM₁₉₇ in European adults. Following randomized blinded comparison of single vaccination with either Vi-CRM₁₉₇ or licensed polysaccharide vaccines (both containing 25·0 µg of Vi antigen), a randomised observer blinded dose ranging trial was performed in the same center to compare three concentrations of Vi-CRM₁₉₇ (1·25 µg, 5·0 µg and 12·5 µg of Vi antigen) with the polysaccharide vaccine. All vaccines were well tolerated. Compared to the polysaccharide vaccine, Vi-CRM₁₉₇ induced a higher incidence of mild to moderate short lasting local pain. All Vi-CRM₁₉₇ formulations induced higher Vi antibody levels compared to licensed control, with clear dose response relationship. Vi-CRM₁₉₇ did not elicit safety concerns, was highly immunogenic and is therefore suitable for further clinical testing in endemic populations of South Asia. ClinicalTrials.gov NCT01123941 NCT01193907.

  18. Safety, immunogenicity and dose ranging of a new Vi-CRM₁₉₇ conjugate vaccine against typhoid fever: randomized clinical testing in healthy adults.

    Directory of Open Access Journals (Sweden)

    Pierre van Damme

    Full Text Available Typhoid fever causes more than 21 million cases of disease and 200,000 deaths yearly worldwide, with more than 90% of the disease burden being reported from Asia. Epidemiological data show high disease incidence in young children and suggest that immunization programs should target children below two years of age: this is not possible with available vaccines. The Novartis Vaccines Institute for Global Health developed a conjugate vaccine (Vi-CRM₁₉₇ for infant vaccination concomitantly with EPI vaccines, either starting at 6 weeks with DTP or at 9 months with measles vaccine. We report the results from a Phase 1 and a Phase 2 dose ranging trial with Vi-CRM₁₉₇ in European adults.Following randomized blinded comparison of single vaccination with either Vi-CRM₁₉₇ or licensed polysaccharide vaccines (both containing 25·0 µg of Vi antigen, a randomised observer blinded dose ranging trial was performed in the same center to compare three concentrations of Vi-CRM₁₉₇ (1·25 µg, 5·0 µg and 12·5 µg of Vi antigen with the polysaccharide vaccine.All vaccines were well tolerated. Compared to the polysaccharide vaccine, Vi-CRM₁₉₇ induced a higher incidence of mild to moderate short lasting local pain. All Vi-CRM₁₉₇ formulations induced higher Vi antibody levels compared to licensed control, with clear dose response relationship.Vi-CRM₁₉₇ did not elicit safety concerns, was highly immunogenic and is therefore suitable for further clinical testing in endemic populations of South Asia.ClinicalTrials.gov NCT01123941 NCT01193907.

  19. Plutonium, americium, and uranium in blow-sand mounds of safety-shot sites at the Nevada Test Site and the Tonopah Test Range

    International Nuclear Information System (INIS)

    Essington, E.H.; Gilbert, R.O.; Wireman, D.L.; Brady, D.N.; Fowler, E.B.

    1977-01-01

    Blow-sand mounds or miniature sand dunes and mounds created by burrowing activities of animals were investigated by the Nevada Applied Ecology Group (NAEG) to determine the influence of mounds on plutonium, americium, and uranium distributions and inventories in areas of the Nevada Test Site and Tonopah Test Range. Those radioactive elements were added to the environment as a result of safety experiments of nuclear devices. Two studies were conducted. The first was to estimate the vertical distribution of americium in the blow-sand mounds and in the desert pavement surrounding the mounds. The second was to estimate the amount or concentration of the radioactive materials accumulated in the mound relative to the desert pavement. Five mound types were identified in which plutonium, americium, and uranium concentrations were measured: grass, shrub, complex, animal, and diffuse. The mount top (that portion above the surrounding land surface datum), the mound bottom (that portion below the mound to a depth of 5 cm below the surrounding land surface datum), and soil from the immediate area surrounding the mound were compared separately to determine if the radioactive elements had concentrated in the mounds. Results of the studies indicate that the mounds exhibit higher concentrations of plutonium, americium, and uranium than the immediate surrounding soil. The type of mound does not appear to have influenced the amount of the radioactive material found in the mound except for the animal mounds where the burrowing activities appear to have obliterated distribution patterns

  20. Apparent field safety of a raccoon poxvirus-vectored plague vaccine in free-ranging prairie dogs (Cynomys spp.), Colorado, USA.

    Science.gov (United States)

    Tripp, Daniel W; Rocke, Tonie E; Streich, Sean P; Abbott, Rachel C; Osorio, Jorge E; Miller, Michael W

    2015-04-01

    Prairie dogs (Cynomys spp.) suffer high rates of mortality from plague. An oral sylvatic plague vaccine using the raccoon poxvirus vector (designated RCN-F1/V307) has been developed for prairie dogs. This vaccine is incorporated into palatable bait along with rhodamine B as a biomarker. We conducted trials in August and September 2012 to demonstrate uptake and apparent safety of the RCN-F1/V307 vaccine in two prairie dog species under field conditions. Free-ranging prairie dogs and other associated small rodents readily consumed vaccine-laden baits during field trials with no apparent adverse effects; most sampled prairie dogs (90%) and associated small rodents (78%) had consumed baits. Visual counts of prairie dogs and their burrows revealed no evidence of prairie dog decline after vaccine exposure. No vaccine-related morbidity, mortality, or gross or microscopic lesions were observed. Poxviruses were not isolated from any animal sampled prior to bait distribution or on sites that received placebo baits. We isolated RCN-F1/V307 from 17 prairie dogs and two deer mice (Peromyscus maniculatus) captured on sites where vaccine-laden baits were distributed. Based on these findings, studies examining the utility and effectiveness of oral vaccination to prevent plague-induced mortality in prairie dogs and associated species are underway.

  1. Apparent field safety of a raccoon poxvirus-vectored plague vaccine in free-ranging prairie dogs (Cynomys spp.), Colorado, USA

    Science.gov (United States)

    Tripp, Daniel W.; Rocke, Tonie E.; Streich, Sean P.; Abbott, Rachel C.; Osorio, Jorge E.; Miller, Michael W.

    2015-01-01

    Prairie dogs (Cynomys spp.) suffer high rates of mortality from plague. An oral sylvatic plague vaccine using the raccoon poxvirus vector (designated RCN-F1/V307) has been developed for prairie dogs. This vaccine is incorporated into palatable bait along with rhodamine B as a biomarker. We conducted trials in August and September 2012 to demonstrate uptake and apparent safety of the RCN-F1/V307 vaccine in two prairie dog species under field conditions. Free-ranging prairie dogs and other associated small rodents readily consumed vaccine-laden baits during field trials with no apparent adverse effects; most sampled prairie dogs (90%) and associated small rodents (78%) had consumed baits. Visual counts of prairie dogs and their burrows revealed no evidence of prairie dog decline after vaccine exposure. No vaccine-related morbidity, mortality, or gross or microscopic lesions were observed. Poxviruses were not isolated from any animal sampled prior to bait distribution or on sites that received placebo baits. We isolated RCN-F1/V307 from 17 prairie dogs and two deer mice (Peromyscus maniculatus) captured on sites where vaccine-laden baits were distributed. Based on these findings, studies examining the utility and effectiveness of oral vaccination to prevent plague-induced mortality in prairie dogs and associated species are underway.

  2. Criteria for onsite transfers of radioactive material

    International Nuclear Information System (INIS)

    Opperman, E.K.; Jackson, E.J.; Eggers, A.G.

    1992-01-01

    A general description of the requirements for making onsite transfers of radioactive material is provided in Chapter 2, along with the required sequencey of activities. Various criteria for package use are identified in Chapters 3-13. These criteria provide protection against undue radiation exposure. Package shielding, containment, and surface contamination requirements are established. Criteria for providing criticality safety are enumerated in Chapter 6. Criteria for providing hazards information are established in Chapter 13. A glossary is provided

  3. Two important general organizational factors: The organizational design of the safety work and the organization autocorrective system. The Italian way to improve them through criteria for the safety organizational rules

    International Nuclear Information System (INIS)

    Moramarco, C.

    1997-01-01

    A complex reality, such as a nuclear power plant, requires the maximum order in the methods of operation. A state of ''organizational confusion'' is the frequent root cause of many errors. An initial situation of organizational confusion, about one or more human allocated functions, generates further malfunctions or lacks and, what is worse, tolerates them because it makes them less visible. Order in the operators society can be improved by improving the quality of the safety organizational design and can be maintained with an effective autocorrective system. (author). 16 refs

  4. Nonreactor nuclear facilities: standards and criteria guide

    International Nuclear Information System (INIS)

    Brynda, W.J.; Junker, L.; Karol, R.C.; Lobner, P.R.; Goldman, L.A.

    1981-09-01

    This guide is a source document that identifies standards, codes, and guides that address the nuclear safety considerations pertinent to nuclear facilities as defined in DOE Order 5480.1, Chapter V, Safety of Nuclear Facilities. The guidance and criteria provided are directed toward areas of safety usually addressed in a Safety Analysis Report. The areas of safety include, but are not limited to, siting, principal design criteria and safety system design guidelines, radiation protection, accident analysis, and quality assurance. The guide is divided into two sections: general guidelines and appendices. Those guidelines that are broadly applicable to most nuclear facilities are presented in the general guidelines. These general guidelines may have limited applicability to subsurface facilities such as waste repositories. Guidelines specific to the various types or categories of nuclear facilities are presented in the appendices. These facility-specific appendices provide guidelines and identify standards and criteria that should be considered in addition to, or in lieu of, the general guidelines

  5. Nonreactor nuclear facilities: Standards and criteria guide

    International Nuclear Information System (INIS)

    Brynda, W.J.; Scarlett, C.H.; Tanguay, G.E.; Lobner, P.R.

    1986-09-01

    This guide is a source document that identifies standards, codes, and guides that address the nuclear safety considerations pertinent to nuclear facilities as defined in DOE 5480.1A, Chapter V, ''Safety of Nuclear Facilities.'' The guidance and criteria provided is directed toward areas of safety usually addressed in a Safety Analysis Report. The areas of safety include, but are not limited to, siting, principal design criteria and safety system design guidelines, radiation protection, accident analysis, conduct of operations, and quality assurance. The guide is divided into two sections: general guidelines and appendices. Those guidelines that are broadly applicable to most nuclear facilities are presented in the general guidelines. Guidelines specific to the various types or categories of nuclear facilities are presented in the appendices. These facility-specific appendices provide guidelines and identify standards and criteria that should be considered in addition to, or in lieu of, the general guidelines. 25 figs., 62 tabs

  6. Generic selection criteria for safety and patient benefit [V]: Comparing the pharmaceutical properties and patient usability of original and generic nasal spray containing ketotifen fumarate.

    Science.gov (United States)

    Wada, Yuko; Ami, Shyoko; Nozawa, Mitsuru; Goto, Miho; Shimokawa, Ken-Ichi; Ishii, Fumiyoshi

    The pH, osmotic pressure (cryoscopy), viscosity, squeeze force, spray angle, and spraying frequency of nasal spray containing ketotifen fumarate (1 brand-name product and 8 generic products) were measured. Based on the results of pH measurement, all products were weakly acidic (4.0 to 5.1). For all products, the osmotic pressure ratio to physiological saline was approximately 1. The viscosity of various products ranged from approximately 1.0 to 1.5 mPa·s. The spray angle of drug solution differed among the products: minimum, 46 degrees (Sawai and Fusachol); and maximum, 68.7 degrees (Sekiton). In particular, TOA, Sawai, Fusachol, and TYK showed significantly smaller angles compared to Zaditen (brand-name product). Container properties varied among the products: minimum squeeze force, 19.0 N (Sekiton); and maximum squeeze force, 43.1 N (Sawai). Based on these results, although all the above products are identical in dosage form and active ingredient, the differences in pharmaceutical properties, such as container operations and drug-solution spraying/attachment, may markedly influence patients' subjective opinions.

  7. Operational safety

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The PNL Safety, Standards and Compliance Program contributed to the development and issuance of safety policies, standards, and criteria; for projects in the nuclear and nonnuclear areas. During 1976 the major emphasis was on developing criteria, instruments and methods to assure that radiation exposure to occupational personnel and to people in the environs of nuclear-related facilities is maintained at the lowest level technically and economically practicable. Progress in 1976 is reported on the preparation of guidelines for radiation exposure; Pu dosimetry studies; the preparation of an environmental monitoring handbook; and emergency instrumentation preparedness

  8. 16 CFR 1031.12 - Membership criteria.

    Science.gov (United States)

    2010-01-01

    ... Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION GENERAL COMMISSION PARTICIPATION AND COMMISSION EMPLOYEE INVOLVEMENT IN VOLUNTARY STANDARDS ACTIVITIES Employee Involvement § 1031.12 Membership criteria. (a) The Commissioners, their special assistants, and Commission officials and employees holding the...

  9. 16 CFR 1031.14 - Observation criteria.

    Science.gov (United States)

    2010-01-01

    ....14 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION GENERAL COMMISSION PARTICIPATION AND COMMISSION EMPLOYEE INVOLVEMENT IN VOLUNTARY STANDARDS ACTIVITIES Employee Involvement § 1031.14 Observation criteria. A Commission official or employee may, on occasion, attend voluntary standards meetings for the...

  10. Safety design

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Shiozawa, Shusaku

    2004-01-01

    JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

  11. Reportable Nuclide Criteria for ORNL Radioactive Waste Management Activities - 13005

    International Nuclear Information System (INIS)

    McDowell, Kip; Forrester, Tim; Saunders, Mark

    2013-01-01

    The U.S. Department of Energy's Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee generates numerous radioactive waste streams. Many of those streams contain a large number of radionuclides with an extremely broad range of concentrations. To feasibly manage the radionuclide information, ORNL developed reportable nuclide criteria to distinguish between those nuclides in a waste stream that require waste tracking versus those nuclides of such minimal activity that do not require tracking. The criteria include tracking thresholds drawn from ORNL onsite management requirements, transportation requirements, and relevant treatment and disposal facility acceptance criteria. As a management practice, ORNL maintains waste tracking on a nuclide in a specific waste stream if it exceeds any of the reportable nuclide criteria. Nuclides in a specific waste stream that screen out as non-reportable under all these criteria may be dropped from ORNL waste tracking. The benefit of these criteria is to ensure that nuclides in a waste stream with activities which meaningfully affect safety and compliance are tracked, while documenting the basis for removing certain isotopes from further consideration. (authors)

  12. Crew Transportation Technical Standards and Design Evaluation Criteria

    Science.gov (United States)

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

    2015-01-01

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

  13. 29 CFR 1904.4 - Recording criteria.

    Science.gov (United States)

    2010-07-01

    ... criteria. (Needlestick and sharps injury cases, tuberculosis cases, hearing loss cases, medical removal... Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RECORDING AND REPORTING OCCUPATIONAL INJURIES AND ILLNESSES Recordkeeping Forms and Recording Criteria § 1904.4...

  14. Range and limits of application of Sec.12, Atomic Energy Act, as a legal basis of the nuclear plant safety ordinance

    International Nuclear Information System (INIS)

    Schmidt-Preuss, Matthias

    2009-01-01

    Ensuring plant safety is a key purpose of nuclear law. Sec.7 II No.3, Atomic Energy Act, is considered the basic norm of nuclear legislation. The main requirement this embodies is ensuring 'the provisions against damage arising from construction and operation of a plant as required in accordance with the state of the art'. These normative requirements constitute the strictest yardstick existing in legislation about technology. Putting it into effect has always been the purpose of the set of nuclear rules and regulations constituting the next lower level of legislation, which so far have developed by evolution and are now to be updated comprehensively in the format of so-called modules as provided for in the concept of the Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU). So far, there has not been a nuclear plant safety ordinance. The Atomic Energy Act has always provided a basis for adopting such an ordinance, especially so in Sec.12 I 1 No.1, Atomic Energy Act. No federal government has so far wanted to make use of it. This makes it all the more remarkable that the BMU took up the subject of a nuclear plant safety ordinance as early as in 2006, starting a dialog with the federal states. This dialog meanwhile has come to a halt. The subject seems to be dormant right now, but certainly has not been shelved. Ensuring plant safety is a key purpose of nuclear law. Sec.7 II No.3, Atomic Energy Act, is considered the basic norm of nuclear legislation. The main requirement this embodies is ensuring 'the provisions against damage arising from construction and operation of a plant as required in accordance with the state of the art'. These normative requirements constitute the strictest yardstick existing in legislation about technology. Putting it into effect has always been the purpose of the set of nuclear rules and regulations constituting the next lower level of legislation, which so far have developed by evolution and are now to be

  15. Multiaxial fatigue criteria for AISI 304 and 2-1/4 Cr-1 Mo steel at 5380C with applications to strain-range partitioning and linear summation of creep and fatigue damage

    International Nuclear Information System (INIS)

    Blass, J.J.

    1982-01-01

    An improved multiaxial fatigue failure criterion was developed based on the results of combined axial-torsional strain cycling tests of AISI 304 and 2-1/4 Cr-1 Mo steel conducted at 538 0 C (1000 0 F). The formulation of this criterion involves the shear and normal components of inelastic strain range on the planes of maximum inelastic shear strain range. Optimum values of certain parameters contained in the formulation were obtained for each material by the method of least squares. The ability of this criterion to correlate the test results was compared with that of the usual (Mises) equivalent inelastic strain range criterion. An improved definition of equivalent inelastic strain range resulting from these considerations was used to generalize the theory of Strain Range Partitioning to multiaxial stress-strain conditions and was also applied to the linear summation of creep and fatigue damage

  16. Design criteria for advanced reactors

    International Nuclear Information System (INIS)

    Dennielou, Y.

    1991-01-01

    Design criteria for advanced reactors are discussed, including safety aspects, site selection, problems related to maintenance and possibility of repairing or replacing structures or components of a nuclear power plant, the human factor considerations. Bearing in mind that some of these criteria are the subject of consensus at international level, the author suggests to establish a table of different operator requirements, to prepare a dossier on the comparison of input data for probabilistic risk analysis, to take into consideration the means to control a severe accident from the very start of the design

  17. Safety criteria: Intercomparison and aggregation of risks

    International Nuclear Information System (INIS)

    Yadigaroglu, G.

    1987-08-01

    Review of the assumptions underlying the use of limit lines in nuclear regulation and risk comparisons led to an alternative approach, where frequency-consequence curves are divided into two components: number of accidents during a specified time period and consequences conditional upon the occurrence of an unwanted event. This approach leads to a neat separation between natural and technological disasters. To compare probability distributions when stochastic dominance principles are not applicable, risk indices based on a linearized moments model (LMM) were established. The LMM permits explicit introduction of individual, societal or group opinions, thus incorporating the various perceptions of risks by society directly into the evaluation process. A questionnaire was applied to elicit empirical data which were then compared with the predictions of the linearized moments model. Furthermore the LMM also allows the comparison of events in those cases where the cumulative probability distributions cross each other. For the evaluation of a probability distribution that crosses the limit line, the risk index of the limit line (calculated by using the LMM) should be compared to the index of the accident under consideration

  18. Safety analysis and lay-out aspects of shieldings against particle radiation at the example of spallation facilities in the megawatt range

    International Nuclear Information System (INIS)

    Hanslik, R.

    2006-08-01

    This paper discusses the shielding of particle radiation from high current accelerators, spallation neutron sources and so called ADS-facilities (Accelerator Driven Systems). ADS-facilities are expected to gain importance in the future for transmutation of long-lived isotopes from fission reactors as well as for energy production. In this paper physical properties of the radiation as well as safety relevant requirements and corresponding shielding concepts are discussed. New concepts for the layout and design of such shielding are presented. Focal point of this work will be the fundamental difference between conventional fission reactor shielding and the safety relevant issues of shielding from high-energy radiation. Key point of this paper is the safety assessment of shielding issues of high current accelerators, spallation targets and ADS-blanket systems as well as neutron scattering instruments at spallation neutron sources. Safety relevant shielding requirements are presented and discussed. For the layout and design of the shielding for spallation sources computer base calculations methods are used. A discussion and comparison of the most important methods like semi-empirical, deterministic and stochastic codes are presented. Another key point within the presented paper is the discussion of shielding materials and their shielding efficiency concerning different types of radiation. The use of recycling material, as a cost efficient solution, is discussed. Based on the conducted analysis, flowcharts for a systematic layout and design of adequate shielding for targets and accelerators have been developed and are discussed in this paper. By use of these flowcharts layout and engineering design of future ADS-facilities can be performed. (orig.)

  19. Nuclear reactor philosophy and criteria

    International Nuclear Information System (INIS)

    Atchison, R.J.

    1979-07-01

    Nuclear power plant safety criteria and principles developed in Canada are directed towards minimizing the chance of failure of the fuel and preventing or reducing to an acceptably low level the escape of fission products should fuel failure occur. Safety criteria and practices are set forth in the Reactor Siting Guide, which is based upon the concept of defence in depth. The Guide specifies that design and construction shall follow the best applicable code, standard or practice; the total of all serious process system failures shall not exceed one in three years; special safety systems are to be physically and functionally separate from process systems and each other; and safety systems shall be testable, with unavailability less than 10 - 3 . Doses to the most exposed member of the public due to normal operation, serious process failures, and dual failures are specified. Licensees are also required to consider the effects of extreme conditions due to airplane crashes, explosions, turbine disintegration, pipe burst, and natural disasters. Safety requirements are changing as nuclear power plant designs evolve and in response to social and economic pressures

  20. Criteria of site assessment

    International Nuclear Information System (INIS)

    Gibbs, P.; Fuchs, H.

    1975-01-01

    The criteria which lead to the choice of a particular site for a nuclear power station are in general very similar to those which would apply to any other type of power station. The principal differences derive from the simpler transport problems for the fuel compared with, say, solid fuel and the special safety considerations which attach to nuclear reactors. The search for a suitable site obviously starts by considering where the power is needed, i.e. where the load centers are and also the existing transmission network which may help to bring the power from a more remote site to the load centers. This economic incentive to put the plant close to loads conflicts directly with the nuclear safety argument which favours more remote siting, and part of the problem of site selection is to reconcile these two matters. In addition, there are many other important matters which will be considered later concerning the adequacy of cooling water supplies, foundation conditions, etc., all of which must be examined in considerable detail. (orig./TK) [de

  1. Radiological criteria in nuclear emergencies

    International Nuclear Information System (INIS)

    Carrillo, D.; Diaz de la Cruz, F.

    1985-01-01

    It is pretended to enlighten the way to adopt the recommendations, from supranational organizations or the practices followed in other countries, to the peculiarities existing in Spain for the specific case of Nuclear Emergency Response Planning. The adaptation has been focalized in the criteria given by the Spanish Nuclear Safety Council and has taken into account the radiological protection levels, which have been considered adequate for Spanish population in case of nuclear accidents. (author)

  2. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical

  3. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical.

  4. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory. Part 2, Chemical constituents

    Energy Technology Data Exchange (ETDEWEB)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical.

  5. Appendix C: safety design rationale

    International Nuclear Information System (INIS)

    Ghose, S.

    1985-01-01

    A brief discussion of the rationale for safety design of fusion plants is presented in the main text. Further detail safety considerations are presented in this appendix in the form of charts and tables. The author present some of the major safety criteria and other criteria used in blanket selection here

  6. The use of economic criteria in providing a basis for safe reactor operation

    International Nuclear Information System (INIS)

    Graham, J.

    1989-01-01

    Probabilistic criteria based upon an acceptance measure of protection for owner investment can complete the range of design probabilistic criteria between those set by acceptance public safety and those set by acceptable reliability in plant operation. Criteria which address the protection of owner investment have the benefit of lowering risk in adjacent risk regions by providing greater reliability in operation as well as less risk to the safety of the public and the environment. Such investment protection criteria are currently being used to extend plant life but they could also be used very beneficially as part of the initial design process. In this paper trial criteria are suggested which address the risk of extended plant shutdown with the resultant necessity to purchase replacement power, and the risk of replacement of expensive plant components. Additional financial assessment is required to ensure that there is a proper correlation between acceptable measures of owner-investment protection and the levels of probabilistic defence suggested, but the trial criteria proposed can be used as important practical design criteria

  7. Arrangement between the Health and Safety Executive of the United Kingdom of Great Britain and Northern Ireland and the Minister of the Interior of the Federal Republic of Germany for a continuing exchange of information on significant matters pertaining to the safety of nuclear installations and on collaboration in the development of regulatory safety criteria

    International Nuclear Information System (INIS)

    1979-01-01

    According to this Agreement, information is exchanged by communication of reports, research results and studies as well as by mutual information on measures and resolutions concerning the safety of nuclear installations. Reports and information also include decisions and enquiries by courts of law on matters of safety. Co-operation in the drafting of safety standards comprises mutual information about work undertaken or planned and the exchange of texts of law, rules and regulations. (NEA) [fr

  8. Achievement report on research and development in the Sunshine Project in fiscal 1977. Studies on hydrogen energy total systems and the safety assuring technologies thereon (Studies on preparing criteria for the safety assuring technologies for hydrogen energy total systems); 1977 nendo suiso energy total system to sono hoan gijutsu ni kansuru kenkyu seika hokokusho. Suiso energy total system no hoan gijutsu kijun no sakusei ni kansuru kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-03-01

    Studies have been made on preparing criteria for the safety assuring technologies for hydrogen energy total systems. The outline of the technological guideline for hydrogen manufacturing processes in the high temperature and pressure water decomposition method is the same as that in the normal pressure water decomposition method. However, its high temperature and pressure environment can cause new safety problems. Considerations should be given on, for example, material problems in structural materials and insulation materials including electrodes and membranes, introduction of gas-liquid separation and pressure balancing devices, problems in electrolyte circulation, and safety problems that may occur because of generation of hydrogen and oxygen under high temperature and pressure conditions. This paper summarizes these matters by surveying literature data. In order to provide basic information to prepare criteria for safety assuring technologies for the gaseous hydrogen liquefaction process, surveys and studies were made based on different items of technological information and experimental study results. Safety assuring technologies were discussed on metal hydrides (promising means for storing hydrogen). Powder is used to enhance hydrogen absorbing performance, whereas the metal hydrides are pulverized as a result of repetition of absorption and discharge of hydrogen. This paper describes also metal dust explosion disaster and its risk of occurrence. (NEDO)

  9. Preliminary investigation on reliability assessment of passive safety system

    International Nuclear Information System (INIS)

    Huang Changfan; Kuang Bo

    2012-01-01

    The reliability evaluation of passive safety system plays an important part in probabilistic safety assessment (PSA) of nuclear power plant applying passive safety design, which depends quantitatively on reliabilities of passive safety system. According to the object of reliability assessment of passive safety system, relevant parameters are identified. Then passive system behavior during accident scenarios are studied. A practical example of this method is given for the case of reliability assessment of AP1000 passive heat removal system in loss of normal feedwater accident. Key and design parameters of PRHRS are identified and functional failure criteria are established. Parameter combinations acquired by Latin hyper~ cube sampling (LHS) in possible parametric ranges are input and calculations of uncertainty propagation through RELAP5/MOD3 code are carried out. Based on the calculations, sensitivity assessment on PRHRS functional criteria and reliability evaluation of the system are presented, which might provide further PSA with PRHR system reliability. (authors)

  10. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  11. Qualitative acceptance criteria for radioactive wastes to be disposed of in deep geological formations

    International Nuclear Information System (INIS)

    1990-05-01

    The present Safety Guide has to be seen as a companion document to the IAEA Safety Series No. 99. It is concerned with the waste form which is an important component of the overall disposal system. Because of the broad range of waste types and conditioned forms and variations in the sites, designs and constructional approaches being considered for deep geological repositories, this report necessarily approaches the waste acceptance criteria in a general way, recognizing that the assignment of quantitative limits to these criteria has to be the responsibility of national authorities. The main objective of this Safety Guide is to set out qualitative waste acceptance criteria as a basis for specifying quantitative limits for the waste forms and packages which are intended to be disposed of in deep geological repositories. It should serve as guidance for assigning such parameter values which would fully comply with the safety assessment and performance of a waste disposal system as a whole. This document is intended to serve both national authorities and regulatory bodies involved in the development of deep underground disposal systems. The qualitative waste acceptance criteria dealt with in the present Safety Guide are primarily concerned with the disposal of high level, intermediate level and long-lived alpha bearing wastes in deep geological repositories. Although some criteria are also applicable in other waste disposal concepts, it has to be borne in mind that the set of criteria presented here shall ensure the isolation capability of a waste disposal system for periods of time much longer than for other waste streams with shorter lifetimes. 51 refs, 1 tab

  12. Packaging design criteria for the Hanford Ecorok Packaging

    International Nuclear Information System (INIS)

    Mercado, M.S.

    1996-01-01

    The Hanford Ecorok Packaging (HEP) will be used to ship contaminated water purification filters from K Basins to the Central Waste Complex. This packaging design criteria documents the design of the HEP, its intended use, and the transportation safety criteria it is required to meet. This information will serve as a basis for the safety analysis report for packaging

  13. Review of fatigue criteria development for HTGR core supports

    International Nuclear Information System (INIS)

    Ho, F.H.; Vollman, R.E.

    1979-10-01

    Fatigue criteria for HTGR core support graphite structure are presented. The criteria takes into consideration the brittle nature of the material, and emphasizes the probabilistic approach in the treatment of strength data. The stress analysis is still deterministic. The conventional cumulative damage approach is adopted here. A specified minimum S-N curve is defined as the curve with 99% probability of survival at a 95% confidence level to accommodate random variability of the material strength. A constant life diagram is constructed to reconcile the effect of mean stress. The linear damage rule is assumed to account for the effect of random cycles. An additional factor of safety of three on cycles is recommended. The uniaxial S-N curve is modified in the medium-to-high cycle range (> 2 x 10 3 cycles) for mutiaxial fatigue effects

  14. FFTF criteria for run to cladding breach experiments

    International Nuclear Information System (INIS)

    Van Keuren, J.C.; Heard, F.J.; Stepnewski, D.D.

    1985-12-01

    The review of experiments proposed for irradiation in FFTF resulted in the development of new criteria for run-to-cladding breach experiments. These criteria have allowed irradiation of aggressive experiments without compromising the safety bases for FFTF. This paper consisting of a set of narrated slides, discusses these criteria and related bases

  15. Long-range hazard assessment of volcanic ash dispersal for a Plinian eruptive scenario at Popocatépetl volcano (Mexico): implications for civil aviation safety

    Science.gov (United States)

    Bonasia, Rosanna; Scaini, Chiara; Capra, Lucia; Nathenson, Manuel; Siebe, Claus; Arana-Salinas, Lilia; Folch, Arnau

    2014-01-01

    Popocatépetl is one of Mexico's most active volcanoes threatening a densely populated area that includes Mexico City with more than 20 million inhabitants. The destructive potential of this volcano is demonstrated by its Late Pleistocene-Holocene eruptive activity, which has been characterized by recurrent Plinian eruptions of large magnitude, the last two of which destroyed human settlements in pre-Hispanic times. Popocatépetl's reawakening in 1994 produced a crisis that culminated with the evacuation of two villages on the northeastern flank of the volcano. Shortly after, a monitoring system and a civil protection contingency plan based on a hazard zone map were implemented. The current volcanic hazards map considers the potential occurrence of different volcanic phenomena, including pyroclastic density currents and lahars. However, no quantitative assessment of the tephra hazard, especially related to atmospheric dispersal, has been performed. The presence of airborne volcanic ash at low and jet-cruise atmospheric levels compromises the safety of aircraft operations and forces re-routing of aircraft to prevent encounters with volcanic ash clouds. Given the high number of important airports in the surroundings of Popocatépetl volcano and considering the potential threat posed to civil aviation in Mexico and adjacent regions in case of a Plinian eruption, a hazard assessment for tephra dispersal is required. In this work, we present the first probabilistic tephra dispersal hazard assessment for Popocatépetl volcano. We compute probabilistic hazard maps for critical thresholds of airborne ash concentrations at different flight levels, corresponding to the situation defined in Europe during 2010, and still under discussion. Tephra dispersal mode is performed using the FALL3D numerical model. Probabilistic hazard maps are built for a Plinian eruptive scenario defined on the basis of geological field data for the "Ochre Pumice" Plinian eruption (4965 14C yr BP

  16. Long-range hazard assessment of volcanic ash dispersal for a Plinian eruptive scenario at Popocatépetl volcano (Mexico): implications for civil aviation safety

    Science.gov (United States)

    Bonasia, Rosanna; Scaini, Chirara; Capra, Lucia; Nathenson, Manuel; Siebe, Claus; Arana-Salinas, Lilia; Folch, Arnau

    2013-01-01

    Popocatépetl is one of Mexico’s most active volcanoes threatening a densely populated area that includes Mexico City with more than 20 million inhabitants. The destructive potential of this volcano is demonstrated by its Late Pleistocene–Holocene eruptive activity, which has been characterized by recurrent Plinian eruptions of large magnitude, the last two of which destroyed human settlements in pre-Hispanic times. Popocatépetl’s reawakening in 1994 produced a crisis that culminated with the evacuation of two villages on the northeastern flank of the volcano. Shortly after, a monitoring system and a civil protection contingency plan based on a hazard zone map were implemented. The current volcanic hazards map considers the potential occurrence of different volcanic phenomena, including pyroclastic density currents and lahars. However, no quantitative assessment of the tephra hazard, especially related to atmospheric dispersal, has been performed. The presence of airborne volcanic ash at low and jet-cruise atmospheric levels compromises the safety of aircraft operations and forces re-routing of aircraft to prevent encounters with volcanic ash clouds. Given the high number of important airports in the surroundings of Popocatépetl volcano and considering the potential threat posed to civil aviation in Mexico and adjacent regions in case of a Plinian eruption, a hazard assessment for tephra dispersal is required. In this work, we present the first probabilistic tephra dispersal hazard assessment for Popocatépetl volcano. We compute probabilistic hazard maps for critical thresholds of airborne ash concentrations at different flight levels, corresponding to the situation defined in Europe during 2010, and still under discussion. Tephra dispersal mode is performed using the FALL3D numerical model. Probabilistic hazard maps are built for a Plinian eruptive scenario defined on the basis of geological field data for the “Ochre Pumice” Plinian eruption (4965 14C

  17. Quantifying reactor safety margins: Application of CSAU [Code Scalability, Applicability and Uncertainty] methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes

    International Nuclear Information System (INIS)

    Wulff, W.; Boyack, B.E.; Duffey, R.B.

    1988-01-01

    Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper. Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PF1/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. 25 refs., 10 figs., 11 tabs

  18. Modern dimensioning criteria for pressure vessels

    International Nuclear Information System (INIS)

    Roche, Roland.

    1975-01-01

    Some ideas on modern dimensioning criteria are given and their advantages with regard to both safety and economy are shown. In general these criteria result from considerations on possible damage to the apparatus in service and the modes of breakdown liable to follow. They are general enough to allow for a variety of dimensioning methods both experimental and theoretical, with special reference to modern computerized digital analysis techniques. As a practical example however some notions are given on the simplest means of computing dimensions in accordance with these criteria [fr

  19. ACL Return to Sport Guidelines and Criteria.

    Science.gov (United States)

    Davies, George J; McCarty, Eric; Provencher, Matthew; Manske, Robert C

    2017-09-01

    Because of the epidemiological incidence of anterior cruciate ligament (ACL) injuries, the high reinjury rates that occur when returning back to sports, the actual number of patients that return to the same premorbid level of competition, the high incidence of osteoarthritis at 5-10-year follow-ups, and the effects on the long-term health of the knee and the quality of life for the patient, individualizing the return to sports after ACL reconstruction (ACL-R) is critical. However, one of the challenging but unsolved dilemmas is what criteria and clinical decision making should be used to return an athlete back to sports following an ACL-R. This article describes an example of a functional testing algorithm (FTA) as one method for clinical decision making based on quantitative and qualitative testing and assessment utilized to make informed decisions to return an athlete to their sports safely and without compromised performance. The methods were a review of the best current evidence to support a FTA. In order to evaluate all the complicated domains of the clinical decision making for individualizing the return to sports after ACL-R, numerous assessments need to be performed including the biopsychosocial concepts, impairment testing, strength and power testing, functional testing, and patient-reported outcomes (PROs). The optimum criteria to use for individualizing the return to sports after ACL-R remain elusive. However, since this decision needs to be made on a regular basis with the safety and performance factors of the patient involved, this FTA provides one method of quantitatively and qualitatively making the decisions. Admittedly, there is no predictive validity of this system, but it does provide practical guidelines to facilitate the clinical decision making process for return to sports. The clinical decision to return an athlete back into competition has significant implications ranging from the safety of the athlete, to performance factors and actual

  20. The Health and Safety Executive's regulatory framework for control of nuclear criticality safety

    International Nuclear Information System (INIS)

    Smith, K.; Simister, D.N.

    1991-01-01

    In the United Kingdom the Health and Safety at Work Act, 1974 is the main legal instrument under which risks to people from work activities are controlled. Certain sections of the Nuclear Installations Act, 1965 which deal with the licensing of nuclear sites and the regulatory control of risks arising from them, including the risk from accidental criticality, are relevant statutory provisions of the Health and Safety at Work Act. The responsibility for safety rests with the operator who has to make and implement arrangements to prevent accidental criticality. The adequacy of these arrangements must be demonstrated in a safety case to the regulatory authorities. Operators are encouraged to treat each plant on its own merits and develop the safety case accordingly. The Nuclear Installations Inspectorate (NII), for its part, assesses the adequacy of the operator's safety case against the industry's own standards and criteria, but more particularly against the NII's safety assessment principles and guides, and international standards. Risks should be made as low as reasonably practicable. Generally, the NII seeks improvements in safety using an enforcement policy which operates at a number of levels, ranging from persuasion through discussion to the ultimate deterrent of withdrawal of a site licence. This paper describes the role of the NII, which includes a specialist criticality expertise, within the Health and Safety Executive, in regulating the nuclear sites from the criticality safety viewpoint. (Author)

  1. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report describes health and safety concerns associated with the Mixed and Low-level Waste Treatment Facility at the Idaho National Engineering Laboratory. Various hazards are described such as fire, electrical, explosions, reactivity, temperature, and radiation hazards, as well as the potential for accidental spills, exposure to toxic materials, and other general safety concerns

  2. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory. Part 1, Waste streams and treatment technologies

    Energy Technology Data Exchange (ETDEWEB)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report describes health and safety concerns associated with the Mixed and Low-level Waste Treatment Facility at the Idaho National Engineering Laboratory. Various hazards are described such as fire, electrical, explosions, reactivity, temperature, and radiation hazards, as well as the potential for accidental spills, exposure to toxic materials, and other general safety concerns.

  3. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report describes health and safety concerns associated with the Mixed and Low-level Waste Treatment Facility at the Idaho National Engineering Laboratory. Various hazards are described such as fire, electrical, explosions, reactivity, temperature, and radiation hazards, as well as the potential for accidental spills, exposure to toxic materials, and other general safety concerns.

  4. Underwater Ranging

    OpenAIRE

    S. P. Gaba

    1984-01-01

    The paper deals with underwater laser ranging system, its principle of operation and maximum depth capability. The sources of external noise and methods to improve signal-to-noise ratio are also discussed.

  5. The development of safety requirements

    International Nuclear Information System (INIS)

    Jorel, M.

    2009-01-01

    This document describes the safety approach followed in France for the design of nuclear reactors. This safety approach is based on safety principles from which stem safety requirements that set limiting values for specific parameters. The improvements in computerized simulation, the use of more adequate new materials, a better knowledge of the concerned physical processes, the changes in the reactor operations (higher discharge burnups for instance) have to be taken into account for the definition of safety criteria and the setting of limiting values. The developments of the safety criteria linked to the risks of cladding failure and loss of primary coolant are presented. (A.C.)

  6. Learners' Epistemic Criteria for Good Scientific Models

    Science.gov (United States)

    Pluta, William J.; Chinn, Clark A.; Duncan, Ravit Golan

    2011-01-01

    Epistemic criteria are the standards used to evaluate scientific products (e.g., models, evidence, arguments). In this study, we analyzed epistemic criteria for good models generated by 324 middle-school students. After evaluating a range of scientific models, but before extensive instruction or experience with model-based reasoning practices,…

  7. Biological Water Quality Criteria

    Science.gov (United States)

    Page contains links to Technical Documents pertaining to Biological Water Quality Criteria, including, technical assistance documents for states, tribes and territories, program overviews, and case studies.

  8. 46 CFR 154.466 - Design criteria.

    Science.gov (United States)

    2010-10-01

    ... GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS FOR... § 154.466 Design criteria. (a) The insulation for a cargo tank without a secondary barrier must be... cargo tank with a secondary barrier must be designed for the secondary barrier at the design temperature...

  9. 46 CFR 154.460 - Design criteria.

    Science.gov (United States)

    2010-10-01

    ... GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS FOR... Barrier § 154.460 Design criteria. At static angles of heel up through 30°, a secondary barrier must (a) If a complete secondary barrier is required in § 154.459, hold all of the liquid cargo in the cargo...

  10. Criteria for maintenance and repair - LMFBR steam generators

    International Nuclear Information System (INIS)

    Essebaggers, J.

    1975-01-01

    The maintenance and repair criteria will be reviewed with respect to the designs presently under construction for the SNR-300 plant. This criteria shall be based upon the philosophy that safety and reliability are of the highest importance at all operating modes, while availability shall be maximized. To maximize the safety of the steam generator, measures have been taken to reduce the possibilities of failure by simplicity in design, choice of material, methods of fabrication and high quality assurance of critical parts of the pressure boundaries. The maintenance and repair program shall meet the same criteria or the intent of these criteria as applied for the original product. (author)

  11. Measuring and managing safety at Wahleach Dam

    International Nuclear Information System (INIS)

    Salmon, G. M.; Cattanach, J. D.; Hartford, D. N. D.

    1996-01-01

    Safety improvements recently implemented at the Wahleach Dam were described as one of the first instances in international dam safety practice where risk concepts have been used in conjunction with acceptable risk criteria to evaluate safety of a dam relative to required level of safety. Erosion was identified as the greatest threat to the safety of the dam. In addressing the deficiencies B.C. Hydro formulated a process which advocates a balanced level of safety,i.e. the probability of failure multiplied by the consequences of failure, integrated over a range of initiators. If the risk posed by the dam is lower than a 'tolerable' risk, the dam is considered to be safe enough. In the case of the Wahleach Dam, the inflow design flood (IDF) was selected to be about one half of the probable maximum flow (PMF), hence it was more likely than not that the spillway could pass floods up to and including the PMF. By accepting the determined level of risk, expenditures of several million dollars for design and construction of dam safety improvements were made redundant. Another byproduct of this new concept of risk assessment was the establishment of improved life safety protection by means of an early warning system for severe floods through the downstream community and emergency authorities. 3 refs., 5 tabs

  12. CCS site characterisation criteria

    Energy Technology Data Exchange (ETDEWEB)

    Bachu, S.; Hawkes, C.; Lawton, D.; Pooladi-Darvish, M.; Perkins, E.

    2009-12-15

    IEA GHG recently commissioned the Alberta Research Counil in Canada to conduct a review of storage site selection criteria and site characterisation methods in order to produce a synthesis report. This report reviews the literature on the subject on the site seleciton and characterisation since the publication of the IPCC Special Report on CCS, and provides a synthesis and classification of criteria. 161 refs.

  13. Green Supplier Selection Criteria

    DEFF Research Database (Denmark)

    Nielsen, Izabela Ewa; Banaeian, Narges; Golinska, Paulina

    2014-01-01

    Green supplier selection (GSS) criteria arise from an organization inclination to respond to any existing trends in environmental issues related to business management and processes, so GSS is integrating environmental thinking into conventional supplier selection. This research is designed...... to determine prevalent general and environmental supplier selection criteria and develop a framework which can help decision makers to determine and prioritize suitable green supplier selection criteria (general and environmental). In this research we considered several parameters (evaluation objectives......) to establish suitable criteria for GSS such as their production type, requirements, policy and objectives instead of applying common criteria. At first a comprehensive and deep review on prevalent and green supplier selection literatures performed. Then several evaluation objectives defined to assess the green...

  14. Safety philosophy and safety technology of the Soviet RBMK reactors

    International Nuclear Information System (INIS)

    Zuend, H.; Jarvis, A.S.; Haennis, H.P.; Tikal, J.

    1986-01-01

    Safety requirements and control in USSR are outlined. Safety criteria and practical application in the case of the RBMK type reactor Chernobyl-4 are discussed. An overview of the Chernobyl-4 reactor accident including its causes is given. Measures to improve the safety of RBMK reactors are described

  15. Parameters and criteria influencing the selection of waste emplacement configurations in mined geologic repositories

    International Nuclear Information System (INIS)

    Bechthold, W.; Closs, K.D.; Papp, R.

    1988-01-01

    Reference concepts for repositories in deep geological formations have been developed in several countries. For these concepts, emplacement configurations vary within a wide range that comprises drift emplacement of unshielded or self-shielded packages and horizontal or vertical borehole emplacement. This is caused by different parameters, criteria, and criteria weighting factors. Examples for parameters are the country's nuclear power program and waste management policy, its geological situation, and safety requirements, examples for criteria and repository area requirements, expenditures of mining and drilling, and efforts for emplacement and, if required, retrieval. Due to the variety of these factors and their ranking in different countries, requirements for a safe, dependable and cost-effective disposal of radioactive waste can be met in various ways

  16. Redefining design criteria for Pu-238 gloveboxes

    International Nuclear Information System (INIS)

    Acosta, S.V.

    1998-01-01

    Enclosures for confinement of special nuclear materials (SNM) have evolved into the design of gloveboxes. During the early stages of glovebox technology, established practices and process operation requirements defined design criteria. Proven boxes that performed and met or exceeded process requirements in one group or area, often could not be duplicated in other areas or processes, and till achieve the same success. Changes in materials, fabrication and installation methods often only met immediate design criteria. Standardization of design criteria took a big step during creation of ''Special-Nuclear Materials R and D Laboratory Project, Glovebox standards''. The standards defined design criteria for every type of process equipment in its most general form. Los Alamos National Laboratory (LANL) then and now has had great success with Pu-238 processing. However with ever changing Environment Safety and Health (ES and H) requirements and Ta-55 Facility Configuration Management, current design criteria are forced to explore alternative methods of glovebox design fabrication and installation. Pu-238 fuel processing operations in the Power Source Technologies Group have pushed the limitations of current design criteria. More than half of Pu-238 gloveboxes are being retrofitted or replaced to perform the specific fuel process operations. Pu-238 glovebox design criteria are headed toward process designed single use glovebox and supporting line gloveboxes. Gloveboxes that will house equipment and processes will support TA-55 Pu-238 fuel processing needs into the next century and extend glovebox expected design life

  17. Multimodal freight investment criteria.

    Science.gov (United States)

    2010-07-01

    Literature was reviewed on multi-modal investment criteria for freight projects, examining measures and techniques for quantifying project benefits and costs, as well as ways to describe the economic importance of freight transportation. : A limited ...

  18. Water Quality Criteria

    Science.gov (United States)

    EPA develops water quality criteria based on the latest scientific knowledge to protect human health and aquatic life. This information serves as guidance to states and tribes in adopting water quality standards.

  19. Aquatic Life Criteria - Ammonia

    Science.gov (United States)

    Documents related to EPA's final 2013 Aquatic Life Ambient Water Quality Criteria for Ammonia (Freshwater). These documents pertain to the safe levels of Ammonia in water that should protect to the majority of species.

  20. Aquatic Life Criteria - Copper

    Science.gov (United States)

    Documents pertain to Aquatic Life Ambient Water Quality criteria for Copper (2007 Freshwater, 2016 Estuarine/marine). These documents contain the safe levels of Copper in water that should protect to the majority of species.

  1. Integrated Criteria Document Chromium

    NARCIS (Netherlands)

    Slooff W; Cleven RFMJ; Janus JA; van der Poel P; van Beelen P; Boumans LJM; Canton JH; Eerens HC; Krajnc EI; de Leeuw FAAM; Matthijsen AJCM; van de Meent D; van der Meulen A; Mohn GR; Wijland GC; de Bruijn PJ; van Keulen A; Verburgh JJ; van der Woerd KF

    1990-01-01

    Betreft de engelse versie van rapport 758701001
    Bij dit rapport behoort een appendix onder hetzelfde nummer getiteld: "Integrated Criteria Document Chromium: Effects" Auteurs: Janus JA; Krajnc EI
    (appendix: see 710401002A)

  2. Dual Criteria Decisions

    DEFF Research Database (Denmark)

    Andersen, Steffen; Harrison, Glenn W.; Lau, Morten Igel

    2014-01-01

    The most popular models of decision making use a single criterion to evaluate projects or lotteries. However, decision makers may actually consider multiple criteria when evaluating projects. We consider a dual criteria model from psychology. This model integrates the familiar tradeoffs between...... to the clear role that income thresholds play in such decision making, but does not rule out a role for tradeoffs between risk and utility or probability weighting....

  3. A STATISTICAL APPROACH FOR DERIVING KEY NFC EVALUATION CRITERIA

    Directory of Open Access Journals (Sweden)

    S.K. KIM

    2014-02-01

    As a result of analyzing the weight of evaluation criteria with the sample of nuclear power experts and the general public, both sides recognized safety as the most important evaluation criterion, and the social factors such as public acceptance appeared to be ranked as more important evaluation criteria by the nuclear energy experts than the general public.

  4. Criteria of energy supply: a challenge for comparison

    International Nuclear Information System (INIS)

    Mueller-Reissmann, K.F.

    1980-01-01

    6 criteria for a judgment of power supply systems, in particular the 'hard' and the 'soft' way, are named: 1) Preservation of existence; 2) efficiency; 3) freedom of action; 4) safety; 5) adaptability; and 6) the principle of social ethics. Finally, the application of these criteria is discussed in a general way. (UA) [de

  5. Playground Safety

    Science.gov (United States)

    ... Prevention Fall Prevention Playground Safety Poisoning Prevention Road Traffic Safety Sports Safety Get Email Updates To receive ... at the Consumer Product Safety Commission’s Playground Safety website . References U.S. Consumer Product Safety Commission. Injuries and ...

  6. Safety performance indicators program

    International Nuclear Information System (INIS)

    Vidal, Patricia G.

    2004-01-01

    In 1997 the Nuclear Regulatory Authority (ARN) initiated a program to define and implement a Safety Performance Indicators System for the two operating nuclear power plants, Atucha I and Embalse. The objective of the program was to incorporate a set of safety performance indicators to be used as a new regulatory tool providing an additional view of the operational performance of the nuclear power plants, improving the ability to detect degradation on safety related areas. A set of twenty-four safety performance indicators was developed and improved throughout pilot implementation initiated in July 1998. This paper summarises the program development, the main criteria applied in each stage and the results obtained. (author)

  7. Study of the chemical interaction between the beryllium powders of different particles size and the air in the temperature range 500-1000degC form the viewpoint of ITER safety

    Energy Technology Data Exchange (ETDEWEB)

    Davydov, D.A. [State Scientific Center of Russian Federation, Moscow (Russian Federation); Konovalov, Y.V.; Gorokhov, V.A.; Levin, V.B.; Chekhlatov, G.M.; Khomutov, A.M.

    1998-01-01

    Under an effect of some factors characteristic for the ITER- operating condition a dense beryllium facing plasma can transit into various forms, changing its structural states. As a result of the bombardment of beryllium plasma facing components by ion fluxes, the production of a dust including the particles from a few micrometers to a few millimeters in size is possible. The specific features in the behaviour of various beryllium forms under emergency conditions are of an essential interest from the viewpoint of ITER safety. Some grades of powders of different average particles size (14-31 micron) have been produced in a given study, and their chemical interaction at high temperatures with air (500-1100degC), test duration effects simulating the emergency situation at ITER in the first approximation have been studied. The temperature dependence of beryllium powders (different particles size after disc abrased) interaction with air in the temperature range 500-1000degC at the exposure of 5 hours long for each temperature and kinetic dependence of interaction of these powders with air at 800degC for the exposure from half an hour to 7 hours long were studied. An analysis of granulometric weight fraction in the metallic and oxidized beryllium powders with different particles size has been done by the photosedimentational technique with the instrument `Analysette-20`. Construction of a mathematical model for the chemical interaction of beryllium powders with air at high temperatures have been carried out. (author)

  8. Design Processes and Criteria for the X-51A Flight Vehicle Airframe

    National Research Council Canada - National Science Library

    Lane, Jeffrey

    2007-01-01

    .... This paper summarizes the X-51A vehicle mission requirements, system design, design processes used for airframe synthesis, design safety factors, success criteria and issues facing the incorporation...

  9. Summarized water quality criteria

    International Nuclear Information System (INIS)

    Kempster, P.L.; Hattingh, W.H.J.; Van Vliet, H.R.

    1980-08-01

    The available world literature from 27 sources on existing water quality criteria are summarized for the 15 main uses of water. The minimum, median and maximum specified values for 96 different determinands are included. Under each water use the criteria are grouped according to the functional significance of the determinands e.g. aesthetic/physical effects, high toxic potential, low toxic potential etc. A synopsis is included summarizing salient facts for each determinand such as the conditions under which it is toxic and its relationship to other determinands. The significance of the criteria is briefly discussed and the importance of considering functional interactions between determinands emphasized in evaluating the potential for toxic or beneficial effects. From the source literature it appears that the toxic potential, in addition to being determined by concentration, is also affected by the origin of the substance concerned, i.e. whether from natural sources or from anthropogenic pollution

  10. General criteria for the project of nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    1979-01-01

    Recommendations are presented establishing the general criteria for the project of nuclear fuel reprocessing plants to be licensed according to the legislation in effect. They apply to all the plant's systems, components and structures which are important to operation safety and to the public's health and safety. (F.E.) [pt

  11. K-effective as a measure of criticality safety

    International Nuclear Information System (INIS)

    Venner, J.; Haley, R.M.; Bowden, R.L.

    2003-01-01

    This paper considers the relation between the neutron multiplication of a system, k-effective, and critical parameters. It aims to investigate whether k-effective is always the most appropriate measure of safety. For simple systems handbook data can be effectively utilized, applying a safety factor to critical masses. In such situations, the criticality safety margin is readily apparent. However, more complex systems may use the calculated value of neutron multiplication to assess the criticality safety of the system under investigation. A problem arises because there is no exact consistency between k-effective and the physical margin of subcriticality, in terms of parameters such as mass. In the UK, commonly accepted safety criteria are applied to limit the k-effective of the system being assessed. These margins of subcriticality have no definitive justification to support the values chosen and might be considered rather arbitrary in nature. This paper aims to answer this question of suitability by investigating the relation between k-effective and the physical critical parameters for a wide range of systems. It concludes that the safety criteria currently applied in the UK are valid, but some difference exists between safety factors applied to the mass of fissile material present and the corresponding value of k-effective. (author)

  12. PA activity by using nuclear power plant safety demonstration and analysis

    International Nuclear Information System (INIS)

    Tsuchiya, Mitsuo; Kamimae, Rie

    1999-01-01

    INS/NUPEC presents one of Public acceptance (PA) methods for nuclear power in Japan, 'PA activity by using Nuclear Power Plant Safety Demonstration and Analysis', by using one of videos which is explained and analyzed accident events (Loss of Coolant Accident). Safety regulations of The National Government are strictly implemented in licensing at each of basic design and detailed design. To support safety regulation activities conducted by the National Government, INS/NLTPEC continuously implement Safety demonstration and analysis. With safety demonstration and analysis, made by assuming some abnormal conditions, what impacts could be produced by the assumed conditions are forecast based on specific design data on a given nuclear power plants. When analysis results compared with relevant decision criteria, the safety of nuclear power plants is confirmed. The decision criteria are designed to help judge if or not safety design of nuclear power plants is properly made. The decision criteria are set in the safety examination guidelines by taking sufficient safety allowance based on the latest technical knowledge obtained from a wide range of tests and safety studies. Safety demonstration and analysis is made by taking the procedure which are summarized in this presentation. In Japan, various PA (Public Acceptance) pamphlets and videos on nuclear energy have been published. But many of them focused on such topics as necessity or importance of nuclear energy, basic principles of nuclear power generation, etc., and a few described safety evaluation particularly of abnormal and accident events in accordance with the regulatory requirements. In this background, INS/NUPEC has been making efforts to prepare PA pamphlets and videos to explain the safety of nuclear power plants, to be simple and concrete enough, using various analytical computations for abnormal and accident events. In results, PA activity of INS/NUPEC is evaluated highly by the people

  13. Nuclear safety

    International Nuclear Information System (INIS)

    2014-01-01

    The Program on Nuclear Safety comprehends Radioprotection, Radioactive Waste Management and Nuclear Material Control. These activities are developed at the Nuclear Safety Directory. The Radioactive Waste Management Department (GRR) was formally created in 1983, to promote research and development, teaching and service activities in the field of radioactive waste. Its mission is to develop and employ technologies to manage safely the radioactive wastes generated at IPEN and at its customer’s facilities all over the country, in order to protect the health and the environment of today's and future generations. The Radioprotection Service (GRP) aims primarily to establish requirements for the protection of people, as workers, contractors, students, members of the general public and the environment from harmful effects of ionizing radiation. Furthermore, it also aims to establish the primary criteria for the safety of radiation sources at IPEN and planning and preparing for response to nuclear and radiological emergencies. The procedures about the management and the control of exposures to ionizing radiation are in compliance with national standards and international recommendations. Research related to the main activities is also performed. The Nuclear Material Control has been performed by the Safeguard Service team, which manages the accountability and the control of nuclear material at IPEN facilities and provides information related to these activities to ABACC and IAEA. (author)

  14. Nuclear safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Program on Nuclear Safety comprehends Radioprotection, Radioactive Waste Management and Nuclear Material Control. These activities are developed at the Nuclear Safety Directory. The Radioactive Waste Management Department (GRR) was formally created in 1983, to promote research and development, teaching and service activities in the field of radioactive waste. Its mission is to develop and employ technologies to manage safely the radioactive wastes generated at IPEN and at its customer’s facilities all over the country, in order to protect the health and the environment of today's and future generations. The Radioprotection Service (GRP) aims primarily to establish requirements for the protection of people, as workers, contractors, students, members of the general public and the environment from harmful effects of ionizing radiation. Furthermore, it also aims to establish the primary criteria for the safety of radiation sources at IPEN and planning and preparing for response to nuclear and radiological emergencies. The procedures about the management and the control of exposures to ionizing radiation are in compliance with national standards and international recommendations. Research related to the main activities is also performed. The Nuclear Material Control has been performed by the Safeguard Service team, which manages the accountability and the control of nuclear material at IPEN facilities and provides information related to these activities to ABACC and IAEA. (author)

  15. NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-07-01

    This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules.

  16. NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

    International Nuclear Information System (INIS)

    1982-07-01

    This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules

  17. Maintenance evaluation using risk based criteria

    International Nuclear Information System (INIS)

    Torres Valle, A.

    1996-01-01

    The maintenance evaluation is currently performed by using economic and, in some case, technical equipment failure criteria, however this is done to a specific equipment level. In general, when statistics are used the analysis for maintenance optimization are made isolated and whit a post mortem character; The integration provided by mean of Probabilistic Safety assessment (PSA) together with the possibilities of its applications, allow for evaluation of maintenance on the basis of broader scope criteria in regard to those traditionally used. The evaluate maintenance using risk based criteria, is necessary to follow a dynamic and systematic approach, in studying the maintenance strategy, to allow for updating the initial probabilistic models, for including operational changes that often take place during operation of complex facilities. This paper proposes a dynamic evaluation system of maintenance task. The system is illustrated by means of a practical example

  18. Radiological design criteria

    International Nuclear Information System (INIS)

    Selby, J.M.; Andersen, B.V.; Carter, L.A.; Waite, D.A.

    1977-01-01

    Many new nuclear facilities are unsatisfactory from a radiation protection point of view, particularly when striving to maintain occupational exposure as low as practicable 'ALAP'. Radiation protection is achieved through physical protective features supplemented by administrative controls. Adequate physical protective feature should be achieved during construction so that supplemental administrative controls may be kept simple and workable. Many nuclear facilities fall short of adequate physical protective features, thus, remedial and sometimes awkward administrative procedures are required to safely conduct work. In reviewing the various handbooks, reports and regulations which deal with radiation protection, it may be noted that there is minimal radiological design guidance for application to nuclear facilities. A set of criteria or codes covering functional areas rather than specific nuclear facility types is badly needed. The following are suggested as functional areas to be considered: characterization of the Facility; siting and access; design exposure limits; layout (people and materials flow); ventilation and effluent control; radiation protection facilities and systems. The application of such radiological design criteria early in the design process would provide some assurance that nuclear facilities will be safe, flexible, and efficient with a minimum of costly retrofitting or administrative restrictions. Criteria which we have found helpful in these functional areas is discussed together with justification for adoption of such criteria and identification of problems which still require solution

  19. Comments on confinement criteria

    International Nuclear Information System (INIS)

    Kurak, V.; Schroer, B.; Swieca, J.A.

    1977-01-01

    For a QED 2 model with SU(n) flavour, the nature of the physical states space is more subtle than one expects on the basis of the loop criterion for confinement. One may have colour confinement without confinement of the fundamental flavour representation. Attempts to formulate confinement criteria in which the quark fields play a more fundamental role are discussed [pt

  20. Radiation safety

    International Nuclear Information System (INIS)

    Van Riessen, A.

    2002-01-01

    Full text: Experience has shown that modem, fully enclosed, XRF and XRD units are generally safe. This experience may lead to complacency and ultimately a lowering of standards which may lead to accidents. Maintaining awareness of radiation safety issues is thus an important role for all radiation safety officers. With the ongoing progress in technology, a greater number of radiation workers are more likely to use a range of instruments/techniques - eg portable XRF, neutron beam analysis, and synchrotron radiation analysis. The source for each of these types of analyses is different and necessitates an understanding of the associated dangers as well as use of specific radiation badges. The trend of 'suitcase science' is resulting in scientists receiving doses from a range of instruments and facilities with no coordinated approach to obtain an integrated dose reading for an individual. This aspect of radiation safety needs urgent attention. Within Australia a divide is springing up between those who work on Commonwealth property and those who work on State property. For example a university staff member may operate irradiating equipment on a University campus and then go to a CSIRO laboratory to operate similar equipment. While at the University State regulations apply and while at CSIRO Commonwealth regulations apply. Does this individual require two badges? Is there a need to obtain two licences? The application of two sets of regulations causes unnecessary confusion and increases the workload of radiation safety officers. Radiation safety officers need to introduce risk management strategies to ensure that both existing and new procedures result in risk minimisation. A component of this strategy includes ongoing education and revising of regulations. AXAA may choose to contribute to both of these activities as a service to its members as well as raising the level of radiation safety for all radiation workers. Copyright (2002) Australian X-ray Analytical

  1. Evaluation of periodic safety status analyses

    International Nuclear Information System (INIS)

    Faber, C.; Staub, G.

    1997-01-01

    In order to carry out the evaluation of safety status analyses by the safety assessor within the periodical safety reviews of nuclear power plants safety goal oriented requirements have been formulated together with complementary evaluation criteria. Their application in an inter-disciplinary coopertion covering the subject areas involved facilitates a complete safety goal oriented assessment of the plant status. The procedure is outlined briefly by an example for the safety goal 'reactivity control' for BWRs. (orig.) [de

  2. Seismic re-evaluation criteria for Bohunice V1 reconstruction

    International Nuclear Information System (INIS)

    Campbell, R.; Schlund, H.; Warnken, L.

    2001-01-01

    Bohunice V1 in Slovakia is a Russian designed two unit WWER 440, Model 230 Pressurized Water Reactor. The plant was not originally designed for earthquake. Subsequent and ongoing reassessments now confirm that the seismic hazard at the site is significant. EBO, the plant owner has contracted with a consortium lead by Siemens AG (REKON) to do major reconstruction of the plant to significantly enhance its safety systems by the addition of new systems and the upgrading of existing systems. As part of the reconstruction, a complete seismic assessment and upgrading is required for existing safety relevant structures, systems and components. It is not practical to conduct this reassessment and upgrading using criteria applied to new design of nuclear power plants. Alternate criteria may be used to achieve adequate safety goals. Utilities in the U.S. have faced several seismic issues with operating NPPs and to resolve these issues, alternate criteria have been developed which are much more cost effective than use of criteria for new design. These alternate criteria incorporate the knowledge obtained from investigation of the performance of equipment in major earthquakes and include provisions for structures and passive equipment to deform beyond the yield point, yet still provide their essential function. IAEA has incorporated features of these alternate criteria into draft Technical Guidelines for application to Bohunice V1 and V2. REKON has developed plant specific criteria and procedures for the Bohunice V1 reconstruction that incorporate major features of the U.S. developed alternate criteria, comply to local codes and which envelop the draft IAEA Technical Guidelines. Included in these criteria and procedures are comprehensive walkdown screening criteria for equipment, piping, HVAC and cable raceways, analytical criteria which include inelastic energy absorption factors defined on an element basis and testing criteria which include specific guidance on interpretation

  3. Hanford Site solid waste acceptance criteria

    International Nuclear Information System (INIS)

    Ellefson, M.D.

    1998-01-01

    Order 5820.2A requires that each treatment, storage, and/or disposal facility (referred to in this document as TSD unit) that manages low-level or transuranic waste (including mixed waste and TSCA PCB waste) maintain waste acceptance criteria. These criteria must address the various requirements to operate the TSD unit in compliance with applicable safety and environmental requirements. This document sets forth the baseline criteria for acceptance of radioactive waste at TSD units operated by WMH. The criteria for each TSD unit have been established to ensure that waste accepted can be managed in a manner that is within the operating requirements of the unit, including environmental regulations, DOE Orders, permits, technical safety requirements, waste analysis plans, performance assessments, and other applicable requirements. Acceptance criteria apply to the following TSD units: the Low-Level Burial Grounds (LLBG) including both the nonregulated portions of the LLBG and trenches 31 and 34 of the 218-W-5 Burial Ground for mixed waste disposal; Central Waste Complex (CWC); Waste Receiving and Processing Facility (WRAP); and T Plant Complex. Waste from all generators, both from the Hanford Site and from offsite facilities, must comply with these criteria. Exceptions can be granted as provided in Section 1.6. Specific waste streams could have additional requirements based on the 1901 identified TSD pathway. These requirements are communicated in the Waste Specification Records (WSRds). The Hanford Site manages nonradioactive waste through direct shipments to offsite contractors. The waste acceptance requirements of the offsite TSD facility must be met for these nonradioactive wastes. This document does not address the acceptance requirements of these offsite facilities

  4. A Statistical Approach for Deriving Key NFC Evaluation Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K; Kang, G. B.; Ko, W. I [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Young, S. R.; Gao, R. X. [Univ. of Science and Technology, Daejeon (Korea, Republic of)

    2014-02-15

    This study suggests 5 evaluation criteria (safety and technology, environmental impact, economic feasibility, social factors, and institutional factors) and 24 evaluation indicators for a NFC (nuclear fuel cycle) derived using factor analysis. To do so, a survey using 1 on 1 interview was given to nuclear energy experts and local residents who live near nuclear power plants. In addition, by conducting a factor analysis, homogeneous evaluation indicators were grouped with the same evaluation criteria, and unnecessary evaluation criteria and evaluation indicators were dropped out. As a result of analyzing the weight of evaluation criteria with the sample of nuclear power experts and the general public, both sides recognized safety as the most important evaluation criterion, and the social factors such as public acceptance appeared to be ranked as more important evaluation criteria by the nuclear energy experts than the general public.

  5. A Statistical Approach for Deriving Key NFC Evaluation Criteria

    International Nuclear Information System (INIS)

    Kim, S. K; Kang, G. B.; Ko, W. I; Young, S. R.; Gao, R. X.

    2014-01-01

    This study suggests 5 evaluation criteria (safety and technology, environmental impact, economic feasibility, social factors, and institutional factors) and 24 evaluation indicators for a NFC (nuclear fuel cycle) derived using factor analysis. To do so, a survey using 1 on 1 interview was given to nuclear energy experts and local residents who live near nuclear power plants. In addition, by conducting a factor analysis, homogeneous evaluation indicators were grouped with the same evaluation criteria, and unnecessary evaluation criteria and evaluation indicators were dropped out. As a result of analyzing the weight of evaluation criteria with the sample of nuclear power experts and the general public, both sides recognized safety as the most important evaluation criterion, and the social factors such as public acceptance appeared to be ranked as more important evaluation criteria by the nuclear energy experts than the general public

  6. Development of an Evaluation Method for Team Safety Culture Competencies using Social Network Analysis

    International Nuclear Information System (INIS)

    Han, Sang Min; Kim, Ar Ryum; Seong, Poong Hyun

    2016-01-01

    In this study, team safety culture competency of a team was estimated through SNA, as a team safety culture index. To overcome the limit of existing safety culture evaluation methods, the concept of competency and SNA were adopted. To estimate team safety culture competency, we defined the definition, range and goal of team safety culture competencies. Derivation of core team safety culture competencies is performed and its behavioral characteristics were derived for each safety culture competency, from the procedures used in NPPs and existing criteria to assess safety culture. Then observation was chosen as a method to provide the input data for the SNA matrix of team members versus insufficient team safety culture competencies. Then through matrix operation, the matrix was converted into the two meaningful values, which are density of team members and degree centralities of each team safety culture competency. Density of tem members and degree centrality of each team safety culture competency represent the team safety culture index and the priority of team safety culture competency to be improved

  7. Development of an Evaluation Method for Team Safety Culture Competencies using Social Network Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Min; Kim, Ar Ryum; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, team safety culture competency of a team was estimated through SNA, as a team safety culture index. To overcome the limit of existing safety culture evaluation methods, the concept of competency and SNA were adopted. To estimate team safety culture competency, we defined the definition, range and goal of team safety culture competencies. Derivation of core team safety culture competencies is performed and its behavioral characteristics were derived for each safety culture competency, from the procedures used in NPPs and existing criteria to assess safety culture. Then observation was chosen as a method to provide the input data for the SNA matrix of team members versus insufficient team safety culture competencies. Then through matrix operation, the matrix was converted into the two meaningful values, which are density of team members and degree centralities of each team safety culture competency. Density of tem members and degree centrality of each team safety culture competency represent the team safety culture index and the priority of team safety culture competency to be improved.

  8. FDR (drive-dynamics-control) - a new driving safety system with active control of brake and drive forces in the dynamic fringe range; FDR, ein neues Fahrsicherheitssystem mit aktiver Regelung der Brems- und Antriebskraefte im fahrdynamischen Grenzbereich

    Energy Technology Data Exchange (ETDEWEB)

    Erhardt, R. [Bosch (R.) GmbH, Stuttgart (Germany); Zanten, A.T. van [Bosch (R.) GmbH, Stuttgart (Germany)

    1995-12-31

    BOSCH is going to introduce a new driving safety system in 1995, the FDR (drive-dynamics-control). Using the measured and estimated dynamic magnitudes as a basis, the system calculates inhowfar the actual vehicle motion differs from the desired stable trace- and direction-consistent handling properties. Depending on the driving situation and driver`s wishes the braking and driving forces at the wheels are adjusted with a considerable divergence in order to achieve the desired handling properties. The system improves the driving stability in all operating states as soon as the dynamic limiting range is reached. It even reduces the risk of skidding in case of extreme steering manoeuvres and also enables the safe control of the vehicle in critical traffic situations. Furthermore the system offers improved basic anti-skid braking system and anti-slip control functions. Due to these advantages it can be expected that the FDR is going to make an important contribution to avoiding accidents and reducing damage. (orig.) [Deutsch] Mit FDR (Fahr-Dynamik-Regelung) wird BOSCH 1995 ein neues Fahrsicherheitssystem einfuehren. Das System berechnet auf der Basis gemessener und geschaetzter fahrdynamischer Groessen, wie stark die tatsaechliche Fahrzeugbewegung von einem gewuenschten stabilen, spur- und richtungstreuen Fahrverhalten abweicht. Die Brems- und Antriebskraefte an den Raedern werden bei deutlicher Abweichung abhaengig von Fahrsituation und Fahrerwunsch so eingestellt, dass die Abweichung minimiert und das gewuenschte Fahrverhalten weitgehend erreicht wird. Das System verbessert die Fahrstabilitaet in allen Betriebszustaenden, sobald der fahrdynamische Grenzbereich erreicht wird. Es reduziert selbst bei extremen Lenkmanoevern die Schleudergefahr drastisch und ermoeglicht auch in kritischen Verkehrssituationen die sicherere Beherrschung des Fahrzeugs. Darueberhinaus bietet das System verbesserte ABS- und ASR-Grundfunktionen. Diese Vorteile lassen erwarten, dass FDR einen

  9. Efficacy and safety of aprepitant for the prevention of chemotherapy-induced nausea and vomiting during the first cycle of moderately emetogenic chemotherapy in Korean patients with a broad range of tumor types.

    Science.gov (United States)

    Kim, Jeong Eun; Jang, Joung-Soon; Kim, Jae-Weon; Sung, Yong Lee; Cho, Chi-Heum; Lee, Myung-Ah; Kim, Do-Jin; Ahn, Myung-Ju; Lee, Kil Yeon; Sym, Sun Jin; Lim, Myong Choel; Jung, Hun; Cho, Eun Kim; Min, Kyung Wan

    2017-03-01

    This study evaluated the efficacy and safety of a 3-day aprepitant regimen for the prevention of chemotherapy-induced nausea and vomiting (CINV) during the first cycle of non-anthracycline plus cyclophosphamide (AC)-based moderately emetogenic chemotherapy (MEC) based on government guidelines in Korean patients. This multicenter, randomized, double-blind, phase IV trial (NCT01636947) enrolled adult South Korean patients with a broad range of tumor types who were scheduled to receive a single dose of ≥1 MEC agent. Patients were randomized to a 3-day regimen of aprepitant (aprepitant regimen) or placebo (control regimen) on top of ondansetron plus dexamethasone. The primary and key secondary efficacy endpoints were the proportions of subjects who achieved no vomiting and complete response (CR) during the overall phase. Of the 494 randomized subjects, 480 were included in the modified intent-to-treat population. Response rates for no vomiting and CR in the overall phase were numerically higher for the aprepitant regimen compared with the control regimen groups, but failed to reach statistical significance (no vomiting 77.2 vs 72.0%; p = 0.191; CR 73.4 vs 70.4%; p = 0.458). Both the aprepitant and control regimens were generally well tolerated. A 3-day aprepitant regimen was numerically better but not statistically superior to a control regimen with respect to the achievement of no vomiting or CR during the overall phase in a non-AC MEC Korean population based on government reimbursement guidelines. ClinicalTrials.gov NCT01636947 ( https://clinicaltrials.Gov/ct2/show/NCT01636947 ).

  10. Safety of nuclear installations

    International Nuclear Information System (INIS)

    Esteves, R.G.

    1987-01-01

    The safety philosophy of a PWR type reactor distinguishing three levels of safety, is presented. At the first level, the concept of reactivity defining coefficients which measure the reactivity variation is introduced. At the second level, the reactor protection system establishing the design criteria to assure the high reliability, is defined. At the third level, the protection barriers to contain the consequences of accident evolution, are defined. (M.C.K.) [pt

  11. Human Systems Design Criteria

    DEFF Research Database (Denmark)

    Rasmussen, Jens

    1982-01-01

    This paper deals with the problem of designing more humanised computer systems. This problem can be formally described as the need for defining human design criteria, which — if used in the design process - will secure that the systems designed get the relevant qualities. That is not only...... the necessary functional qualities but also the needed human qualities. The author's main argument is, that the design process should be a dialectical synthesis of the two points of view: Man as a System Component, and System as Man's Environment. Based on a man's presentation of the state of the art a set...... of design criteria is suggested and their relevance discussed. The point is to focus on the operator rather than on the computer. The crucial question is not to program the computer to work on its own conditions, but to “program” the operator to function on human conditions....

  12. Containment penetration design criteria and implementation

    International Nuclear Information System (INIS)

    Perry, R.F.; Rigamonti, G.; Dainora, J.

    1975-01-01

    A rational design criteria is presented which serves as a basis for the design and analysis of containment piping penetrations. The criteria includes the effect of temperature as well as mechanical loads for the full range of plant conditions. With this criteria various penetration flued head designs have been compared and optimization achieved. Sleeve wall dimensions and containment loads have been determined without reference to piping configuration. An interaction theory which allows the implementation of the criteria for the determination of design loads and minimum sleeve wall thickness. The interaction theory developed applies to elastic-perfectly plastic cylinders (pipes and sleeves) and accounts for the simultaneous load resultants of transverse shear force, bending moment, torsional moment, and axial force in addition to internal pipe pressure. Application of the theory developed to the determination of sleeve thickness and containment design loads is presented in detail. (Auth.)

  13. Intelligent intefrace design criteria

    International Nuclear Information System (INIS)

    Sicard, Y.; Siebert, S.; Thebault, M.H.

    1990-01-01

    Optimum adequation between control means and the capacities of the teams of operators is sought for to achieve computerization of control and monitoring interfaces. Observation of the diagnosis activity of populations of operators in incident situations on a simulator enables design criteria well-suited to the characteristics of the detection, interpretation of symptoms and incident location tasks to be defined. A software tool based on a qualitative approach enables the design process to be systematized

  14. DOE Standard: Fire protection design criteria

    International Nuclear Information System (INIS)

    1999-07-01

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

  15. Top-level regulatory criteria for the standard MHTGR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-10-15

    The Licensing Plan for the Standard MHTGR (Ref. 1) describes a program to support a U.S. Nuclear Regulatory Commission (NRC) design review and approval. The Plan calls for the submittal of Top-Level Regulatory Criteria to the NRC for concurrence with their completeness and acceptability for the MHTGR program. The Top-Level Regulatory Criteria are defined as the standards for judging licensability that directly specify acceptable limits for protection of the public health and safety and the environment. The criteria proposed herein are for normal plant operation and a broad spectrum of anticipated events, including accidents. The approach taken is to define a set of criteria which are general as opposed to being design specific. Specifically, it is recommended that criteria be met which: 1. Are less prescriptive than current regulation, thereby encouraging maximum flexibility in design approaches. 2. Are measurable. 3. Are not more strict than the criteria for current power plants.

  16. Current quality criteria

    International Nuclear Information System (INIS)

    Sajaroff, Pedro M.

    1999-01-01

    The Safety Series 50-C/SG-Q 'Quality Assurance for Safety in Nuclear Power Plants and other Nuclear Installations' published by the IAEA in 1996, contains the Code and 14 Safety Guides. It is more addressed to operating organizations than to regulatory authorities. Priority is assigned to programme implementation and to work effectiveness. Emphasis is putted on doing work correctly from the very beginning (right first time attitude), and not on finding and correcting errors instead of preventing them. The essential aspects for assuring quality (that is, for achieving and maintaining an acceptable level of safety) cover the ten basic requirements established by the Code: Management (QA programme, training and qualification, non-conformance control and corrective actions, document control and records); Performance (work, design, procurement, inspection and testing for acceptance); and Assessment (management self-assessment, independent assessment). These requirements are applicable, on a graded approach basis, to the overall QA programme of the operating organization and its partial programmes during design, construction, commissioning, operation and decommissioning stages. The requirements can be also applied, with due adaptations, to other nuclear installations. The research and development activities related to such stages are comprised under the quality 'umbrella' as well. Therefore the scope of the traditional quality assurance is expanded and consolidated as a management tool more oriented to user. Responsibilities for achieving quality objectives are assigned to all involved individuals and organizations at all levels: management, working groups and assessment teams. Managers' is important but individual responsibilities are unavoidable. For the Argentine Nuclear Regulatory Authority, performance-based quality assurance was not a novelty. In line with the regulatory philosophy (which is essentially based on performance) the regulatory standard AR 3.6.1 (Rev.1

  17. Criteria for software modularization

    Science.gov (United States)

    Card, David N.; Page, Gerald T.; Mcgarry, Frank E.

    1985-01-01

    A central issue in programming practice involves determining the appropriate size and information content of a software module. This study attempted to determine the effectiveness of two widely used criteria for software modularization, strength and size, in reducing fault rate and development cost. Data from 453 FORTRAN modules developed by professional programmers were analyzed. The results indicated that module strength is a good criterion with respect to fault rate, whereas arbitrary module size limitations inhibit programmer productivity. This analysis is a first step toward defining empirically based standards for software modularization.

  18. Safety class methodology

    International Nuclear Information System (INIS)

    Donner, E.B.; Low, J.M.; Lux, C.R.

    1992-01-01

    DOE Order 6430.1A, General Design Criteria (GDC), requires that DOE facilities be evaluated with respect to ''safety class items.'' Although the GDC defines safety class items, it does not provide a methodology for selecting safety class items. The methodology described in this paper was developed to assure that Safety Class Items at the Savannah River Site (SRS) are selected in a consistent and technically defensible manner. Safety class items are those in the highest of four categories determined to be of special importance to nuclear safety and, merit appropriately higher-quality design, fabrication, and industrial test standards and codes. The identification of safety class items is approached using a cascading strategy that begins at the 'safety function' level (i.e., a cooling function, ventilation function, etc.) and proceeds down to the system, component, or structure level. Thus, the items that are required to support a safety function are SCls. The basic steps in this procedure apply to the determination of SCls for both new project activities, and for operating facilities. The GDC lists six characteristics of SCls to be considered as a starting point for safety item classification. They are as follows: 1. Those items whose failure would produce exposure consequences that would exceed the guidelines in Section 1300-1.4, ''Guidance on Limiting Exposure of the Public,'' at the site boundary or nearest point of public access 2. Those items required to maintain operating parameters within the safety limits specified in the Operational Safety Requirements during normal operations and anticipated operational occurrences. 3. Those items required for nuclear criticality safety. 4. Those items required to monitor the release of radioactive material to the environment during and after a Design Basis Accident. Those items required to achieve, and maintain the facility in a safe shutdown condition 6. Those items that control Safety Class Item listed above

  19. Radiation protection - Performance criteria for service laboratories performing biological dosimetry by cytogenetics

    International Nuclear Information System (INIS)

    2004-01-01

    This International Standard provides criteria for quality assurance and quality control, evaluation of the performance and the accreditation of biological dosimetry by cytogenetic service laboratories. This International Standard addresses: a) the confidentiality of personal information, for the customer and the service laboratory, b) the laboratory safety requirements, c) the calibration sources and calibration dose ranges useful for establishing the reference dose-effect curves allowing the dose estimation from chromosome aberration frequency, and the minimum detection levels, d) the scoring procedure for unstable chromosome aberrations used for biological dosimetry, e) the criteria for converting a measured aberration frequency into an estimate of absorbed dose, f) the reporting of results, g) the quality assurance and quality control, h) informative annexes containing examples of a questionnaire, instructions for customers, a data sheet for recording aberrations and a sample report

  20. Operational and environmental safety

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    The responsibility of the DOE Office of Operational and Environmental Safety is to assure that DOE-controlled activities are conducted in a manner that will minimize risks to the public and employees and will provide protection for property and the environment. The program supports the various energy technologies by identifying and resolving safety problems; developing and issuing safety policies, standards, and criteria; assuring compliance with DOE, Federal, and state safety regulations; and establishing procedures for reporting and investigating accidents in DOE operations. Guidelines for the radiation protection of personnel; radiation monitoring at nuclear facilities; an assessment of criticality accidents by fault tree analysis; and the preparation of environmental, safety, and health standards applicable to geothermal energy development are discussed

  1. Balancing safety and economics

    International Nuclear Information System (INIS)

    Kroeger, W.; Fischer, P.U.

    2000-01-01

    The safety requirements of NPPs have always aimed at limiting societal risks. This risk approach initially resulted in deterministic design criteria and concepts. In the 1980s the paradigm 'safety at all costs' arose and often led to questionable backfitting measures. Conflicts between new requirements, classical design concepts and operational demands were often ignored. The design requirements for advanced reactors ensure enhanced protection against severe accidents. Still, it is questionable whether the 'no-damage-outside-the-fence' criteria can be achieved deterministically and at competitive costs. Market deregulation and utility privatisation call for a balance between safety and costs, without jeopardising basic safety concepts. An ideal approach must be risk-based and imply modern PSAs and new methods for cost-benefit and ALARA analyses, embed nuclear risks in a wider risk spectrum, but also make benefits transparent within the context of a broader life experience. Governments should define basic requirements, minimum standards and consistent comparison criteria, and strengthen operator responsibility. Internationally sufficient and binding safety requirements must be established and nuclear technology transfer handled in a responsible way, while existing plants, with their continuous backfitting investments, should receive particular attention. (orig.)

  2. Auto Safety

    Science.gov (United States)

    ... Safe Videos for Educators Search English Español Auto Safety KidsHealth / For Parents / Auto Safety What's in this ... by teaching some basic rules. Importance of Child Safety Seats Using a child safety seat (car seat) ...

  3. Combining the IADPSG criteria with the WHO diagnostic criteria for ...

    African Journals Online (AJOL)

    Macrosomia or at least one adverse outcome were more likely in GDM patients who met the diagnostic criteria by both the IADPSG and WHO criteria (P = 0.001). Conclusion: A diagnosis of GDM that meets both the WHO and IADPSG criteria provides stronger prediction for adverse pregnancy outcome than a diagnosis that ...

  4. Siting Criteria for Low and Intermediate Level Radioactive Waste Disposal in Egypt (Proposal approach)

    International Nuclear Information System (INIS)

    Abdellatif, M.M.

    2012-01-01

    The objective of radioactive waste disposal is to isolate waste from the surrounding media so that it does not result in undue radiation exposure to humans and the environment. The required degree of isolation can be obtained by implementing various disposal methods and suitable criteria. Near surface disposal method has been practiced for some decades, with a wide variation in sites, types and amounts of wastes, and facility designs employed. Experience has shown that the effective and safe isolation of waste depends on the performance of the overall disposal system, which is formed by three major components or barriers: the site, the disposal facility and the waste form. The site selection process for low-level and intermediate level radioactive waste disposal facility addressed a wide range of public health, safety, environmental, social and economic factors. Establishing site criteria is the first step in the sitting process to identify a site that is capable of protecting public health, safety and the environment. This paper is concerning a proposal approach for the primary criteria for near surface disposal facility that could be applicable in Egypt.

  5. Design loads, loading combinations and structural acceptance criteria for BWR containments in the United States

    International Nuclear Information System (INIS)

    Edwards, N.W.

    1979-01-01

    The definition of loads, loading combinations, and structural acceptance criteria used for the design and evaluation of BWR containments in the Unites States has become much more comprehensive over the past decade. The Mark I pressure suppression containment vessels were designed for a static design pressure, a design temperature, dead load and static equivalent earthquake. The current Mark III containments are being designed to accommodate many more loads such as safety relief valve discharge loads, and suppression pool hydrodynamic loadings associated with the steam condensation phenomena as well as pressure and temperature transients for a range of pipe break sizes. Consistent with the more comprehensive definition of loads and loading combinations, the ASME Code presently establishes structural acceptance criteria with different margins of safety by the definition of Service Level Assignments A, B, C and D. Acting in a responsible manner, United States utilities are currently evaluating and modifying existing containment vessels to account for the more detailed load definition and structural acceptance criteria. (orig.)

  6. Criteria for high-level waste disposal

    International Nuclear Information System (INIS)

    Sousselier, Y.

    1981-01-01

    Disposal of radioactive wastes is storage without the intention of retrieval. But in such storage, it may be useful and in some cases necessary to have the possibility of retrieval at least for a certain period of time. In order to propose some criteria for HLW disposal, one has to examine how this basic concept is to be applied. HLW is waste separated as a raffinate in the first cycle of solvent extraction in reprocessing. Such waste contains the bulk of fission products which have long half lives, therefore the safety of a disposal site, at least after a certain period of time, must be intrinsic, i.e. not based on human intervention. There is a consensus that such a disposal is feasible in a suitable geological formation in which the integrity of the container will be reinforced by several additional barriers. Criteria for disposal can be proposed for all aspects of the question. The author discusses the aims of the safety analysis, particularly the length of time for this analysis, and the acceptable dose commitments resulting from the release of radionuclides, the number and role of each barrier, and a holistic analysis of safety external factors. (Auth.)

  7. Depression, antidepressants and driving safety.

    Science.gov (United States)

    Hill, Linda L; Lauzon, Vanessa L; Winbrock, Elise L; Li, Guohua; Chihuri, Stanford; Lee, Kelly C

    2017-12-01

    The purpose of this study was to review to review the reported associations of depression and antidepressants with motor vehicle crashes. A literature search for material published in the English language between January, 1995, and October, 2015, in bibliographic databases was combined with a search for other relevant material referenced in the retrieved articles. Retrieved articles were systematically reviewed for inclusion criteria: 19 epidemiological studies (17 case-control and 2 cohort studies) fulfilled the inclusion criteria by estimating the crash risk associated with depression and/or psychotropic medications in naturalistic settings. The estimates of the odds ratio (OR) of crash involvement associated with depression ranged from 1.78 to 3.99. All classes of antidepressants were reported to have side effects with the potential to affect driving safety. The majority of studies of antidepressant effects on driving reported an elevated crash risk, and ORs ranged from 1.19 to 2.03 for all crashes, and 3.19 for fatal crashes. In meta-analysis, depression was associated with approximately 2-fold increased crash risk (summary OR = 1.90; 95% CI, 1.06 to 3.39), and antidepressants were associated with approximately 40% increased crash risk (summary OR = 1.40; 95%CI, 1.18 to 1.66). Based on the findings of the studies reviewed, depression, antidepressants or the combination of depression and antidepressants may pose a potential hazard to driving safety. More research is needed to understand the individual contributions of depression and the medications used to treat depression.

  8. National Recommended Water Quality Criteria

    Data.gov (United States)

    U.S. Environmental Protection Agency — The National Recommended Water Quality Criteria is a compilation of national recommended water quality criteria for the protection of aquatic life and human health...

  9. Radioactive facilities classification criteria

    International Nuclear Information System (INIS)

    Briso C, H.A.; Riesle W, J.

    1992-01-01

    Appropriate classification of radioactive facilities into groups of comparable risk constitutes one of the problems faced by most Regulatory Bodies. Regarding the radiological risk, the main facts to be considered are the radioactive inventory and the processes to which these radionuclides are subjected. Normally, operations are ruled by strict safety procedures. Thus, the total activity of the radionuclides existing in a given facility is the varying feature that defines its risk. In order to rely on a quantitative criterion and, considering that the Annual Limits of Intake are widely accepted references, an index based on these limits, to support decisions related to radioactive facilities, is proposed. (author)

  10. Consideration of Criteria for a Conceptual Near Surface Radioactive Waste disposal Facility in Kenya

    Energy Technology Data Exchange (ETDEWEB)

    Nderitu, Stanley Werugia; Kim, Changlak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-05-15

    The purpose of the criteria is to limit the consequences of events which could lead to radiation exposures. This study will present an approach for establishing radiological waste acceptance criteria using a safety assessment methodology and illustrate some of its application in establishing limits on the total activity and the activity concentrations of radioactive waste to be disposed in a conceptual near surface disposal facility in Kenya. The approach will make use of accepted methods and computational schemes currently used in assessing the safety of near surface disposal facilities. The study will mainly focus on post-closure periods. The study will employ some specific inadvertent human intrusion scenarios in the development of example concentration ranges for the disposal of near-surface wastes. The overall goal of the example calculations is to illustrate the application of the scenarios in a performance assessment to assure that people in the future cannot receive a dose greater than an established limit. The specific performance assessments will use modified scenarios and data to establish acceptable disposal concentrations for specific disposal sites and conditions. Safety and environmental impacts assessments is required in the post-closure phase to support particular decisions in development, operation, and closure of a near surface repository.

  11. Diagnostic Criteria for Pediatric MS

    Directory of Open Access Journals (Sweden)

    J Gordon Millichap

    2013-06-01

    Full Text Available Investigators at Northwestern University Feinberg School of Medicine and Ann & Robert H. Lurie Children’s Hospital of Chicago review the diagnostic criteria for pediatric multiple sclerosis, the differential diagnosis, the 2010 McDonald criteria, and Callen criteria.

  12. Analysis and consideration for the US criteria of nuclear fuel cycle facilities to resist natural disasters

    International Nuclear Information System (INIS)

    Shen Hong

    2013-01-01

    Natural disasters pose a threat to the safety of nuclear facilities. Fukushima nuclear accident tells us that nuclear safety in siting, design and construction shall be strengthened in case of external events caused by natural disasters. This paper first analyzes the DOE criteria of nuclear fuel cycle facilities to resist natural disasters. Then to develop our national criteria for natural disaster resistance of nuclear fuel cycle facilities is suggested, so as to ensure the safety of these facilities. (authors)

  13. Proceedings of the workshop on structural design criteria for HTR

    International Nuclear Information System (INIS)

    Breitbach, G.; Schubert, F.; Nickel, H.

    1989-04-01

    The papers demonstrate the status of high temperature reactor technology with regard to its realization in the nuclear power industry of various countries (FRG, USA, Japan) as well as to the development of safety rules in Germany. The design criteria for HTR could be presented. The criteria already determine definitely and almost completely the relevant requirements of the component rules. The informations include the technical boundary conditions with regard to safety, the metallic high temperature components, a particular section dealing with the reactor pressure vessel, especially with the prestressed concrete vessel, and the structural graphite components. (DG)

  14. The Derivation of Evaluation Criteria of Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Kim, S. K.; Ko, W. I.

    2013-01-01

    This study suggests the evaluation criteria and evaluation indicators derived using a factor analysis. As a result of a factor analysis, 5 evaluation criteria (safety (technological feature), environmental impact, economic feasibility, sociality, institution) and 24 evaluation indicators were selected. Particularly, the level of legislation for the management of radioactive waste, the level of establishment of safety standards of the country, and the level of application of international safety standards were analyzed to be qualitative evaluation indicators that should be considered in the aspect of the institution. The purpose of an analysis on diverse nuclear fuel cycles is to select the optimum nuclear fuel cycle suitable for the environment of one's own country. Accordingly, diverse evaluation criteria and evaluation indicators are necessary. In addition, individual evaluation criteria can be explained with various evaluation indicators. For example, the evaluation criteria for economic feasibility can be explained with evaluation indicators such as the unit cost or total cost. However, if too many evaluation indicators are included in one evaluation criterion, the evaluation is not easy, and if too few evaluation indicators are established, the evaluation criteria cannot be explained sufficiently, and thus the evaluation can be distorted. Accordingly, not only should the evaluation indicators be composed of an appropriate number of units, but they should also not be overlapped, and ambiguous evaluation indicators should be dropped out and necessary evaluation indicators must be included

  15. Safety system status monitoring

    International Nuclear Information System (INIS)

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide

  16. Safety system status monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide.

  17. Performance objectives and criteria for plant evaluations

    International Nuclear Information System (INIS)

    1983-04-01

    Maintenance organization and administration should ensure effective implementation and control of maintenance activities. The criteria are: A. The organizational structure is clearly defined. B. Staffing and resources are sufficient to accomplish assigned tasks. C. Responsibilities and authority of each management, supervisory, and professional position are clearly defined. D. Personnel clearly understand their authority, responsibilities, accountabilities, and interfaces with supporting groups. E. Administrative controls are employed for maintenance activities important to plant safety and reliability. F. Performance appraisals are effectively utilized to enhance individual performance

  18. Managing for safety at nuclear installations

    International Nuclear Information System (INIS)

    1996-01-01

    This publication, by the Health and Safety Executive's (HSE's) Nuclear Safety Division (NSD), provides a statement of the criteria the Nuclear Installations Inspectorate (NII) uses to judge the adequacy of any proposed or existing system for managing a nuclear installation in so far as it affects safety. These criteria have been developed from the basic HSE model, described in the publication Successful health and safety management that applies to industry generally, in order to meet the additional needs for managing nuclear safety. In addition, the publication identifies earlier studies upon which this work was based together with the key management activities and outputs. (Author)

  19. Decision criteria in PSA applications

    International Nuclear Information System (INIS)

    Holmberg, J.E.; Pulkkinen, U.; Rosqvist, T.; Simola, K.

    2001-11-01

    Along with the adoption of risk informed decision making principles, the need for formal probabilistic decision rule or criteria has been risen. However, there are many practical and theoretical problems in the application of probabilistic criteria. One has to think what is the proper way to apply probabilistic rules together with deterministic ones and how the criteria are weighted with respect to each other. In this report, we approach the above questions from the decision theoretic point of view. We give a short review of the most well known probabilistic criteria, and discuss examples of their use. We present a decision analytic framework for evaluating the criteria, and we analyse how the different criteria behave under incompleteness or uncertainty of the PSA model. As the conclusion of our analysis we give recommendations on the application of the criteria in different decision situations. (au)

  20. Safety analysis and lay-out aspects of shieldings against particle radiation at the example of spallation facilities in the megawatt range; Sicherheitstechnische Analyse und Auslegungsaspekte von Abschirmungen gegen Teilchenstrahlung am Beispiel von Spallationsanlagen im Megawatt Bereich

    Energy Technology Data Exchange (ETDEWEB)

    Hanslik, R.

    2006-08-15

    This paper discusses the shielding of particle radiation from high current accelerators, spallation neutron sources and so called ADS-facilities (Accelerator Driven Systems). ADS-facilities are expected to gain importance in the future for transmutation of long-lived isotopes from fission reactors as well as for energy production. In this paper physical properties of the radiation as well as safety relevant requirements and corresponding shielding concepts are discussed. New concepts for the layout and design of such shielding are presented. Focal point of this work will be the fundamental difference between conventional fission reactor shielding and the safety relevant issues of shielding from high-energy radiation. Key point of this paper is the safety assessment of shielding issues of high current accelerators, spallation targets and ADS-blanket systems as well as neutron scattering instruments at spallation neutron sources. Safety relevant shielding requirements are presented and discussed. For the layout and design of the shielding for spallation sources computer base calculations methods are used. A discussion and comparison of the most important methods like semi-empirical, deterministic and stochastic codes are presented. Another key point within the presented paper is the discussion of shielding materials and their shielding efficiency concerning different types of radiation. The use of recycling material, as a cost efficient solution, is discussed. Based on the conducted analysis, flowcharts for a systematic layout and design of adequate shielding for targets and accelerators have been developed and are discussed in this paper. By use of these flowcharts layout and engineering design of future ADS-facilities can be performed. (orig.)

  1. ACR appropriateness criteria jaundice.

    Science.gov (United States)

    Lalani, Tasneem; Couto, Corey A; Rosen, Max P; Baker, Mark E; Blake, Michael A; Cash, Brooks D; Fidler, Jeff L; Greene, Frederick L; Hindman, Nicole M; Katz, Douglas S; Kaur, Harmeet; Miller, Frank H; Qayyum, Aliya; Small, William C; Sudakoff, Gary S; Yaghmai, Vahid; Yarmish, Gail M; Yee, Judy

    2013-06-01

    A fundamental consideration in the workup of a jaundiced patient is the pretest probability of mechanical obstruction. Ultrasound is the first-line modality to exclude biliary tract obstruction. When mechanical obstruction is present, additional imaging with CT or MRI can clarify etiology, define level of obstruction, stage disease, and guide intervention. When mechanical obstruction is absent, additional imaging can evaluate liver parenchyma for fat and iron deposition and help direct biopsy in cases where underlying parenchymal disease or mass is found. Imaging techniques are reviewed for the following clinical scenarios: (1) the patient with painful jaundice, (2) the patient with painless jaundice, and (3) the patient with a nonmechanical cause for jaundice. The ACR Appropriateness Criteria are evidence-based guidelines for specific clinical conditions that are reviewed every 2 years by a multidisciplinary expert panel. The guideline development and review include an extensive analysis of current medical literature from peer-reviewed journals and the application of a well-established consensus methodology (modified Delphi) to rate the appropriateness of imaging and treatment procedures by the panel. In those instances where evidence is lacking or not definitive, expert opinion may be used to recommend imaging or treatment. Copyright © 2013 American College of Radiology. Published by Elsevier Inc. All rights reserved.

  2. Criteria for decommissioning

    International Nuclear Information System (INIS)

    Ricci, P.F.

    1988-01-01

    In this paper the authors describe three risk acceptability criteria as parts of a strategy to clean up decommissioned facilities, related to both the status quo and to a variety of alternative technical clean-up options. The acceptability of risk is a consideration that must enter into any decision to establish when a site is properly decommissioned. To do so, both the corporate and public aspects of the acceptability issue must be considered. The reasons for discussion the acceptability of risk are to: Legitimize the process for making cleanup decisions; Determine who is at risk, who benefits, and who bears the costs of site cleanup, for each specific cleanup option, including the do nothing option; Establish those factors that, taken as a whole, determine measures of acceptability; Determine chemical-specific aggregate and individual risk levels; and Establish levels for cleanup. The choice of these reasons is pragmatic. The method consistent with these factors is risk-risk-effectiveness: the level of cleanup must be consistent with the foreseeable use of the site and budget constraints. Natural background contamination is the level below which further cleanup is generally inefficient. Case-by-case departures from natural background are to be considered depending on demonstrated risk. For example, a hot spot is obviously a prima facie exception, but should be rebuttable. Rebuttability means that, through consensus, the ''hot spot'' is shown not to be associated with exposure

  3. Criteria for performance evaluation

    Directory of Open Access Journals (Sweden)

    David J. Weiss

    2009-03-01

    Full Text Available Using a cognitive task (mental calculation and a perceptual-motor task (stylized golf putting, we examined differential proficiency using the CWS index and several other quantitative measures of performance. The CWS index (Weiss and Shanteau, 2003 is a coherence criterion that looks only at internal properties of the data without incorporating an external standard. In Experiment 1, college students (n = 20 carried out 2- and 3-digit addition and multiplication problems under time pressure. In Experiment 2, experienced golfers (n = 12, also college students, putted toward a target from nine different locations. Within each experiment, we analyzed the same responses using different methods. For the arithmetic tasks, accuracy information (mean absolute deviation from the correct answer, MAD using a coherence criterion was available; for golf, accuracy information using a correspondence criterion (mean deviation from the target, also MAD was available. We ranked the performances of the participants according to each measure, then compared the orders using Spearman's rextsubscript{s}. For mental calculation, the CWS order correlated moderately (rextsubscript{s} =.46 with that of MAD. However, a different coherence criterion, degree of model fit, did not correlate with either CWS or accuracy. For putting, the ranking generated by CWS correlated .68 with that generated by MAD. Consensual answers were also available for both experiments, and the rankings they generated correlated highly with those of MAD. The coherence vs. correspondence distinction did not map well onto criteria for performance evaluation.

  4. Nuclear safety

    International Nuclear Information System (INIS)

    1991-02-01

    This book reviews the accomplishments, operations, and problems faced by the defense Nuclear Facilities Safety Board. Specifically, it discusses the recommendations that the Safety Board made to improve safety and health conditions at the Department of Energy's defense nuclear facilities, problems the Safety Board has encountered in hiring technical staff, and management problems that could affect the Safety Board's independence and credibility

  5. Elements of nuclear safety

    CERN Document Server

    Libmann, Jacques

    1996-01-01

    This basically educational book is intended for all involved in nuclear facility safety. It dissects the principles and experiences conducive to the adoption of attitudes compliant with what is now known as "safety culture". This book is accessible to a wide range of readers.

  6. Communication's Role in Safety Management and Performance for the Road Safety Practices

    OpenAIRE

    Salim Keffane (s)

    2014-01-01

    Communication among organizations could play an important role in increasing road safety. To get in-depth knowledge of its role, this study measured managers' and employees' perceptions of the communication's role on six safety management and performance criteria for road safety practices by conducting a survey using a questionnaire among 165 employees and 135 managers. Path analysis using AMOS-19 software shows that some of the safety management road safety practices have high correlation wi...

  7. Drug Safety

    Science.gov (United States)

    ... over-the-counter drug. The FDA evaluates the safety of a drug by looking at Side effects ... clinical trials The FDA also monitors a drug's safety after approval. For you, drug safety means buying ...

  8. Nuclear safety

    International Nuclear Information System (INIS)

    Tarride, Bruno

    2015-10-01

    The author proposes an overview of methods and concepts used in the nuclear industry, at the design level as well as at the exploitation level, to ensure an acceptable safety level, notably in the case of nuclear reactors. He first addresses the general objectives of nuclear safety and the notion of acceptable risk: definition and organisation of nuclear safety (relationships between safety authorities and operators), notion of acceptable risk, deterministic safety approach and main safety principles (safety functions and confinement barriers, concept of defence in depth). Then, the author addresses the safety approach at the design level: studies of operational situations, studies of internal and external aggressions, safety report, design principles for important-for-safety systems (failure criterion, redundancy, failure prevention, safety classification). The next part addresses safety during exploitation and general exploitation rules: definition of the operation domain and of its limits, periodic controls and tests, management in case of incidents, accidents or aggressions

  9. Water quality criteria for lead

    Energy Technology Data Exchange (ETDEWEB)

    Nagpal, N.K.

    1987-01-01

    This report is one in a series that establishes water quality criteria for British Columbia. The report sets criteria for lead to protect a number of water uses, including drinking water, freshwater and marine aquatic life, wildlife, livestock, irrigation, and recreation. The criteria are set as either maximum concentrations of total lead that should not be exceeded at any time, or average concentrations that should not be exceeded over a 30-day period. Actual values are summarized.

  10. Criteria for Authorship in Bioethics

    OpenAIRE

    Resnik, David B.; Master, Zubin

    2011-01-01

    Multiple authorship is becoming increasingly common in bioethics research. There are well-established criteria for authorship in empirical bioethics research but not for conceptual research. It is important to develop criteria for authorship in conceptual publications to prevent undeserved authorship and uphold standards of fairness and accountability. This article explores the issue of multiple authorship in bioethics and develops criteria for determining who should be an author on a concept...

  11. 10 CFR 32.23 - Same: Safety criteria.

    Science.gov (United States)

    2010-01-01

    ... normal use and disposal of a single exempt unit, it is unlikely that the external radiation dose in any... units likely to accumulate in one location during marketing, distribution, installation, and servicing of the product, it is unlikely that the external radiation dose in any one year, or the dose...

  12. 10 CFR 32.27 - Same: Safety criteria.

    Science.gov (United States)

    2010-01-01

    ... exempt units likely to accumulate in one location during marketing, distribution, installation, and servicing of the product, it is unlikely that the external radiation dose in any one year, or the dose... marketing, distribution, installation, and servicing of the product, the probability is low that the...

  13. 32 CFR 636.33 - Vehicle safety inspection criteria.

    Science.gov (United States)

    2010-07-01

    ... a muffler in good working order and in constant operation. (10) Mirror—every vehicle, from which the... obstructed by any sign, poster, or other nontransparent material. Windshields and rear windows will not have...

  14. Preliminary Acceptance Criteria for Safety Analysis of KALIMER-600 SFR

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Chang, Won Pyo; Jeong, Hae Yong

    2010-01-01

    The KALIMER-600 event categorization in the function of occurrence frequency has been made by traditional engineering judgment with information from some reference plants such as CRBR, PRISM and EFR. The dividing line between DBE and BDBE is the frequency of 10 -7 per plant-year. Each event belongs to one of five categories based upon its nominal frequency per reactor-year (f) as a criterion. (1) Moderate frequency Event (MF): f ≥ 10 -1 (2) Infrequent Event (IE): 10 -1 > f ≥ 10 -2 (3) Unlikely Event (UE): 10 -2 > f ≥ 10 -4 (4) Extremely Unlikely Event (XU): 10 -4 > f ≥ 10 -7 (5) Beyond DBE (BDBE): > f ≥ 10 -4

  15. 76 FR 20070 - Commercial Space Transportation Safety Approval Performance Criteria

    Science.gov (United States)

    2011-04-11

    ... DEPARTMENT OF TRANSPORTATION Federal Aviation Administration Commercial Space Transportation... Commercial Space Transportation (AST), 800 Independence Avenue SW., Room 331, Washington, DC 20591, telephone.... Nield, Associate Administrator for Commercial Space Transportation. [FR Doc. 2011-8534 Filed 4-8-11; 8...

  16. Design criteria for the 218-group criticality safety reference library

    International Nuclear Information System (INIS)

    Westfall, R.M.; Ford, W.E. III; Webster, C.C.

    1978-01-01

    The generation of a 218-group neutron cross-section library from ENDF/B-IV data is described. Experience in selecting broad-group subsets and applying them in the analysis of critical experiments is related. Recommendations on the use of the 218-group library are made. 3 figures, 5 tables

  17. Safety culture

    International Nuclear Information System (INIS)

    Keen, L.J.

    2003-01-01

    Safety culture has become a topic of increasing interest for industry and regulators as issues are raised on safety problems around the world. The keys to safety culture are organizational effectiveness, effective communications, organizational learning, and a culture that encourages the identification and resolution of safety issues. The necessity of a strong safety culture places an onus on all of us to continually question whether the safety measures already in place are sufficient, and are being applied. (author)

  18. Combining the IADPSG criteria with the WHO diagnostic criteria for ...

    African Journals Online (AJOL)

    The International Association of Diabetes in Pregnancy Study Group (IADPSG) and World Health ... Macrosomia or at least one adverse outcome were more likely in GDM patients who ... criteria for GDM in the ADA's more recent position statement.[18] .... at risk for postpartum type 2 DM;[27] the IADPSG criteria on the other ...

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

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

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

  20. Radwaste characteristics and Disposal Facility Waste Acceptance Criteria

    International Nuclear Information System (INIS)

    Sung, Suk Hyun; Jeong, Yi Yeong; Kim, Ki Hong

    2008-01-01

    The purpose of Radioactive Waste Acceptance Criteria (WAC) is to verify a radioactive waste compliance with radioactive disposal facility requirements in order to maintain a disposal facility's performance objectives and to ensure its safety. To develop WAC which is conformable with domestic disposal site conditions, we furthermore analysed the WAC of foreign disposal sites similar to the Kyung-Ju disposal site and the characteristics of various wastes which are being generated from Korea nuclear facilities. Radioactive WAC was developed in the technical cooperation with the Korea Atomic Energy Research Institute in consideration of characteristics of the wastes which are being generated from various facilities, waste generators' opinions and other conditions. The established criteria was also discussed and verified at an advisory committee which was comprised of some experts from universities, institutes and the industry. So radioactive WAC was developed to accept all wastes which are being generated from various nuclear facilities as much as possible, ensuring the safety of a disposal facility. But this developed waste acceptance criteria is not a criteria to accept all the present wastes generated from various nuclear facilities, so waste generators must seek an alternative treatment method for wastes which were not worth disposing of, and then they must treat the wastes more to be acceptable at a disposal site. The radioactive disposal facility WAC will continuously complement certain criteria related to a disposal concentration limit for individual radionuclide in order to ensure a long-term safety.

  1. Reload core safety verification

    International Nuclear Information System (INIS)

    Svetlik, M.; Minarcin, M.

    2003-01-01

    This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

  2. Regulatory criteria for final disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Petraitis, E.; Ciallella, N.; Siraky, G.

    1998-01-01

    This paper describes briefly the legislative and regulatory framework in which the final disposal of radioactive wastes is carried out in Argentina. It also presents the criteria developed by the Nuclear Regulatory Authority (ARN) to assess the long-term safety of final disposal systems for high level radioactive wastes. (author)

  3. 47 CFR 101.105 - Interference protection criteria.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 5 2010-10-01 2010-10-01 false Interference protection criteria. 101.105 Section 101.105 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO.... (ii) To accommodate co-primary Direct Broadcast Satellite Service earth stations, an MVDDS...

  4. 78 FR 79010 - Criteria to Certify Coal Mine Rescue Teams

    Science.gov (United States)

    2013-12-27

    ... coal requires more heat to combust; (3) anthracite dust does not propagate an explosion; and (4) there... to Certify Coal Mine Rescue Teams AGENCY: Mine Safety and Health Administration, Labor. ACTION... updated the coal mine rescue team certification criteria. The Mine Improvement and New Emergency Response...

  5. Looking for Improvement in Last Planner System: Defining Selection Criteria

    DEFF Research Database (Denmark)

    Lindhard, Søren; Wandahl, Søren

    2013-01-01

    criteria was carried out. Six flows are identified as relevant: workforce, material, and machinery which comprise the needed resources and safety, climate conditions, and space which affect the pace of the work. Because of the importance to progress in the workflow, and the on-schedule completeness...

  6. At-reactor storage concepts criteria for preliminary assessment

    International Nuclear Information System (INIS)

    Boydston, L.A.

    1981-12-01

    The licensing, safety, and environmental considerations of four wet and four dry at-reactor storage concepts are presented. Physical criteria for each concept are examined to determine the minimum site and facility requirements which must be met by a utility which desires to expand its at-reactor spent fuel storage capability

  7. Ecological radiation protection criteria for nuclear power

    International Nuclear Information System (INIS)

    Kryshev, I.I.

    1993-01-01

    By now a large quantity of radioactive hazards of all sizes and shapes has accumulated in Russia. They include RBMK, VVER, and BN (fast-neutron) nuclear power plants, nuclear fuel processing plants, radioactive waste dumps, ships with nuclear power units, etc. In order to evaluate the radioecological situation correctly, the characteristics of the radioactive contamination must be compiled in these areas with some system of criteria which will provide an acceptable level of ecological safety. Currently health criteria for radiation protection are, which are oriented to man's radiation protection, predominate. Here the concept of a thresholdless linear dose-response dependence, which has been confirmed experimentally only at rather high doses (above 1 Gy), is taken as the theoretical basis for evaluating and normalizing radiation effects. According to one opinion, protecting people against radiation is sufficient to protect other types of organisms, although they are not necessarily of the same species. However, from the viewpoint of ecology, this approach is incorrect, because it does not consider radiation dose differences between man and other living organisms. The article discusses dose-response dependences for various organisms, biological effects of ionizing radiation, and appropriate radiation protection criteria

  8. Inappropriate prescribing: criteria, detection and prevention.

    LENUS (Irish Health Repository)

    O'Connor, Marie N

    2012-06-01

    Inappropriate prescribing is highly prevalent in older people and is a major healthcare concern because of its association with negative healthcare outcomes including adverse drug events, related morbidity and hospitalization. With changing population demographics resulting in increasing proportions of older people worldwide, improving the quality and safety of prescribing in older people poses a global challenge. To date a number of different strategies have been used to identify potentially inappropriate prescribing in older people. Over the last two decades, a number of criteria have been published to assist prescribers in detecting inappropriate prescribing, the majority of which have been explicit sets of criteria, though some are implicit. The majority of these prescribing indicators pertain to overprescribing and misprescribing, with only a minority focussing on the underprescribing of indicated medicines. Additional interventions to optimize prescribing in older people include comprehensive geriatric assessment, clinical pharmacist review, and education of prescribers as well as computerized prescribing with clinical decision support systems. In this review, we describe the inappropriate prescribing detection tools or criteria most frequently cited in the literature and examine their role in preventing inappropriate prescribing and other related healthcare outcomes. We also discuss other measures commonly used in the detection and prevention of inappropriate prescribing in older people and the evidence supporting their use and their application in everyday clinical practice.

  9. mathematical models for prediction of safety factors for a simply

    African Journals Online (AJOL)

    HOD

    Keywords: reliability, code calibration, load factor, safety factor, design, steel beam. 1. INTRODUCTION ... safety factors for the design of a simply supported steel beam using regression .... 5 design criteria for a solid timber portal frame.

  10. Code on the safety of nuclear power plants: Siting

    International Nuclear Information System (INIS)

    1988-01-01

    This Code provides criteria and procedures that are recommended for safety in nuclear power plant siting. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  11. Clinical Criteria for Physician Aid in Dying.

    Science.gov (United States)

    Orentlicher, David; Pope, Thaddeus Mason; Rich, Ben A

    2016-03-01

    More than 20 years ago, even before voters in Oregon had enacted the first aid in dying (AID) statute in the United States, Timothy Quill and colleagues proposed clinical criteria AID. Their proposal was carefully considered and temperate, but there were little data on the practice of AID at the time. (With AID, a physician writes a prescription for life-ending medication for a terminally ill, mentally capacitated adult.) With the passage of time, a substantial body of data on AID has developed from the states of Oregon and Washington. For more than 17 years, physicians in Oregon have been authorized to provide a prescription for AID. Accordingly, we have updated the clinical criteria of Quill, et al., based on the many years of experience with AID. With more jurisdictions authorizing AID, it is critical that physicians can turn to reliable clinical criteria. As with any medical practice, AID must be provided in a safe and effective manner. Physicians need to know (1) how to respond to a patient's inquiry about AID, (2) how to assess patient decision making capacity, and (3) how to address a range of other issues that may arise. To ensure that physicians have the guidance they need, Compassion & Choices convened the Physician Aid-in-Dying Clinical Criteria Committee, in July 2012, to create clinical criteria for physicians who are willing to provide AID to patients who request it. The committee includes experts in medicine, law, bioethics, hospice, nursing, social work, and pharmacy. Using an iterative consensus process, the Committee drafted the criteria over a one-year period.

  12. Acceptability criteria for final underground disposal of radioactive waste

    International Nuclear Information System (INIS)

    Sousselier, Y.

    1984-01-01

    Specialists now generally agree that the underground disposal of suitably immobilized radioactive waste offers a means of attaining the basic objective of ensuring the immediate and long-term protection of man and the environment throughout the requisite period of time and in all foreseeable circumstances. Criteria of a more general as well as a more specific nature are practical means through which this basic protection objective can be reached. These criteria, which need not necessarily be quantified, enable the authorities to gauge the acceptability of a given project and provide those responsible for waste management with a basis for making decisions. In short, these principles constitute the framework of a suitably safety-oriented waste management policy. The more general criteria correspond to the protection objectives established by the national authorities on the basis of principles and recommendations formulated by international organizations, in particular the ICRP and the IAEA. They apply to any underground disposal system considered as a whole. The more specific criteria provide a means of evaluating the degree to which the various components of the disposal system meet the general criteria. They must also take account of the interaction between these components. As the ultimate aim is the overall safety of the disposal system, individual components can be adjusted to compensate for the performance of others with respect to the criteria. This is the approach adopted by the international bodies and national authorities in developing acceptability criteria for the final underground radioactive disposal systems to be used during the operational and post-operational phases respectively. The main criteria are reviewed and an attempt is made to assess the importance of the specific criteria according to the different types of disposal systems. (author)

  13. Efficacy and Safety of Sarecycline, a Novel, Once-Daily, Narrow Spectrum Antibiotic for the Treatment of Moderate to Severe Facial Acne Vulgaris: Results of a Phase 2, Dose-Ranging Study.

    Science.gov (United States)

    Leyden, James J; Sniukiene, Vilma; Berk, David R; Kaoukhov, Alexandre

    2018-03-01

    There is a need for new oral antibiotics for acne with improved safety profiles and targeted antibacterial spectra. Sarecycline is a novel, tetracycline-class antibiotic specifically designed for acne, offering a narrow spectrum of activity compared with currently available tetracyclines, including less activity against enteric Gram-negative bacteria. This phase 2 study evaluated the efficacy and safety of three doses of sarecycline for moderate to severe facial acne vulgaris. In this multicenter, double-blind, placebo-controlled study, patients aged 12 to 45 years were randomized to once-daily sarecycline 0.75 mg/kg, 1.5 mg/kg, 3.0 mg/kg, or placebo. Efficacy analyses included change from baseline in inflammatory and noninflammatory lesion counts at week 12, with between-group comparisons using analysis of covariance. Safety assessments included adverse events (AEs), clinical laboratories, vital signs, electrocardiograms, and physical examinations. Overall, 285 randomized patients received at least one dose of study drug. At week 12, sarecycline 1.5 mg/kg and 3.0 mg/kg groups demonstrated significantly reduced inflammatory lesions from baseline (52.7% and 51.8%, respectively) versus placebo (38.3%; P=0.02 and P=0.03, respectively). Sarecycline was safe and well tolerated, with similar gastrointestinal AE rates in sarecycline and placebo groups. Vertigo and photosensitivity AEs occurred in less than 1% of patients when pooling sarecycline groups; no vulvovaginal candidiasis AEs occurred. Discontinuation rates due to AEs were low. No serious AEs occurred. Once-daily sarecycline 1.5 mg/kg significantly reduced inflammatory lesions versus placebo and was safe and well tolerated with low rates of AEs, including gastrointestinal AEs. Sarecycline 3.0 mg/kg did not result in additional efficacy versus 1.5 mg/kg. Sarecycline may represent a novel, once-daily treatment for patients with moderate to severe acne. It offers a narrow antibacterial spectrum relative to other

  14. Safety detaching hook specification.

    CSIR Research Space (South Africa)

    Roux, JD

    1999-05-01

    Full Text Available hydraulic tensioning system The impactor must subject the safety detaching hook to an impact energy of 150 kJ minimum. A high speed digital imaging system, capable of recording at least 2000 full frames per second, shall be used to record video images... document compiled by the Contractor, detailing all major events in the production phase, including but not necessarily limited to inspection requirements, test procedures and acceptance/rejection criteria, sampling plans and equipment to be employed...

  15. Diagnostic criteria for vascular dementia

    NARCIS (Netherlands)

    Scheltens, P.; Hijdra, A. H.

    1998-01-01

    The term vascular dementia implies the presence of a clinical syndrome (dementia) caused by, or at least assumed to be caused by, a specific disorder (cerebrovascular disease). In this review, the various sets of criteria used to define vascular dementia are outlined. The various sets of criteria

  16. Safety case plan 2008

    International Nuclear Information System (INIS)

    2008-07-01

    Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy) Posiva is preparing to submit the construction license application for a spent fuel repository by the end of the year 2012. The long-term safety section supporting the license application is based on a safety case, which, according to the internationally adopted definition, is a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. In 2005, Posiva presented a plan to prepare such a safety case. The present report provides a revised plan of the safety case contents mentioned above. The update of the safety case plan takes into account the recommendations made by the Radiation and Nuclear Safety Authority (STUK) about improving the focus and further developing the plan. Accordingly, particular attention is given to the quality management of the safety case work, the management of uncertainties and the scenario methodology. The quality management is based on the ISO 9001:2000 standard process thinking enhanced with special features arising from STUK's YVL Guides. The safety case production process is divided into four main sub-processes. The conceptualisation and methodology sub-process defines the framework for the assessment. The critical data handling and modelling sub-process links Posiva's main technical and scientific activities to the production of the safety case. The assessment sub-process analyses the consequences of the evolution of the disposal system in various scenarios, classified either as part of the expected evolution or as disruptive scenarios. The compliance and confidence sub-process is responsible for final evaluation of compliance of the assessment results with the regulatory criteria and the overall confidence in the safety case. As in the previous safety case plan, the safety case will be based on several reports, but

  17. The safety features of an integrated maritime reactor

    International Nuclear Information System (INIS)

    Miyakoshi, Junichi; Yamada, Nobuyuki; Kuwahara, Shin-ichi

    1975-01-01

    The EFDR-80, a typical integrated maritime reactor, which is being developed in West Germany is outlined. The safety features of the integrated maritime reactor are presented with the analysis of reactor accidents and hazards, and are compared with those of the separated maritime reactor. Furthermore, the safety criteria of maritime reactors in Japan and West Germany are compared, and some of the differences are presented from the viewpoint of reactor design and safety analysis. In this report the authors express an earnest desire that the definite and reasonable safety criteria of the integrated maritime reactor should be established and that the safety criteria of the nuclear ship should be standardized internationally. (auth.)

  18. Failure Criteria for Reinforced Materials

    DEFF Research Database (Denmark)

    Rathkjen, Arne

    Failure of materials is often characterized as ductile yielding, brittle fracture, creep rupture, etc., and different criteria given in terms of different parameters have been used to describe different types of failure. Only criteria expressing failure in terms of stress are considered in what...... place until the matrix, the continuous component of the composite, fails. When an isotropic matrix is reinforced as described above, the result is an anisotropic composite material. Even if the material is anisotropic, it usually exhibits a rather high degree of symmetry and such symmetries place...... certain restrictions on the form of the failure criteria for anisotropic materials. In section 2, some failure criteria for homogenous materials are reviewed. Both isotropic and anisotropic materials are described, and in particular the constraints imposed on the criteria from the symmetries orthotropy...

  19. NWTS program criteria for mined geologic disposal of nuclear waste: program objectives, functional requirements, and system performance criteria

    International Nuclear Information System (INIS)

    1982-03-01

    The NWTS-33 series, of which this document is a part, provides guidance for the National Waste Terminal Storage (NWTS) program in the development and implementation of licensed mined geologic disposal systems for solidified high-level and TRU wastes. Program objectives, functional requirements, and system performance criteria are found in this document. At the present time final criteria have not been issued by the Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA). The criteria in these documents have been developed on the basis of DOE's judgment of what is required to protect the health and safety of the public and the quality of the environment. It is expected that these criteria will be consistent with regulatory standards. The criteria will be re-evaluated on a periodic basis to ensure that they remain consistent with national waste management policy and regulatory requirements. A re-evaluation will be made when final criteria are promulgated by the NRC and EPA. A background section that briefly describes the mined geologic disposal system and explains the hierarchy and application of the NWTS criteria is included in Section 2.0. Secton 3.0 presents the program objectives, Section 4.0 functional requirements, Secton 5.0 the system performance criteria, and Section 6.0 quality assurance and standards. A draft of this document was issued for public comment in April 1981. Appendix A contains the DOE responses to the comments received. Appendix B is a glossary

  20. Nuclear power plant safety

    International Nuclear Information System (INIS)

    Otway, H.J.

    1974-01-01

    Action at the international level will assume greater importance as the number of nuclear power plants increases, especially in the more densely populated parts of the world. Predictions of growth made prior to October 1973 [9] indicated that, by 1980, 14% of the electricity would be supplied by nuclear plants and by the year 2000 this figure would be about 50%. This will make the topic of international co-operation and standards of even greater importance. The IAEA has long been active in providing assistance to Member States in the siting design and operation of nuclear reactors. These activities have been pursued through advisory missions, the publication of codes of practice, guide books, technical reports and in arranging meetings to promote information exchange. During the early development of nuclear power, there was no well-established body of experience which would allow formulation of internationally acceptable safety criteria, except in a few special cases. Hence, nuclear power plant safety and reliability matters often received an ad hoc approach which necessarily entailed a lack of consistency in the criteria used and in the levels of safety required. It is clear that the continuation of an ad hoc approach to safety will prove inadequate in the context of a world-wide nuclear power industry, and the international trade which this implies. As in several other fields, the establishment of internationally acceptable safety standards and appropriate guides for use by regulatory bodies, utilities, designers and constructors, is becoming a necessity. The IAEA is presently planning the development of a comprehensive set of basic requirements for nuclear power plant safety, and the associated reliability requirements, which would be internationally acceptable, and could serve as a standard frame of reference for nuclear plant safety and reliability analyses

  1. Power plants and safety 1982

    International Nuclear Information System (INIS)

    1982-01-01

    The papers of this volume deal with the whole range of safety issues from planning and construction to the operation of power plants, and discuss also issues like availability and safety of power plants, protective clothes and their incommodating effect, alternatives for rendering hot-water generators safe and the safety philosophy in steam turbine engineering. (HAG) [de

  2. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  3. Geoscientific evaluation factors and criteria for siting and site evaluation. Progress report

    International Nuclear Information System (INIS)

    Stroem, A.; Ericsson, Lars O.; Svemar, C.; Almen, K.E.; Andersson, Johan

    1999-03-01

    thereby involved. Requirements and preferences regarding the deep repository, and thereby the rock, are primarily formulated with respect to function and not directly for individual parameter values. In a similar manner the evaluation factors have been arranged per geoscientific discipline. A geoscientific parameter that can be measured or estimated in site investigations is considered to be a suitable evaluation factor if one of the following conditions is fulfilled: a direct requirement or an essential preference has been formulated for the parameter, or a the parameter is expected to have a great influence on the result of one or more important function analyses. Based on a preliminary list of possible evaluation factors, the level of knowledge that can or should be reached after the feasibility study, site investigation and detailed characterization have been completed is also discussed. It is not reasonable to designate a geoscientific parameter as an evaluation factor if the parameter cannot be measured or estimated with sufficient accuracy. Criteria for site evaluation will also be determined in the future work. When it comes to repository performance, criteria consist of indicative values or value ranges of outcomes of performance assessments. The criteria can be changed during the course of the siting work as the information available on the sites changes. But requirements and preferences remain the same. Even though the overall evaluation of the suitability of the sites is determined within the framework of an integrated safety assessment and an integrated construction analysis, the specified criteria should provide good guidance regarding the results of such an integrated assessment/analysis

  4. Geoscientific evaluation factors and criteria for siting and site evaluation. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Stroem, A.; Ericsson, Lars O.; Svemar, C. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Almen, K.E. [KEA GEO-konsult AB (Sweden); Andersson, Johan [Golder Grundteknik KB (Sweden)

    1999-03-01

    thereby involved. Requirements and preferences regarding the deep repository, and thereby the rock, are primarily formulated with respect to function and not directly for individual parameter values. In a similar manner the evaluation factors have been arranged per geoscientific discipline. A geoscientific parameter that can be measured or estimated in site investigations is considered to be a suitable evaluation factor if one of the following conditions is fulfilled: a direct requirement or an essential preference has been formulated for the parameter, or a the parameter is expected to have a great influence on the result of one or more important function analyses. Based on a preliminary list of possible evaluation factors, the level of knowledge that can or should be reached after the feasibility study, site investigation and detailed characterization have been completed is also discussed. It is not reasonable to designate a geoscientific parameter as an evaluation factor if the parameter cannot be measured or estimated with sufficient accuracy. Criteria for site evaluation will also be determined in the future work. When it comes to repository performance, criteria consist of indicative values or value ranges of outcomes of performance assessments. The criteria can be changed during the course of the siting work as the information available on the sites changes. But requirements and preferences remain the same. Even though the overall evaluation of the suitability of the sites is determined within the framework of an integrated safety assessment and an integrated construction analysis, the specified criteria should provide good guidance regarding the results of such an integrated assessment/analysis 14 refs, figs, tabs

  5. MULTIPLE CRITERIA DECISION MAKING IN STRATEGIC PLANNING OF TABLE EGG PRODUCTION

    Directory of Open Access Journals (Sweden)

    Ana Crnčan

    2016-06-01

    Full Text Available The main research objective was to analyze and evaluate different systems of table egg production by using the multiple criteria analysis, the method of Analytic Hierarchy Process (AHP in decision making within strategic planning of production. The survey involved 79 producers of table eggs registered in the Records on laying hens’ farms in the Republic of Croatia. In the first stage, the research defined the criteria and sub-criteria for system evaluation which were compared in pairs in order to determine the weight or importance for each of them. Alternatives were evaluation based on definition of priorities of examinees and the extent to which they meet each of the defined criteria and sub-criteria. Intensity of examinees’ preferences were entered into the Expert Choice software in order to evaluate ranking results of egg production systems. Defined model consisted of a quantitative criterion of economic indicators, and the other two referred to qualitative criteria, market indicators and technical-technological factors. Each criterion had its corresponding sub-criteria that were evenly distributed in numerical order. Based on individual assessments of the examinees, overall cumulative evaluation was obtained for the table egg production systems. Accordingly, the most acceptable alternative to egg production is the indoor keeping system (priority 0.301. It is followed by the free-range system of keeping laying hens (priority 0.253. The third-ranked alternative is egg production by hens kept in conventional cages (priority 0.226, while the fourth-ranked least acceptable alternative, as of the total evaluation, is the ecological system of egg production (priority 0.220. Taking into account the obtained results of multiple criteria evaluation as well as EU and world trends in changing consumers’ habits including food safety and quality as well as customers’ preferences towards local market and local products, it is recommended that eggs

  6. Application of leak-before-break criteria to pressurized water reactors

    International Nuclear Information System (INIS)

    Roege, P.; Day, B.; Beckjord, E.; Golay, M.

    1986-01-01

    The possibility of consequential damage to safety-related systems or components after postulated pipe breaks in Light Water Reactors has led to the installation of pipe restraints capable of withstanding the loads in such an accident. These restraints are a significant part of initial capital cost, and because of their size and location, impede plant maintenance. The Piping Review Committee of the U.S. Nuclear Regulatory Commission has concluded that, subject to fulfillment of certain criteria, the pipe restraints for pressurized water reactor main coolant piping are not necessary, because the failure mode of this piping is such that it will leak before it will break, and the leakage of reactor coolant is large enough to detect. In this study, we examine the piping systems of a 4-loop 1,150 MWe pressurized water reactor, determining the crack size that would be stable from a fracture mechanics point of view, and the range of leak rates that would ensue. We then consider the sensitivity of conventional leak detection systems, and find that pipe sizes down to 45 cm in diameter would meet the leak-before-break criteria. Improvements in the sensitivity of conventional leak detectors would extend this range down to pipe sizes down to the range of 20 - 45 cm in diameter. The development of local leak detection systems which would respond to leaks in compartments or confined areas would make it possible to apply the criteria to sizes as low as 10 - 20 cm in diameter, which appear to be the limit of the net cost savings of eliminating pipe restraints and adding additional leak detection instrumentation. Extending the leak-before-break concept into this smallest pipe range may require improved precision in crack definition, flow modeling, and leak detection. Better detection of leaks may also require use of new detection methods coupled to novel approaches to piping system design. (J.P.N.)

  7. Efficacy and safety of dupilumab in adults with moderate-to-severe atopic dermatitis inadequately controlled by topical treatments: a randomised, placebo-controlled, dose-ranging phase 2b trial.

    Science.gov (United States)

    Thaçi, Diamant; Simpson, Eric L; Beck, Lisa A; Bieber, Thomas; Blauvelt, Andrew; Papp, Kim; Soong, Weily; Worm, Margitta; Szepietowski, Jacek C; Sofen, Howard; Kawashima, Makoto; Wu, Richard; Weinstein, Steven P; Graham, Neil M H; Pirozzi, Gianluca; Teper, Ariel; Sutherland, E Rand; Mastey, Vera; Stahl, Neil; Yancopoulos, George D; Ardeleanu, Marius

    2016-01-02

    Data from early-stage studies suggested that interleukin (IL)-4 and IL-13 are requisite drivers of atopic dermatitis, evidenced by marked improvement after treatment with dupilumab, a fully-human monoclonal antibody that blocks both pathways. We aimed to assess the efficacy and safety of several dose regimens of dupilumab in adults with moderate-to-severe atopic dermatitis inadequately controlled by topical treatments. In this randomised, placebo-controlled, double-blind study, we enrolled patients aged 18 years or older who had an Eczema Area and Severity Index (EASI) score of 12 or higher at screening (≥16 at baseline) and inadequate response to topical treatments from 91 study centres, including hospitals, clinics, and academic institutions, in Canada, Czech Republic, Germany, Hungary, Japan, Poland, and the USA. Patients were randomly assigned (1:1:1:1:1:1), stratified by severity (moderate or severe, as assessed by Investigator's Global Assessment) and region (Japan vs rest of world) to receive subcutaneous dupilumab: 300 mg once a week, 300 mg every 2 weeks, 200 mg every 2 weeks, 300 mg every 4 weeks, 100 mg every 4 weeks, or placebo once a week for 16 weeks. We used a central randomisation scheme, provided by an interactive voice response system. Drug kits were coded, providing masking to treatment assignment, and allocation was concealed. Patients on treatment every 2 weeks and every 4 weeks received volume-matched placebo every week when dupilumab was not given to ensure double blinding. The primary outcome was efficacy of dupilumab dose regimens based on EASI score least-squares mean percentage change (SE) from baseline to week 16. Analyses included all randomly assigned patients who received one or more doses of study drug. This trial is registered with ClinicalTrials.gov, number NCT01859988. Between May 15, 2013, and Jan 27, 2014, 452 patients were assessed for eligibility, and 380 patients were randomly assigned. 379 patients received one or more

  8. Analysing and Comparing Encodability Criteria

    Directory of Open Access Journals (Sweden)

    Kirstin Peters

    2015-08-01

    Full Text Available Encodings or the proof of their absence are the main way to compare process calculi. To analyse the quality of encodings and to rule out trivial or meaningless encodings, they are augmented with quality criteria. There exists a bunch of different criteria and different variants of criteria in order to reason in different settings. This leads to incomparable results. Moreover it is not always clear whether the criteria used to obtain a result in a particular setting do indeed fit to this setting. We show how to formally reason about and compare encodability criteria by mapping them on requirements on a relation between source and target terms that is induced by the encoding function. In particular we analyse the common criteria full abstraction, operational correspondence, divergence reflection, success sensitiveness, and respect of barbs; e.g. we analyse the exact nature of the simulation relation (coupled simulation versus bisimulation that is induced by different variants of operational correspondence. This way we reduce the problem of analysing or comparing encodability criteria to the better understood problem of comparing relations on processes.

  9. Defining and Measuring Safety Climate: A Review of the Construction Industry Literature.

    Science.gov (United States)

    Schwatka, Natalie V; Hecker, Steven; Goldenhar, Linda M

    2016-06-01

    Safety climate measurements can be used to proactively assess an organization's effectiveness in identifying and remediating work-related hazards, thereby reducing or preventing work-related ill health and injury. This review article focuses on construction-specific articles that developed and/or measured safety climate, assessed safety climate's relationship with other safety and health performance indicators, and/or used safety climate measures to evaluate interventions targeting one or more indicators of safety climate. Fifty-six articles met our inclusion criteria, 80% of which were published after 2008. Our findings demonstrate that researchers commonly defined safety climate as perception based, but the object of those perceptions varies widely. Within the wide range of indicators used to measure safety climate, safety policies, procedures, and practices were the most common, followed by general management commitment to safety. The most frequently used indicators should and do reflect that the prevention of work-related ill health and injury depends on both organizational and employee actions. Safety climate scores were commonly compared between groups (e.g. management and workers, different trades), and often correlated with subjective measures of safety behavior rather than measures of ill health or objective safety and health outcomes. Despite the observed limitations of current research, safety climate has been promised as a useful feature of research and practice activities to prevent work-related ill health and injury. Safety climate survey data can reveal gaps between management and employee perceptions, or between espoused and enacted policies, and trigger communication and action to narrow those gaps. The validation of safety climate with safety and health performance data offers the potential for using safety climate measures as a leading indicator of performance. We discuss these findings in relation to the related concept of safety culture and

  10. NWTS program criteria for mined geologic disposal of nuclear waste: program objectives, functional requirements, and system performance criteria

    International Nuclear Information System (INIS)

    1981-04-01

    At the present time, final repository criteria have not been issued by the responsible agencies. This document describes general objectives, requirements, and criteria that the DOE intends to apply in the interim to the National Waste Terminal Storage (NWTS) Program. These objectives, requirements, and criteria have been developed on the basis of DOE's analysis of what is needed to achieve the National objective of safe waste disposal in an environmentally acceptable and economic manner and are expected to be consistent with anticipated regulatory standards. The qualitative statements in this document address the broad issues of public and occupational health and safety, institutional acceptability, engineering feasibility, and economic considerations. A comprehensive set of criteria, general and project specific, of which these are a part, will constitute a portion of the technical basis for preparation and submittal by the DOE of formal documents to support future license applications for nuclear waste repositories

  11. NWTS program criteria for mined geologic disposal of nuclear waste: program objectives, functional requirements, and system performance criteria

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-04-01

    At the present time, final repository criteria have not been issued by the responsible agencies. This document describes general objectives, requirements, and criteria that the DOE intends to apply in the interim to the National Waste Terminal Storage (NWTS) Program. These objectives, requirements, and criteria have been developed on the basis of DOE's analysis of what is needed to achieve the National objective of safe waste disposal in an environmentally acceptable and economic manner and are expected to be consistent with anticipated regulatory standards. The qualitative statements in this document address the broad issues of public and occupational health and safety, institutional acceptability, engineering feasibility, and economic considerations. A comprehensive set of criteria, general and project specific, of which these are a part, will constitute a portion of the technical basis for preparation and submittal by the DOE of formal documents to support future license applications for nuclear waste repositories.

  12. Public safety around dams

    Energy Technology Data Exchange (ETDEWEB)

    Bourassa, H [Centre d' expertise hydrique du Quebec, Quebec, PQ (Canada)

    2009-07-01

    Fourty public dams are managed on a real-time basis by the Centre d'expertise hydrique du Quebec (CEHQ). This presentation described the public dams owned by the CEHQ and discussed the public safety measures at the dams. The dams serve various purposes, including protection against floods; industrial or drinking water supply; resort or recreational activities; hydroelectric development; and wildlife conservation. Trigger events were also discussed, such as the complaint at Rapides-des-Cedres dam and deaths that occurred in 2004 when water from a dam was released without warning. Several photographs were presented to illustrate that people were unaware of the danger. Initiatives aimed at raising awareness and studying public safety issues were discussed. A pilot project was launched and a permanent committee was created to evaluate all aspects of public safety at the dams owned by CEHQ. The first tasks of the committee were to establish requirements for waterway safety barriers, both upstream and downstream, for all public dams; to establish requirements for safety signage for all public dams; and to develop criteria to decide on safety signage at each dam. figs.

  13. Public safety around dams

    Energy Technology Data Exchange (ETDEWEB)

    Bourassa, H. [Centre d' expertise hydrique du Quebec, Quebec, PQ (Canada)

    2009-07-01

    Fourty public dams are managed on a real-time basis by the Centre d'expertise hydrique du Quebec (CEHQ). This presentation described the public dams owned by the CEHQ and discussed the public safety measures at the dams. The dams serve various purposes, including protection against floods; industrial or drinking water supply; resort or recreational activities; hydroelectric development; and wildlife conservation. Trigger events were also discussed, such as the complaint at Rapides-des-Cedres dam and deaths that occurred in 2004 when water from a dam was released without warning. Several photographs were presented to illustrate that people were unaware of the danger. Initiatives aimed at raising awareness and studying public safety issues were discussed. A pilot project was launched and a permanent committee was created to evaluate all aspects of public safety at the dams owned by CEHQ. The first tasks of the committee were to establish requirements for waterway safety barriers, both upstream and downstream, for all public dams; to establish requirements for safety signage for all public dams; and to develop criteria to decide on safety signage at each dam. figs.

  14. Phase 2, Randomized, Double-Blind, Dose-Ranging Study Evaluating the Safety, Tolerability, Population Pharmacokinetics, and Efficacy of Oral Torezolid Phosphate in Patients with Complicated Skin and Skin Structure Infections▿ † ‡

    Science.gov (United States)

    Prokocimer, P.; Bien, P.; Surber, J.; Mehra, P.; DeAnda, C.; Bulitta, J. B.; Corey, G. R.

    2011-01-01

    Torezolid (TR-700) is the active moiety of the prodrug torezolid phosphate ([TP] TR-701), a second-generation oxazolidinone with 4- to 16-fold greater potency than linezolid against Gram-positive species including methicillin-resistant Staphylococcus aureus (MRSA). A double-blind phase 2 study evaluated three levels (200, 300, or 400 mg) of oral, once-daily TP over 5 to 7 days for complicated skin and skin structure infections (cSSSI). Patients 18 to 75 years old with cSSSI caused by suspected or confirmed Gram-positive pathogens were randomized 1:1:1. Of 188 treated patients, 76.6% had abscesses, 17.6% had extensive cellulitis, and 5.9% had wound infections. S. aureus, the most common pathogen, was isolated in 90.3% of patients (139/154) with a baseline pathogen; 80.6% were MRSA. Cure rates in clinically evaluable patients were 98.2% at 200 mg, 94.4% at 300 mg, and 94.4% at 400 mg. Cure rates were consistent across diagnoses, regardless of lesion size or the presence of systemic signs of infection. Clinical cure rates in patients with S. aureus isolated at baseline were 96.6% overall and 96.8% for MRSA. TP was safe and well tolerated at all dose levels. No patients discontinued treatment due to an adverse event. Three-stage hierarchical population pharmacokinetic modeling yielded a geometric mean clearance of 8.28 liters/h (between-patient variability, 32.3%), a volume of the central compartment of 71.4 liters (24.0%), and a volume of the peripheral compartment of 27.9 liters (35.7%). Results of this study show a high degree of efficacy at all three dose levels without significant differences in the safety profile and support the continued evaluation of TP for the treatment of cSSSI in phase 3 trials. PMID:21115795

  15. Aquatic Life Criteria - Tributyltin (TBT)

    Science.gov (United States)

    Documents pertaining to 2004 Final Acute and Chronic Ambient Aquatic Life Water Quality Criteria for Tributyltin (TBT) for freshwater and saltwater. These documents include the safe levels of TBT that should protect the majority of species.

  16. Floorball game skills (evaluation criteria)

    OpenAIRE

    Chlumský, Marek

    2013-01-01

    Title: Playing skills in floorball (evaluation criteria). Target: To create a list of playing skills which an ideal player should demonstrate. Find and verify the evaluation criteria of these skills and inspire trainers to develop these skills in the best way. Methods: Informal interviews, individually structured interviews, analysis and verification of data, pilot testing. Results: Defined playing skills in floorball, developed scale of values of floorball playing skills, creation of exercis...

  17. Reliability criteria for voltage stability

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Carson W; Silverstein, Brian L [Bonneville Power Administration, Portland, OR (United States)

    1994-12-31

    In face of costs pressures, there is need to allocate scare resources more effectively in order to achieve voltage stability. This naturally leads to development of probabilistic criteria and notions of rick management. In this paper it is presented a discussion about criteria for long term voltage stability limited to the case in which the time frames are topically several minutes. (author) 14 refs., 1 fig.

  18. Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

    International Nuclear Information System (INIS)

    2014-01-01

    The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

  19. Eye bank procedures: donor selection criteria.

    Science.gov (United States)

    Sousa, Sidney Júlio de Faria E; Sousa, Stella Barretto de Faria E

    2018-01-01

    Eye banks use sterile procedures to manipulate the eye, antiseptic measures for ocular surface decontamination, and rigorous criteria for donor selection to minimize the possibility of disease transmission due to corneal grafting. Donor selection focuses on analysis of medical records and specific post-mortem serological tests. To guide and standardize procedures, eye bank associations and government agencies provide lists of absolute and relative contraindications for use of the tissue based on donor health history. These lists are guardians of the Hippocratic principle "primum non nocere." However, each transplantation carries risk of transmission of potentially harmful agents to the recipient. The aim of the procedures is not to eliminate risk, but limit it to a reasonable level. The balance between safety and corneal availability needs to be maintained by exercising prudence without disproportionate rigor.

  20. Prospective, randomized, double-blind, Phase 2 dose-ranging study comparing efficacy and safety of imipenem/cilastatin plus relebactam with imipenem/cilastatin alone in patients with complicated urinary tract infections.

    Science.gov (United States)

    Sims, Matthew; Mariyanovski, Valeri; McLeroth, Patrick; Akers, Wayne; Lee, Yu-Chieh; Brown, Michelle L; Du, Jiejun; Pedley, Alison; Kartsonis, Nicholas A; Paschke, Amanda

    2017-09-01

    The β-lactamase inhibitor relebactam can restore imipenem activity against imipenem non-susceptible pathogens. To explore relebactam's safety, tolerability and efficacy, we conducted a randomized (1:1:1), controlled, Phase 2 trial comparing imipenem/cilastatin+relebactam 250 mg, imipenem/cilastatin+relebactam 125 mg and imipenem/cilastatin alone in adults with complicated urinary tract infections (cUTI) or acute pyelonephritis, regardless of baseline pathogen susceptibility. Treatment was administered intravenously every 6 h for 4-14 days, with optional step-down to oral ciprofloxacin. The primary endpoint was favourable microbiological response rate (pathogen eradication) at discontinuation of intravenous therapy (DCIV) in the microbiologically evaluable (ME) population. Non-inferiority of imipenem/cilastatin+relebactam over imipenem/cilastatin alone was defined as lower bounds of the 95% CI for treatment differences being above -15%. At DCIV, 71 patients in the imipenem/cilastatin + 250 mg relebactam, 79 in the imipenem/cilastatin + 125 mg relebactam and 80 in the imipenem/cilastatin-only group were ME; 51.7% had cUTI and 48.3% acute pyelonephritis. Microbiological response rates were 95.5%, 98.6% and 98.7%, respectively, confirming non-inferiority of both imipenem/cilastatin + relebactam doses to imipenem/cilastatin alone. Clinical response rates were 97.1%, 98.7% and 98.8%, respectively. All 23 ME patients with imipenem non-susceptible pathogens had favourable DCIV microbiological responses (100% in each group). Among all 298 patients treated, 28.3%, 29.3% and 30.0% of patients, respectively, had treatment-emergent adverse events. The most common treatment-related adverse events across groups (1.0%-4.0%) were diarrhoea, nausea and headache. Imipenem/cilastatin + relebactam (250 or 125 mg) was as effective as imipenem/cilastatin alone for treatment of cUTI. Both relebactam-containing regimens were well tolerated. (NCT01505634).

  1. Investigation of effective decision criteria for multiobjective optimization in IMRT.

    Science.gov (United States)

    Holdsworth, Clay; Stewart, Robert D; Kim, Minsun; Liao, Jay; Phillips, Mark H

    2011-06-01

    To investigate how using different sets of decision criteria impacts the quality of intensity modulated radiation therapy (IMRT) plans obtained by multiobjective optimization. A multiobjective optimization evolutionary algorithm (MOEA) was used to produce sets of IMRT plans. The MOEA consisted of two interacting algorithms: (i) a deterministic inverse planning optimization of beamlet intensities that minimizes a weighted sum of quadratic penalty objectives to generate IMRT plans and (ii) an evolutionary algorithm that selects the superior IMRT plans using decision criteria and uses those plans to determine the new weights and penalty objectives of each new plan. Plans resulting from the deterministic algorithm were evaluated by the evolutionary algorithm using a set of decision criteria for both targets and organs at risk (OARs). Decision criteria used included variation in the target dose distribution, mean dose, maximum dose, generalized equivalent uniform dose (gEUD), an equivalent uniform dose (EUD(alpha,beta) formula derived from the linear-quadratic survival model, and points on dose volume histograms (DVHs). In order to quantatively compare results from trials using different decision criteria, a neutral set of comparison metrics was used. For each set of decision criteria investigated, IMRT plans were calculated for four different cases: two simple prostate cases, one complex prostate Case, and one complex head and neck Case. When smaller numbers of decision criteria, more descriptive decision criteria, or less anti-correlated decision criteria were used to characterize plan quality during multiobjective optimization, dose to OARs and target dose variation were reduced in the final population of plans. Mean OAR dose and gEUD (a = 4) decision criteria were comparable. Using maximum dose decision criteria for OARs near targets resulted in inferior populations that focused solely on low target variance at the expense of high OAR dose. Target dose range, (D

  2. NEW CRITERIA FOR ASSIGNING WASTE CONTAINING TECH-NOGENIC RADIONUCLIDES TO THE RADIOACTIVE WASTE

    Directory of Open Access Journals (Sweden)

    I. K. Romanovich

    2010-01-01

    Full Text Available The article contains detailed description of criteria for assigning of liquid and gaseous industrial waste containing technogenicradionuclides to the radioactive waste, presented in the new Basic Sanitary Rulesof Radiation Safety (OSPORB-99/2010. The analysisof shortcomings and discrepancies of the previously used in Russia system of criteria for assigning waste to the radioactive waste is given.

  3. 47 CFR 90.545 - TV/DTV interference protection criteria.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 5 2010-10-01 2010-10-01 false TV/DTV interference protection criteria. 90.545... the 763-775 and 793-805 MHz Bands § 90.545 TV/DTV interference protection criteria. Public safety base... reception of the signals of existing TV and DTV broadcast stations transmitting on TV Channels 62, 63, 64...

  4. 10 CFR Appendix S to Part 50 - Earthquake Engineering Criteria for Nuclear Power Plants

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Earthquake Engineering Criteria for Nuclear Power Plants S... FACILITIES Pt. 50, App. S Appendix S to Part 50—Earthquake Engineering Criteria for Nuclear Power Plants... nuclear power plant structures, systems, and components important to safety to withstand the effects of...

  5. Vaccine Safety

    Science.gov (United States)

    ... During Pregnancy Frequently Asked Questions about Vaccine Recalls Historical Vaccine Safety Concerns FAQs about GBS and Menactra ... CISA Resources for Healthcare Professionals Evaluation Current Studies Historical Background 2001-12 Publications Technical Reports Vaccine Safety ...

  6. SAFETY FIRST

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    Ensuring safety while peacefully utilizing nuclear energy is a top priority for China A fter a recent earthquake in Japan caused radioactive leaks at a nuclear power plant in Tokyo, the safety of nuclear energy has again aroused public attention.

  7. ITER safety

    International Nuclear Information System (INIS)

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  8. Water Safety

    Science.gov (United States)

    ... Staying Safe Videos for Educators Search English Español Water Safety KidsHealth / For Parents / Water Safety What's in ... remains your best measure of protection. Making Kids Water Wise It's important to teach your kids proper ...

  9. Safety first. Yes, but which safety?; Primat der Sicherheit. Ja, aber welche Sicherheit ist gemeint?

    Energy Technology Data Exchange (ETDEWEB)

    Roehlig, Klaus-Juergen [Technische Univ. Clausthal, Clausthal-Zellerfeld (Germany). Inst. fuer Endlagerforschung; Eckhardt, Anne [risicare GmbH, Zollikerberg (Switzerland)

    2017-09-01

    The site selection law in Germany and the final report of the final repository commission state the central objective to find a repository site that will guarantee safety for the next million of years. Decision makers, concerned and interested people have obviously different opinions and acceptance criteria with respect to the tools for the demonstration of safety (safety case). Possible solutions for a broad acceptance of safety definitions are discussed.

  10. Social Advertising Quality: Assessment Criteria

    Directory of Open Access Journals (Sweden)

    S. B. Kalmykov

    2017-01-01

    Full Text Available Purpose: the The purpose of the publication is development of existing criterial assessment in social advertising sphere. The next objectives are provided for its achievement: to establish research methodology, to develop the author’s version of necessary notional apparatus and conceptual generalization, to determine the elements of social advertising quality, to establish the factors of its quality, to conduct the systematization of existing criteria and measuring instruments of quality assessment, to form new criteria of social advertising quality, to apply received results for development of criterial assessment to determine the further research perspectives. Methods: the methodology of research of management of social advertising interaction with target audience, which has dynamic procedural character with use of sociological knowledge multivariate paradigmatic status, has been proposed. Results: the primary received results: the multivariate paradigmatic research basis with use of works of famous domestic and foreign scientists in sociology, qualimetry and management spheres; the definitions of social advertising, its quality, sociological quality provision system, target audience behavior model during social advertising interaction are offered; the quality factors with three groups by level of effect on consumer are established; the systematization of existing quality and its measure instruments assessment criteria by detected social advertising quality elements are conducted; the two new criteria and its management quality assessment measuring instruments in social advertising sphere are developed; the one of the common groups of production quality criteria – adaptability with considering of new management quality criteria and conducted systematization of existing social advertising creative quality assessment criteria development; the perspective of further perfection of quality criterial assessment based on social advertising

  11. Food safety

    Science.gov (United States)

    ... safety URL of this page: //medlineplus.gov/ency/article/002434.htm Food safety To use the sharing features on this page, please enable JavaScript. Food safety refers to the conditions and practices that preserve the quality of food. These practices prevent contamination and foodborne ...

  12. Safety analyses for NHR-200

    Energy Technology Data Exchange (ETDEWEB)

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The NHR-200 is a commercial 200-MW District Heating Reactor developed in China. It is designed on the basis of design, construction and four-year operating experience of the 5MW Experimental Heating Reactor (NHR-5). It has special safety features which are briefly described in this paper. Accident classification and safety criteria are also explained. Some typical and serious accidents are studied theoretically, and their results are detailed in this paper. They demonstrate the excellent safety characteristics of HR-200. (author). 4 refs, 9 figs, 1 tab.

  13. Study on the KALIMER safety approach

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Han, Do Hee; Kim, Young Cheol.

    1997-01-01

    This study describes KALIMER's safety approach, how to establish the safety criteria and temperature limit, how to define safety evaluation events, and some safety research and development needs items. It is recommended that the KALIMER's approach to safety use seven levels of safety design and a defense-in-depth design approach with particular emphasis on inherent passive features. In order to establish as set DBEs for KALIMER safety evaluation, the procedure is explained how to define safety evaluation events. Final selection is to be determined later with the final establishment of design concepts. On the basis of preliminary studies and evaluation of the plant safety related areas, the KALIMER and PRISM have following three main difference that may require special research and development for KALIMER. (author). 7 refs., 6 tabs., 6 figs

  14. Perspectives on dam safety in Canada

    International Nuclear Information System (INIS)

    Halliday, R.

    2004-01-01

    Canadian dam safety issues were reviewed from the perspective of a water resources engineer who is not a dam safety practitioner. Several external factors affecting dam safety were identified along with perceived problems in dam safety administration. The author claims that the main weakness in safety practices can be attributed to provincial oversights and lack of federal engagement. Some additions to the Canadian Dam Safety Guidelines were proposed to address these weaknesses. Canada has hundreds of large dams and high hazard dams whose failure would result in severe downstream consequences. The safety of dams built on boundary waters shared with the United States have gained particular attention from the International Joint Commission. This paper also examined safety criteria for concerns such as aging dams, sabotage and global climate change that may compromise the safety of a dam. 26 refs

  15. ITER operating limit definition criteria

    International Nuclear Information System (INIS)

    Ciattaglia, S.; Barabaschi, P.; Carretero, J.A.; Chiocchio, S.; Hureau, D.; Girard, J.Ph.; Gordon, C.; Portone, A.; Rodrigo, L. Rodriguez; Roldan, C.; Saibene, G.; Uzan-Elbez, J.

    2009-01-01

    The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction. OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent. This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.

  16. Criteria for saturated magnetization loop

    International Nuclear Information System (INIS)

    Harres, A.; Mikhov, M.; Skumryev, V.; Andrade, A.M.H. de; Schmidt, J.E.; Geshev, J.

    2016-01-01

    Proper estimation of magnetization curve parameters is vital in studying magnetic systems. In the present article, criteria for discrimination non-saturated (minor) from saturated (major) hysteresis loops are proposed. These employ the analysis of (i) derivatives of both ascending and descending branches of the loop, (ii) remanent magnetization curves, and (iii) thermomagnetic curves. Computational simulations are used in order to demonstrate their validity. Examples illustrating the applicability of these criteria to well-known real systems, namely Fe_3O_4 and Ni fine particles, are provided. We demonstrate that the anisotropy-field value estimated from a visual examination of an only apparently major hysteresis loop could be more than two times lower than the real one. - Highlights: • Proper estimation of hysteresis-loop parameters is vital in magnetic studies. • We propose criteria for discrimination minor from major hysteresis loops. • The criteria analyze magnetization, remanence and ZFC/FC curves and/or their derivatives. • Examples of their application on real nanoparticles systems are given. • Using the criteria could avoid twofold or bigger saturation-field underestimation errors.

  17. Criteria for saturated magnetization loop

    Energy Technology Data Exchange (ETDEWEB)

    Harres, A. [Departamento de Física, UFSM, Santa Maria, 97105-900 Rio Grande do Sul (Brazil); Mikhov, M. [Faculty of Physics, University of Sofia, 1164 Sofia (Bulgaria); Skumryev, V. [Institució Catalana de Recerca i Estudis Avançats, 08010 Barcelona (Spain); Departament de Física, Universitat Autònoma de Barcelona, 08193 Barcelona (Spain); Andrade, A.M.H. de; Schmidt, J.E. [Instituto de Física, UFRGS, Porto Alegre, 91501-970 Rio Grande do Sul (Brazil); Geshev, J., E-mail: julian@if.ufrgs.br [Departament de Física, Universitat Autònoma de Barcelona, 08193 Barcelona (Spain); Instituto de Física, UFRGS, Porto Alegre, 91501-970 Rio Grande do Sul (Brazil)

    2016-03-15

    Proper estimation of magnetization curve parameters is vital in studying magnetic systems. In the present article, criteria for discrimination non-saturated (minor) from saturated (major) hysteresis loops are proposed. These employ the analysis of (i) derivatives of both ascending and descending branches of the loop, (ii) remanent magnetization curves, and (iii) thermomagnetic curves. Computational simulations are used in order to demonstrate their validity. Examples illustrating the applicability of these criteria to well-known real systems, namely Fe{sub 3}O{sub 4} and Ni fine particles, are provided. We demonstrate that the anisotropy-field value estimated from a visual examination of an only apparently major hysteresis loop could be more than two times lower than the real one. - Highlights: • Proper estimation of hysteresis-loop parameters is vital in magnetic studies. • We propose criteria for discrimination minor from major hysteresis loops. • The criteria analyze magnetization, remanence and ZFC/FC curves and/or their derivatives. • Examples of their application on real nanoparticles systems are given. • Using the criteria could avoid twofold or bigger saturation-field underestimation errors.

  18. Safety handbook

    International Nuclear Information System (INIS)

    1990-01-01

    The purpose of the Australian Nuclear Science and Technology Organization's Safety Handbook is to outline simply the fundamental procedures and safety precautions which provide an appropriate framework for safe working with any potential hazards, such as fire and explosion, welding, cutting, brazing and soldering, compressed gases, cryogenic liquids, chemicals, ionizing radiations, non-ionising radiations, sound and vibration, as well as safety in the office. It also specifies the organisation for safety at the Lucas Heights Research Laboratories and the responsibilities of individuals and committees. It also defines the procedures for the scrutiny and review of all operations and the resultant setting of safety rules for them. ills

  19. Evaluation and construction of diagnostic criteria for inclusion body myositis

    Science.gov (United States)

    Mammen, Andrew L.; Amato, Anthony A.; Weiss, Michael D.; Needham, Merrilee

    2014-01-01

    Objective: To use patient data to evaluate and construct diagnostic criteria for inclusion body myositis (IBM), a progressive disease of skeletal muscle. Methods: The literature was reviewed to identify all previously proposed IBM diagnostic criteria. These criteria were applied through medical records review to 200 patients diagnosed as having IBM and 171 patients diagnosed as having a muscle disease other than IBM by neuromuscular specialists at 2 institutions, and to a validating set of 66 additional patients with IBM from 2 other institutions. Machine learning techniques were used for unbiased construction of diagnostic criteria. Results: Twenty-four previously proposed IBM diagnostic categories were identified. Twelve categories all performed with high (≥97%) specificity but varied substantially in their sensitivities (11%–84%). The best performing category was European Neuromuscular Centre 2013 probable (sensitivity of 84%). Specialized pathologic features and newly introduced strength criteria (comparative knee extension/hip flexion strength) performed poorly. Unbiased data-directed analysis of 20 features in 371 patients resulted in construction of higher-performing data-derived diagnostic criteria (90% sensitivity and 96% specificity). Conclusions: Published expert consensus–derived IBM diagnostic categories have uniformly high specificity but wide-ranging sensitivities. High-performing IBM diagnostic category criteria can be developed directly from principled unbiased analysis of patient data. Classification of evidence: This study provides Class II evidence that published expert consensus–derived IBM diagnostic categories accurately distinguish IBM from other muscle disease with high specificity but wide-ranging sensitivities. PMID:24975859

  20. Safety analysis to support a safe operating envelope for fuel

    International Nuclear Information System (INIS)

    Gibb, R.A.; Reid, P.J.

    1998-01-01

    This paper presents an approach for defining a safe operating envelope for fuel. 'Safe operating envelope' is defined as an envelope of fuel parameters defined for application in safety analysis that can be related to, or used to define, the acceptable range of fuel conditions due to operational transients or deviations in fuel manufacturing processes. The paper describes the motivation for developing such a methodology. The methodology involved four steps: the update of fission product inventories, the review of sheath failure criteria, a review of input parameters to be used in fuel modelling codes, and the development of an improved fission product release code. This paper discusses the aspects of fuel sheath failure criteria that pertain to operating or manufacturing conditions and to the evaluation and selection of modelling input data. The other steps are not addressed in this paper since they have been presented elsewhere. (author)

  1. Packaging design criteria modified fuel spacer burial box. Revision 1

    International Nuclear Information System (INIS)

    Stevens, P.F.

    1994-01-01

    Various Hanford facilities must transfer large radioactively contaminated items to burial/storage. Presently, there are eighteen Fuel Spacer Burial Boxes (FSBBs) available on the Hanford Site for transport of such items. Previously, the FSBBS were transported from a rail car to the burial trench via a drag-off operation. To allow for the lifting of the boxes into the burial trench, it will be necessary to improve the packagings lifting attachments and provide structural reinforcement. Additional safety improvements to the packaging system will be provided by the addition of a positive closure system and package ventilation. FSBBs that are modified in such a manner are referred to as Modified Fuel Spacer Burial Boxes (MFSBs). The criteria provided by this PDC will be used to demonstrate that the transfer of the MFSB will provide an equivalent degree of safety as would be provided by a package meeting offsite transportation requirements. This fulfills the onsite transportation safety requirements implemented in WHC-CM-2-14, Hazardous Material Packaging and Shipping. A Safety Analysis Report for Packaging (SARP) will be prepared to evaluate the safety of the transfer operation. Approval of the SARP is required to authorize transfer. Criteria are also established to ensure burial requirements are met

  2. Food irradiation: its role in food safety

    Energy Technology Data Exchange (ETDEWEB)

    Qureshi, R U

    1986-12-31

    There are food safety criteria generally defined by international groups and specifically defined by individual countries. Food irradiation will be discussed in the light of food safety regulations. The merits and acceptability of food irradiation in promoting trade within and between countries will also be discussed. The need for public awareness and training of technical personnel will be highlighted

  3. 40 CFR 258.10 - Airport safety.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Airport safety. 258.10 Section 258.10 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) SOLID WASTES CRITERIA FOR MUNICIPAL SOLID WASTE LANDFILLS Location Restrictions § 258.10 Airport safety. (a) Owners or operators of new...

  4. Annual report on occupational safety 1987

    International Nuclear Information System (INIS)

    1988-01-01

    This report presents detailed information on occupational safety relating to the Company's employees for 1987. Data are quoted in tables and text, together with data from the previous year for comparison where available. The report is presented under the following headings: radiological and non-radiological safety, incidents, appendices (statutory dose limits, nuclear incident criteria for reporting to ministers). (author)

  5. Safety objectives for nuclear activities in Canada

    International Nuclear Information System (INIS)

    1982-04-01

    This report by the Advisory Committee on Nuclear Safety presents a concise statement of the basic safety objectives which the Committee considers underlie, or should underlie, the regulations and the licensing and compliance practices of the Atomic Energy Control Board. The report also includes a number of general criteria for achieving these objectives

  6. Food irradiation: its role in food safety

    International Nuclear Information System (INIS)

    Qureshi, R. U.

    1985-01-01

    There are food safety criteria generally defined by international groups and specifically defined by individual countries. Food irradiation will be discussed in the light of food safety regulations. The merits and acceptability of food irradiation in promoting trade within and between countries will also be discussed. The need for public awareness and training of technical personnel will be highlighted

  7. Tonopah Test Range - Index

    Science.gov (United States)

    Capabilities Test Operations Center Test Director Range Control Track Control Communications Tracking Radars Photos Header Facebook Twitter YouTube Flickr RSS Tonopah Test Range Top TTR_TOC Tonopah is the testing range of choice for all national security missions. Tonopah Test Range (TTR) provides research and

  8. Design criteria of integrated reactors based on transients

    International Nuclear Information System (INIS)

    Zanocco, P.; Gimenez, M.; Delmastro, D.

    1999-01-01

    A new tendency in integrated reactors conceptual design is to include safety criteria through accident analysis. In this work, the effect of design parameters in a Loss of Heat Sink transient using design maps is analyzed. Particularly, geometry related parameters and reactivity coefficients are studied. Also the effect of primary relief/safety valve during the transient is evaluated. A design map for valve area vs. coolant density reactivity coefficient is obtained. A computer code (HUARPE) is developed in order to simulate these transients. Coolant, steam dome, pressure vessel structures and core models are implemented. This code is checked against TRAC with satisfactory results. (author)

  9. Typical design/qualification acceptance criteria for newly installed pipelines and equipment components of VVER-type NPPs

    International Nuclear Information System (INIS)

    Masopust, R.

    2003-01-01

    This paper describes in general the typical design/qualification acceptance criteria and seismic acceptance criteria in particular that are applicable for important to safety newly installed pipelines and equipment components of VVER-type already existing NPPs, specifically during the design verification phase of this newly installed equipment. These criteria are currently used for VVER 440-213 and VVER 1000 NPPs in Czech Republic and in Slovakia. The similar criteria are also used in Hungary. (author)

  10. Nuclear Safety

    International Nuclear Information System (INIS)

    1978-09-01

    In this short paper it has only been possible to deal in a rather general way with the standards of safety used in the UK nuclear industry. The record of the industry extending over at least twenty years is impressive and, indeed, unique. No other industry has been so painstaking in protection of its workers and in its avoidance of damage to the environment. Headings are: introduction; how a nuclear power station works; radiation and its effects (including reference to ICRP, the UK National Radiological Protection Board, and safety standards); typical radiation doses (natural radiation, therapy, nuclear power programme and other sources); safety of nuclear reactors - design; key questions (matters of concern which arise in the public mind); safety of operators; safety of people in the vicinity of a nuclear power station; safety of the general public; safety bodies. (U.K.)

  11. User perspectives on relevance criteria

    DEFF Research Database (Denmark)

    Maglaughlin, Kelly L.; Sonnenwald, Diane H.

    2002-01-01

    , partially relevant, or not relevant to their information need; and explained their decisions in an interview. Analysis revealed 29 criteria, discussed positively and negatively, that were used by the participants when selecting passages that contributed or detracted from a document's relevance......This study investigates the use of criteria to assess relevant, partially relevant, and not-relevant documents. Study participants identified passages within 20 document representations that they used to make relevance judgments; judged each document representation as a whole to be relevant...... matter, thought catalyst), full text (e.g., audience, novelty, type, possible content, utility), journal/publisher (e.g., novelty, main focus, perceived quality), and personal (e.g., competition, time requirements). Results further indicate that multiple criteria are used when making relevant, partially...

  12. Criteria for authorship in bioethics.

    Science.gov (United States)

    Resnik, David B; Master, Zubin

    2011-10-01

    Multiple authorship is becoming increasingly common in bioethics research. There are well-established criteria for authorship in empirical bioethics research but not for conceptual research. It is important to develop criteria for authorship in conceptual publications to prevent undeserved authorship and uphold standards of fairness and accountability. This article explores the issue of multiple authorship in bioethics and develops criteria for determining who should be an author on a conceptual publication in bioethics. Authorship in conceptual research should be based on contributing substantially to: (1) identifying a topic, problem, or issue to study; (2) reviewing and interpreting the relevant literature; (3) formulating, analyzing, and evaluating arguments that support one or more theses; (4) responding to objections and counterarguments; and (5) drafting the manuscript. Authors of conceptual publications should participate substantially in at least two of areas (1)-(5) and also approve the final version. [corrected].

  13. National Risk Assessment in The Netherlands : A Multi-Criteria Decision Analysis Approach

    NARCIS (Netherlands)

    Pruyt, E.; Wijnmalen, D.J.D.

    2010-01-01

    Nowadays, National Safety and Security issues receive much attention in many countries. In 2007, the Dutch government approved a National Safety and Security Strategy based on a multi-criteria analysis approach to classify potential threats and hazards. The general methodology of this Dutch National

  14. Controlling criteria for radiation exposure of astronauts and space workers

    International Nuclear Information System (INIS)

    Katoh, Kazuaki

    1989-01-01

    Space workers likely to suffer from radiation exposure in the outer space are currently limited to the U.S. and Soviet Union, and only a small amount of data and information is available concerning the techniques and criteria for control of radiation exposure in this field. Criteria used in the Soviet Union are described first. The criteria (TRS-75), called the Radiation Safety Criteria for Space Navigation, are tentative ones set up in 1975. They are based on risk assessment. The standard radiation levels are established based on unit flight time: 50rem for 1 month, 80rem for 3 months, 110rem for 6 months and 150rem for 12 months. These are largely different from the emergency exposure limit of 100mSv (10rem) specified in a Japanese law, and the standard annual exposure value of 50mSv (5rem) for workers in nuclear power plants at normal times. For the U.S., J.A. Angelo, Jr., presented a paper titled 'Radiation Protection Issues and Techniques concerning Extended Manned Space Missions' at an IAEA meeting held in 1988. Though the criteria shown in the paper are not formal ones at the national level, similar criteria are expected to be adopted by the nation in the near future. The exposure limits recommended in the paper include a depth dose of 1-4Sv for the whole life span of a worker. (Nogami, K.)

  15. Safety of WWER type nuclear power plants - viewing from Hungary

    International Nuclear Information System (INIS)

    Voeroess, L.

    1991-01-01

    An evaluation of WWER type nuclear power plants operating in Hungary is given, relative to the safety requirements accepted internationally; how safe can they be regarded and what can be done to assure a high level of safety in all case. After an overview of general safety criteria, an overall description of WWER-440 type nuclear reactors is presented. Design safety, operational safety issues are treated in detail. Safety inspection and safety-related research and development is discussed. Regarding the future, five different issues associated with nuclear reactor safety should be considered. (R.P.) 20 refs.; 12 figs.; 3 tabs

  16. UO3 deactivation end point criteria

    International Nuclear Information System (INIS)

    Stefanski, L.D.

    1994-01-01

    The UO 3 Deactivation End Point Criteria are necessary to facilitate the transfer of the UO 3 Facility from the Office of Facility Transition and Management (EM-60) to the office of Environmental Restoration (EM-40). The criteria were derived from a logical process for determining end points for the systems and spaces at the UO 3 , Facility based on the objectives, tasks, and expected future uses pertinent to that system or space. Furthermore, the established criteria meets the intent and supports the draft guidance for acceptance criteria prepared by EM-40, open-quotes U.S. Department of Energy office of Environmental Restoration (EM-40) Decontamination and Decommissioning Guidance Document (Draft).close quotes For the UO 3 Facility, the overall objective of deactivation is to achieve a safe, stable and environmentally sound condition, suitable for an extended period, as quickly and economically as possible. Once deactivated, the facility is kept in its stable condition by means of a methodical surveillance and maintenance (S ampersand M) program, pending ultimate decontamination and decommissioning (D ampersand D). Deactivation work involves a range of tasks, such as removal of hazardous material, elimination or shielding of radiation fields, partial decontamination to permit access for inspection, installation of monitors and alarms, etc. it is important that the end point of each of these tasks be established clearly and in advance, for the following reasons: (1) End points must be such that the central element of the deactivation objective - to achieve stability - is unquestionably achieved. (2) Much of the deactivation work involves worker exposure to radiation or dangerous materials. This can be minimized by avoiding unnecessary work. (3) Each task is, in effect, competing for resources with other deactivation tasks and other facilities. By assuring that each task is appropriately bounded, DOE's overall resources can be used most fully and effectively

  17. Inherent/passive safety for fusion

    International Nuclear Information System (INIS)

    Piet, S.J.

    1986-06-01

    The concept of inherent or passive passive safety for fusion energy is explored, defined, and partially quantified. Four levels of safety assurance are defined, which range from true inherent safety to passive safety to protection via active engineered safeguard systems. Fusion has the clear potential for achieving inherent or passive safety, which should be an objective of fusion research and design. Proper material choice might lead to both inherent safety and high mass power density, improving both safety and economics. When inherent safety is accomplished, fusion will be well on the way to achieving its ultimate potential and to be truly different and superior

  18. 76 FR 67020 - Railroad Safety Advisory Committee; Notice of Meeting

    Science.gov (United States)

    2011-10-28

    ... Device Distraction, Critical Incident, Track Safety Standards, Dark Territory, Passenger Safety, and... requisite range of views and expertise necessary to discharge its responsibilities. See the RSAC Web site...

  19. OSCILLATION CRITERIA FOR FORCED SUPERLINEAR DIFFERENCE EQUATIONS

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Using Riccati transformation techniques,some oscillation criteria for the forced second-order superlinear difference equations are established.These criteria are dis- crete analogues of the criteria for differential equations proposed by Yan.

  20. AERB information booklet: personal protective equipment- safety footwear

    International Nuclear Information System (INIS)

    1992-01-01

    The main classes of safety footwear required for industrial operations in the units of Department of Atomic Energy are the following; leather safety boots and shoes, firemen's leather boots - Wellington type, electrical safety shoes, chemical safety shoes, shoes suitable for mining operations. The criteria to be adopted for selection of safety shoes for nuclear installations are given. (M.K.V.). 5 annexures, 1 appendix

  1. The dialectical thinking about deterministic and probabilistic safety analysis

    International Nuclear Information System (INIS)

    Qian Yongbai; Tong Jiejuan; Zhang Zuoyi; He Xuhong

    2005-01-01

    There are two methods in designing and analysing the safety performance of a nuclear power plant, the traditional deterministic method and the probabilistic method. To date, the design of nuclear power plant is based on the deterministic method. It has been proved in practice that the deterministic method is effective on current nuclear power plant. However, the probabilistic method (Probabilistic Safety Assessment - PSA) considers a much wider range of faults, takes an integrated look at the plant as a whole, and uses realistic criteria for the performance of the systems and constructions of the plant. PSA can be seen, in principle, to provide a broader and realistic perspective on safety issues than the deterministic approaches. In this paper, the historical origins and development trend of above two methods are reviewed and summarized in brief. Based on the discussion of two application cases - one is the changes to specific design provisions of the general design criteria (GDC) and the other is the risk-informed categorization of structure, system and component, it can be concluded that the deterministic method and probabilistic method are dialectical and unified, and that they are being merged into each other gradually, and being used in coordination. (authors)

  2. Safety requirements for a nuclear power plant electric power system

    Energy Technology Data Exchange (ETDEWEB)

    Fouad, L F; Shinaishin, M A

    1988-06-15

    This work aims at identifying the safety requirements for the electric power system in a typical nuclear power plant, in view of the UNSRC and the IAEA. Description of a typical system is provided, followed by a presentation of the scope of the information required for safety evaluation of the system design and performance. The acceptance and design criteria that must be met as being specified by both regulatory systems, are compared. Means of implementation of such criteria as being described in the USNRC regulatory guides and branch technical positions on one hand and in the IAEA safety guides on the other hand are investigated. It is concluded that the IAEA regulations address the problems that may be faced with in countries having varying grid sizes ranging from large stable to small potentially unstable ones; and that they put emphasis on the onsite standby power supply. Also, in this respect the Americans identify the grid as the preferred power supply to the plant auxiliaries, while the IAEA leaves the possibility that the preferred power supply could be either the grid or the unit main generator depending on the reliability of each. Therefore, it is found that it is particularly necessary in this area of electric power supplies to deal with the IAEA and the American sets of regulations as if each complements and not supplements the other. (author)

  3. Accident analysis and DOE criteria

    International Nuclear Information System (INIS)

    Graf, J.M.; Elder, J.C.

    1982-01-01

    In analyzing the radiological consequences of major accidents at DOE facilities one finds that many facilities fall so far below the limits of DOE Order 6430 that compliance is easily demonstrated by simple analysis. For those cases where the amount of radioactive material and the dispersive energy available are enough for accident consequences to approach the limits, the models and assumptions used become critical. In some cases the models themselves are the difference between meeting the criteria or not meeting them. Further, in one case, we found that not only did the selection of models determine compliance but the selection of applicable criteria from different chapters of Order 6430 also made the difference. DOE has recognized the problem of different criteria in different chapters applying to one facility, and has proceeded to make changes for the sake of consistency. We have proposed to outline the specific steps needed in an accident analysis and suggest appropriate models, parameters, and assumptions. As a result we feed DOE siting and design criteria will be more fairly and consistently applied

  4. Aversive Stimulation -- Criteria for Application.

    Science.gov (United States)

    O'Donnell, Patrick A.; Ohlson, Glenn A.

    Criteria for applying aversive stimulation with severely handicapped children are examined, and practical and ethical issues are considered. Factors seen to influence punishment outcomes include timing, intensity, and schedule of reinforcement. Suggested is the need for further research on the comparative effectiveness of positive and negative…

  5. Risk based seismic design criteria

    International Nuclear Information System (INIS)

    Kennedy, R.P.

    1999-01-01

    In order to develop a risk based seismic design criteria the following four issues must be addressed: (1) What target annual probability of seismic induced unacceptable performance is acceptable? (2) What minimum seismic margin is acceptable? (3) Given the decisions made under Issues 1 and 2, at what annual frequency of exceedance should the safe-shutdown-earthquake (SSE) ground motion be defined? (4) What seismic design criteria should be established to reasonably achieve the seismic margin defined under Issue 2? The first issue is purely a policy decision and is not addressed in this paper. Each of the other three issues are addressed. Issues 2 and 3 are integrally tied together so that a very large number of possible combinations of responses to these two issues can be used to achieve the target goal defined under Issue 1. Section 2 lays out a combined approach to these two issues and presents three potentially attractive combined resolutions of these two issues which reasonably achieves the target goal. The remainder of the paper discusses an approach which can be used to develop seismic design criteria aimed at achieving the desired seismic margin defined in resolution of Issue 2. Suggestions for revising existing seismic design criteria to more consistently achieve the desired seismic margin are presented. (orig.)

  6. Quality criteria for phase change materials selection

    International Nuclear Information System (INIS)

    Vitorino, Nuno; Abrantes, João C.C.; Frade, Jorge R.

    2016-01-01

    Highlights: • Selection criteria of phase change materials for representative applications. • Selection criteria based on reliable solutions for latent heat transfer. • Guidelines for the role of geometry and heat transfer mechanisms. • Performance maps based on PCM properties, operating conditions, size and time scales. - Abstract: Selection guidelines are primary criterion for optimization of materials for specific applications in order to meet simultaneous and often conflicting requirements. This is mostly true for technologies and products required to meet the main societal needs, such as energy. In this case, gaps between supply and demand require strategies for energy conversion and storage, including thermal storage mostly based on phase change materials. Latent heat storage is also very versatile for thermal management and thermal control by allowing high storage density within narrow temperature ranges without strict dependence between stored thermal energy and temperature. Thus, this work addressed the main issues of latent heat storage from a materials selection perspective, based on expected requirements of applications in thermal energy storage or thermal regulation. Representative solutions for the kinetics of latent heat charge/discharge were used to derive optimization guidelines for high energy density, high power, response time (from fast response to thermal inertia), etc. The corresponding property relations were presented in graphical forms for a wide variety of prospective phase change materials, and for wide ranges of operating conditions, and accounting for changes in geometry and mechanisms.

  7. A comparison of international criteria for the ultimate storage of radioactive wastes

    International Nuclear Information System (INIS)

    Mielke, H.

    1985-01-01

    In countries other than the Federal Republic of Germany and internationally there are no comprehensive codes referring to criteria and safety requirements except those of the IAEA and USA. In other countries there exist safety goals for the ultimate storage or for purely geological criteria. The degree of detailing regulations differs widely abroad and internationally. Safety goals abroad and internationally as well as measures for their realisation in the ultimate storage of radioactive wastes in deep geological formations are in line with the German safety goals. The IAEA refers to general aspects of geological, waste technology and ultimate storage technology criteria. In the USA, ultimate storage technology criteria have been quantified in part. The quantitative geological criteria existing in Great Britain and in the Netherlands are only relevant in as much as safety analyses must be performed for a specific site to provide evidence for the safety of this site. The comparison shows that most requirements pronounced abroad are also made for the Federal Republic of Germany. Some requirements are more specified in the Federal Republic of Germany, some are more detailed abroad. (orig./HP) [de

  8. Interim performance criteria for photovoltaic energy systems. [Glossary included

    Energy Technology Data Exchange (ETDEWEB)

    DeBlasio, R.; Forman, S.; Hogan, S.; Nuss, G.; Post, H.; Ross, R.; Schafft, H.

    1980-12-01

    This document is a response to the Photovoltaic Research, Development, and Demonstration Act of 1978 (P.L. 95-590) which required the generation of performance criteria for photovoltaic energy systems. Since the document is evolutionary and will be updated, the term interim is used. More than 50 experts in the photovoltaic field have contributed in the writing and review of the 179 performance criteria listed in this document. The performance criteria address characteristics of present-day photovoltaic systems that are of interest to manufacturers, government agencies, purchasers, and all others interested in various aspects of photovoltaic system performance and safety. The performance criteria apply to the system as a whole and to its possible subsystems: array, power conditioning, monitor and control, storage, cabling, and power distribution. They are further categorized according to the following performance attributes: electrical, thermal, mechanical/structural, safety, durability/reliability, installation/operation/maintenance, and building/site. Each criterion contains a statement of expected performance (nonprescriptive), a method of evaluation, and a commentary with further information or justification. Over 50 references for background information are also given. A glossary with definitions relevant to photovoltaic systems and a section on test methods are presented in the appendices. Twenty test methods are included to measure performance characteristics of the subsystem elements. These test methods and other parts of the document will be expanded or revised as future experience and needs dictate.

  9. Evaluating Dependence Criteria for Caffeine.

    Science.gov (United States)

    Striley, Catherine L W; Griffiths, Roland R; Cottler, Linda B

    2011-12-01

    Background: Although caffeine is the most widely used mood-altering drug in the world, few studies have operationalized and characterized Diagnostic and Statistical Manual IV (DSM-IV) substance dependence criteria applied to caffeine. Methods: As a part of a nosological study of substance use disorders funded by the National Institute on Drug Abuse, we assessed caffeine use and dependence symptoms among high school and college students, drug treatment patients, and pain clinic patients who reported caffeine use in the last 7 days and also reported use of alcohol, nicotine, or illicit drugs within the past year ( n =167). Results: Thirty-five percent met the criteria for dependence when all seven of the adopted DSM dependence criteria were used. Rates of endorsement of several of the most applicable diagnostic criteria were as follows: 26% withdrawal, 23% desire to cut down or control use, and 44% continued use despite harm. In addition, 34% endorsed craving, 26% said they needed caffeine to function, and 10% indicated that they talked to a physician or counselor about problems experienced with caffeine. There was a trend towards increased caffeine dependence among those dependent on nicotine or alcohol. Within a subgroup that had used caffeine, alcohol, and nicotine in the past year, 28% fulfilled criteria for caffeine dependence compared to 50% for alcohol and 80% for nicotine. Conclusion: The present study adds to a growing literature suggesting the reliability, validity, and clinical utility of the caffeine dependence diagnosis. Recognition of caffeine dependence in the DSM-V may be clinically useful.

  10. Safety culture

    International Nuclear Information System (INIS)

    1991-01-01

    The response to a previous publication by the International Nuclear Safety Advisory Group (INSAG), indicated a broad international interest in expansion of the concept of Safety Culture, in such a way that its effectiveness in particular cases may be judged. This report responds to that need. In its manifestation, Safety Culture has two major components: the framework determined by organizational policy and by managerial action, and the response of individuals in working within and benefiting by the framework. 1 fig

  11. Safety; Avertissement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

  12. Transferability of habitat suitability criteria for fishes in warmwater streams

    Science.gov (United States)

    Freeman, Mary C.; Bowen, Z.H.; Crance, J.H.

    1997-01-01

    We developed habitat suitability criteria and tested their transferability for nine fishes inhabiting unregulated Piedmont and Coastal Plain streams in Alabama. Cr iteria for optimal habitat were defined as ranges of depth, velocity, substrate type and cover type for which a species' suitability index (proportional abundance divided by proportional habitat availability, scaled from 0 to 1) equalled or exceeded 0.4. We evaluated the transferability of criteria between study sites by testing the null hypothesis that species occurrence in a sample was independent of whether or not the sample was taken in optimal habitat. We also tested criteria transference to a large, flow-regulated river sampled during low flow periods. Depth, velocity and most substrate criteria developed for the bronze darter Percina palmaris successfully transferred between unregulated streams and to the flow-regulated river samples. All criteria developed for a pair of closely related, allopatric darter species, Etheostoma chuckwachattee and E. jordani, transferred sucessfully when applied between species (in the unregulated sites) and to the regulated river samples. In contrast, criteria for the Alabama shiner Cyprinella callistia failed nearly all tests of transferability. Criteria for E. stigmaeum, P. nigrofasciata, an undescribed Percina species, and a pair of related, allopatric Cyprinella species transferred inconsistently. The species with good criteria transference had high suitability indices for shallow depths, fast current velocities and coarse substrates, characteristic of riffle species. We suggest that microhabitat criteria for riffle fishes are more likely to provide a transferable measure of habitat quality than criteria for fishes that, although restricted to fluvial habitats, commonly occupy a variety of pool and riffle habitats.

  13. Does Employee Safety Matter for Patients Too? Employee Safety Climate and Patient Safety Culture in Health Care.

    Science.gov (United States)

    Mohr, David C; Eaton, Jennifer Lipkowitz; McPhaul, Kathleen M; Hodgson, Michael J

    2015-04-22

    We examined relationships between employee safety climate and patient safety culture. Because employee safety may be a precondition for the development of patient safety, we hypothesized that employee safety culture would be strongly and positively related to patient safety culture. An employee safety climate survey was administered in 2010 and assessed employees' views and experiences of safety for employees. The patient safety survey administered in 2011 assessed the safety culture for patients. We performed Pearson correlations and multiple regression analysis to examine the relationships between a composite measure of employee safety with subdimensions of patient safety culture. The regression models controlled for size, geographic characteristics, and teaching affiliation. Analyses were conducted at the group level using data from 132 medical centers. Higher employee safety climate composite scores were positively associated with all 9 patient safety culture measures examined. Standardized multivariate regression coefficients ranged from 0.44 to 0.64. Medical facilities where staff have more positive perceptions of health care workplace safety climate tended to have more positive assessments of patient safety culture. This suggests that patient safety culture and employee safety climate could be mutually reinforcing, such that investments and improvements in one domain positively impacts the other. Further research is needed to better understand the nexus between health care employee and patient safety to generalize and act upon findings.

  14. Periodic safety analyses; Les essais periodiques

    Energy Technology Data Exchange (ETDEWEB)

    Gouffon, A; Zermizoglou, R

    1990-12-01

    The IAEA Safety Guide 50-SG-S8 devoted to 'Safety Aspects of Foundations of Nuclear Power Plants' indicates that operator of a NPP should establish a program for inspection of safe operation during construction, start-up and service life of the plant for obtaining data needed for estimating the life time of structures and components. At the same time the program should ensure that the safety margins are appropriate. Periodic safety analysis are an important part of the safety inspection program. Periodic safety reports is a method for testing the whole system or a part of the safety system following the precise criteria. Periodic safety analyses are not meant for qualification of the plant components. Separate analyses are devoted to: start-up, qualification of components and materials, and aging. All these analyses are described in this presentation. The last chapter describes the experience obtained for PWR-900 and PWR-1300 units from 1986-1989.

  15. Implications of stress range for inelastic analysis

    International Nuclear Information System (INIS)

    Karabin, M.E.; Dhalla, A.K.

    1981-01-01

    The elastic stress range over a complete load cycle is routinely used to formulate simplified rules regarding the inelastic behavior of structures operating at elevated temperature. For example, a 300 series stainless steel structure operating at elevated temperature, in all probability, would satisfy the ASME Boiler and Pressure Vessel Code criteria if the linearized elastic stress range is less than three times the material yield strength. However, at higher elastic stress ranges it is difficult to judge, a priori, that a structural component would comply with inelastic Code criteria after a detailed inelastic analysis. The purpose of this paper is to illustrate that it is not the elastic stress range but the stress intensities at specific times during a thermal transient which provide a better insight into the inelastic response of the structure. The specific example of the CRBRP flued head design demonstrates that the temperature differential between various parts of the structure can be changed by modifying the insulation pattern and heat flow path in the structure, without significantly altering the elastic stress range over a complete load cycle. However, the modified design did reduce the stress intensity during steady state elevated temperature operation. This modified design satisfied the inelastic Code criteria whereas the initial design failed to comply with the strain accumulation criterion

  16. Visit safety

    CERN Document Server

    2012-01-01

    Experiment areas, offices, workshops: it is possible to have co-workers or friends visit these places.     You already know about the official visits service, the VIP office, and professional visits. But do you know about the safety instruction GSI-OHS1, “Visits on the CERN site”? This is a mandatory General Safety Instruction that was created to assist you in ensuring safety for all your visits, whatever their nature—especially those that are non-official. Questions? The HSE Unit will be happy to answer them. Write to safety-general@cern.ch.   The HSE Unit

  17. Monitoring circuit for reactor safety systems

    Science.gov (United States)

    Keefe, Donald J.

    1976-01-01

    The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned.

  18. Monitoring circuit for reactor safety systems

    International Nuclear Information System (INIS)

    Keefe, D.J.

    1976-01-01

    The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned. 3 claims, 2 figures

  19. Qualification criteria to certify a package for air transport of plutonium

    International Nuclear Information System (INIS)

    1977-12-01

    The document describes qualification criteria developed by the U.S. Nuclear Regulatory Commission to certify a package for air transport of plutonium. Included in the document is a discussion of aircraft accident conditions and a summary of the technical basis for the qualification criteria. The criteria require prototype packages to be subjected to various individual and sequential tests that simulate the conditions produced in severe aircraft accidents. Specific post-test acceptance standards are prescribed for each of the three safety functions of a package. The qualification criteria also prescribe certain operational controls to be exercised during transport

  20. On the consistency of risk acceptance criteria with normative theories for decision-making

    Energy Technology Data Exchange (ETDEWEB)

    Abrahamsen, E.B. [University of Stavanger, 4036 Stavanger (Norway)], E-mail: eirik.abrahamsen@uis.no; Aven, T. [University of Stavanger, 4036 Stavanger (Norway)

    2008-12-15

    In evaluation of safety in projects it is common to use risk acceptance criteria to support decision-making. In this paper, we discuss to what extent the risk acceptance criteria is in accordance with the normative theoretical framework of the expected utility theory and the rank-dependent utility theory. We show that the use of risk acceptance criteria may violate the independence axiom of the expected utility theory and the comonotonic independence axiom of the rank-dependent utility theory. Hence the use of risk acceptance criteria is not in general consistent with these theories. The level of inconsistency is highest for the expected utility theory.

  1. On the consistency of risk acceptance criteria with normative theories for decision-making

    International Nuclear Information System (INIS)

    Abrahamsen, E.B.; Aven, T.

    2008-01-01

    In evaluation of safety in projects it is common to use risk acceptance criteria to support decision-making. In this paper, we discuss to what extent the risk acceptance criteria is in accordance with the normative theoretical framework of the expected utility theory and the rank-dependent utility theory. We show that the use of risk acceptance criteria may violate the independence axiom of the expected utility theory and the comonotonic independence axiom of the rank-dependent utility theory. Hence the use of risk acceptance criteria is not in general consistent with these theories. The level of inconsistency is highest for the expected utility theory

  2. Nuclear Safety Bureau: safety objectives and principles for the proposed ANSTO reactor

    International Nuclear Information System (INIS)

    Westall, D.

    1993-01-01

    Siting criteria and safety assessment principles were previously promulgated by the Australian Atomic Energy Commission (AAEC), and have been applied by ANSTO and the Nuclear Safety Bureau (NSB). The NSB is revising these criteria and principles to take account of evolving nuclear safety standards and practices. The NSB Safety and Siting Assessment Principles (SSAP) are presented and it is estimated that it will provide a comprehensive basis for the safety assessment of research reactors in Australia, and be applicable to all stages of a reactor project: siting: design and construction; operation; modification; and decommissioning. The SSAP are similar to the principles promulgated by the AAEC, in that probabilistic safety criteria are set for assessment of design, however these criteria are complimentary to a deterministic design basis approach. This is a similar approach to that recently published by the UK Nuclear Installations Inspectorate 4 . Siting principles are now also included, where they were previously separate, and require a consideration of the consequences of severe accidents which are an extension of accidents catered for by the design of the plant. Criteria for radiation doses due to normal operations and design basis accidents are included in the principles for safety assessment. 9 refs

  3. Assessment of Clinical Criteria for Sepsis

    Science.gov (United States)

    Seymour, Christopher W.; Liu, Vincent X.; Iwashyna, Theodore J.; Brunkhorst, Frank M.; Rea, Thomas D.; Scherag, André; Rubenfeld, Gordon; Kahn, Jeremy M.; Shankar-Hari, Manu; Singer, Mervyn; Deutschman, Clifford S.; Escobar, Gabriel J.; Angus, Derek C.

    2016-01-01

    IMPORTANCE The Third International Consensus Definitions Task Force defined sepsis as “life-threatening organ dysfunction due to a dysregulated host response to infection.” The performance of clinical criteria for this sepsis definition is unknown. OBJECTIVE To evaluate the validity of clinical criteria to identify patients with suspected infection who are at risk of sepsis. DESIGN, SETTINGS, AND POPULATION Among 1.3 million electronic health record encounters from January 1, 2010, to December 31, 2012, at 12 hospitals in southwestern Pennsylvania, we identified those with suspected infection in whom to compare criteria. Confirmatory analyses were performed in 4 data sets of 706 399 out-of-hospital and hospital encounters at 165 US and non-US hospitals ranging from January 1, 2008, until December 31, 2013. EXPOSURES Sequential [Sepsis-related] Organ Failure Assessment (SOFA) score, systemic inflammatory response syndrome (SIRS) criteria, Logistic Organ Dysfunction System (LODS) score, and a new model derived using multivariable logistic regression in a split sample, the quick Sequential [Sepsis-related] Organ Failure Assessment (qSOFA) score (range, 0–3 points, with 1 point each for systolic hypotension [≤100 mm Hg], tachypnea [≥22/min], or altered mentation). MAIN OUTCOMES AND MEASURES For construct validity, pairwise agreement was assessed. For predictive validity, the discrimination for outcomes (primary: in-hospital mortality; secondary: in-hospital mortality or intensive care unit [ICU] length of stay ≥3 days) more common in sepsis than uncomplicated infection was determined. Results were expressed as the fold change in outcome over deciles of baseline risk of death and area under the receiver operating characteristic curve (AUROC). RESULTS In the primary cohort, 148 907 encounters had suspected infection (n = 74 453 derivation; n = 74 454 validation), of whom 6347 (4%) died. Among ICU encounters in the validation cohort (n = 7932 with suspected

  4. Regulatory review of safety cases and safety assessments - associated challenges

    International Nuclear Information System (INIS)

    Bennett, D.G.; Ben Belfadhel, M.; Metcalf, P.E.

    2006-01-01

    Regulatory reviews of safety cases and safety assessments are essential for credible decision making on the licensing or authorization of radioactive waste disposal facilities. Regulatory review also plays an important role in developing the safety case and in establishing stakeholders' confidence in the safety of the facility. Reviews of safety cases for radioactive waste disposal facilities need to be conducted by suitably qualified and experienced staff, following systematic and well planned review processes. Regulatory reviews should be sufficiently comprehensive in their coverage of issues potentially affecting the safety of the disposal system, and should assess the safety case against clearly established criteria. The conclusions drawn from a regulatory review, and the rationale for them should be reproducible and documented in a transparent and traceable way. Many challenges are faced when conducting regulatory reviews of safety cases. Some of these relate to issues of project and programme management, and resources, while others derive from the inherent difficulties of assessing the potential long term future behaviour of engineered and environmental systems. The paper describes approaches to the conduct of regulatory reviews and discusses some of the challenges faced. (author)

  5. Reflexive criteria of sociological research

    Directory of Open Access Journals (Sweden)

    R T Ubaydullaeva

    2014-12-01

    Full Text Available The article is devoted to the sociological criteria of explaining the way of thinking and actions of subjects, their spiritual and moral positions and intellectual forces that form the laws of social life. The author seeks to adapt such categories as ‘meaning of life’, ‘human dignity’, ‘rationality’ etc. for the purposes of sociological analysis by methodological construction of some real life dichotomies such as ‘subjective meaning and social function’, ‘the real and the ideal’, ‘the demanded and the excluded’. Thus, the author studies economic, political and technical processes in terms of both positivity and negativity of social interaction and states that given the increasing differentiation of the society and the contradictory trends of social development the reflexive criteria that take into account the socio-cultural nature of the man help to find one’s own model of development.

  6. New facility shield design criteria

    International Nuclear Information System (INIS)

    Howell, W.P.

    1981-07-01

    The purpose of the criteria presented here is to provide standard guidance for the design of nuclear radiation shields thoughout new facilities. These criteria are required to assure a consistent and integrated design that can be operated safely and economically within the DOE standards. The scope of this report is confined to the consideration of radiation shielding for contained sources. The whole body dose limit established by the DOE applies to all doses which are generally distributed throughout the trunk of the body. Therefore, where the whole body is the critical organ for an internally deposited radionuclide, the whole body dose limit applies to the sum of doses received must assure control of the concentration of radionuclides in the building atmosphere and thereby limit the dose from internal sources

  7. Pickering seismic safety margin

    International Nuclear Information System (INIS)

    Ghobarah, A.; Heidebrecht, A.C.; Tso, W.K.

    1992-06-01

    A study was conducted to recommend a methodology for the seismic safety margin review of existing Canadian CANDU nuclear generating stations such as Pickering A. The purpose of the seismic safety margin review is to determine whether the nuclear plant has sufficient seismic safety margin over its design basis to assure plant safety. In this review process, it is possible to identify the weak links which might limit the seismic performance of critical structures, systems and components. The proposed methodology is a modification the EPRI (Electric Power Research Institute) approach. The methodology includes: the characterization of the site margin earthquake, the definition of the performance criteria for the elements of a success path, and the determination of the seismic withstand capacity. It is proposed that the margin earthquake be established on the basis of using historical records and the regional seismo-tectonic and site specific evaluations. The ability of the components and systems to withstand the margin earthquake is determined by database comparisons, inspection, analysis or testing. An implementation plan for the application of the methodology to the Pickering A NGS is prepared

  8. Design provisions for safety

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1983-01-01

    Design provisions for safety of nuclear power plants are based on a well balanced concept: the public is protected against a release of radioactive material by multiple barriers. These barriers are protected according to a 'defence-in-depth' principle. The reactor safety concept is primarily aimed at the prevention of accidents, especially fuel damage. Additionally, measures for consequence limitation are provided in order to prevent a severe release of radioactivity to the environment. However, it is difficult to judge the overall effectiveness of such devices. In a comprehensive safety analysis it has to be shown that the protection systems and safeguards work with sufficient reliability in the event of an accident. For the reliability assessment deterministic criteria (single failure, redundancy, fail-safe, demand for diversity) play an important role. Increasing efforts have been made to assess reliability quantitatively by means of probabilistic methods. It is now usual to perform reliability analyses of essential systems of nuclear power plants in the course of licensing procedures. As an additional level of emergency measures for a further reduction of hazards a reasonable amount of accident information has to be transferred. Operational experience may be considered as an important feedback to the design of plant safety features. Operator training has to include, besides skill in performing of operating procedures, the training of a flexible response to different accident situations. Experience has shown that the design provisions for safety could prevent dangerous release of the radioactive material to the environment after an accident has occurred. For future developments of reactor safety, extensive analyses of operating experience are of great importance. The main goal should be to enhance the reliability of measures for accident prevention, which prevent the core from meltdown or other damages

  9. Safety Principles

    Directory of Open Access Journals (Sweden)

    V. A. Grinenko

    2011-06-01

    Full Text Available The offered material in the article is picked up so that the reader could have a complete representation about concept “safety”, intrinsic characteristics and formalization possibilities. Principles and possible strategy of safety are considered. A material of the article is destined for the experts who are taking up the problems of safety.

  10. Safety Systems

    Science.gov (United States)

    Halligan, Tom

    2009-01-01

    Colleges across the country are rising to the task by implementing safety programs, response strategies, and technologies intended to create a secure environment for teachers and students. Whether it is preparing and responding to a natural disaster, health emergency, or act of violence, more schools are making campus safety a top priority. At…

  11. Safety First

    Science.gov (United States)

    Taft, Darryl

    2011-01-01

    Ned Miller does not take security lightly. As director of campus safety and emergency management at the Des Moines Area Community College (DMACC), any threat requires serious consideration. As community college administrators adopt a more proactive approach to campus safety, many institutions are experimenting with emerging technologies, including…

  12. NASA balloon design and flight - Philosophy and criteria

    Science.gov (United States)

    Smith, I. S., Jr.

    1993-01-01

    The NASA philosophy and criteria for the design and flight of scientific balloons are set forth and discussed. The thickness of balloon films is standardized at 20.3 microns to isolate potential film problems, and design equations are given for specific balloon parameters. Expressions are given for: flight-stress index, total required thickness, cap length, load-tape rating, and venting-duct area. The balloon design criteria were used in the design of scientific balloons under NASA auspices since 1986, and the resulting designs are shown to be 95 percent effective. These results represent a significant increase in the effectiveness of the balloons and therefore indicate that the design criteria are valuable. The criteria are applicable to four balloon volume classes in combination with seven payload ranges.

  13. Study on dilatation of multi-criteria evaluation method (II)

    International Nuclear Information System (INIS)

    Tabaru, Yasuhiko; Tomizawa, Masao; Sasaki, Shigeo

    2003-01-01

    In the study on FBR-cycle practical application strategy conducted by JNC, as part of development of evaluation system aiming at comparative evaluation of promising concept for FBR-cycle system, they grade the value of the concepts under the criteria evaluation such as economical efficiency and environmental load. In order that this system functions effectively in selecting promising concepts, we believe that it is important to extend the range of criteria evaluation and improve objectivity and persuasiveness of it. This is why since the last fiscal year we have been studying on evaluation methodology of and investigation examples on external economical efficiency (effects on the environment and human health, safety, energy security, nuclear non-proliferation, etc.) relevant to introduction of FBR, which had not been included in the conventional evaluation of economical efficiency. In this work, we have especially focused on the external economical efficiency relevant to energy security which is peculiar to FBR-cycle and studied on its evaluation methodology and investigation examples. Firstly, we summarized up on the concept and current situation of energy security and the position of nuclear energy in energy security. Then we identified the necessity of clarifying the importance of energy security with the middle-term point of view allowing for deficiency of fossil or uranium resources, and also the importance of the role of FBR as an improvement action for it. Secondly, we studied on the current energy economic model and examined the possibility of applying energy security for quantitative evaluation. As a result, we have concluded that the general equilibrium GTAP (Global Trade Analysis Project) in which fossil resource market around the world is modeled should be effective for quantitative evaluation of long term energy security. Finally, assuming that we will conduct the quantitative evaluation of long term energy security using GTAP model in the future, we

  14. Unmanned Aerial Vehicle Mishap Taxonomy for Range Safety Reviews

    Science.gov (United States)

    2016-02-01

    Wind /Turbulence ................................................................................................. 5-3 5.1.3 Rain...5.1.8 Bird strike............................................................................................................. 5-5 5.2 Radio...intervals were mentioned in reports as corrective actions for these scenarios. One instance of fuel nozzle failure in a turbine -powered UAV resulted in

  15. Evaluating Dependence Criteria for Caffeine

    OpenAIRE

    Striley, Catherine L.W.; Griffiths, Roland R.; Cottler, Linda B.

    2011-01-01

    Background: Although caffeine is the most widely used mood-altering drug in the world, few studies have operationalized and characterized Diagnostic and Statistical Manual IV (DSM-IV) substance dependence criteria applied to caffeine. Methods: As a part of a nosological study of substance use disorders funded by the National Institute on Drug Abuse, we assessed caffeine use and dependence symptoms among high school and college students, drug treatment patients, and pain clinic patients who re...

  16. Sampling criteria in multicollection searching.

    Science.gov (United States)

    Gilio, A.; Scozzafava, R.; Marchetti, P. G.

    In the first stage of the document retrieval process, no information concerning relevance of a particular document is available. On the other hand, computer implementation requires that the analysis be made only for a sample of retrieved documents. This paper addresses the significance and suitability of two different sampling criteria for a multicollection online search facility. The inevitability of resorting to a logarithmic criterion in order to achieve a "spread of representativeness" from the multicollection is demonstrated.

  17. Position paper: Seismic design criteria

    International Nuclear Information System (INIS)

    Farnworth, S.K.

    1995-01-01

    The purpose of this paper is to document the seismic design criteria to be used on the Title 11 design of the underground double-shell waste storage tanks and appurtenant facilities of the Multi-Function Waste Tank Facility (MWTF) project, and to provide the history and methodologies for determining the recommended Design Basis Earthquake (DBE) Peak Ground Acceleration (PGA) anchors for site-specific seismic response spectra curves. Response spectra curves for use in design are provided in Appendix A

  18. Safety culture at ANAV

    International Nuclear Information System (INIS)

    Fernandez Madrid, B.

    2010-01-01

    Recent safety culture assessments detected various actions, practices and behaviours that did not follow the standards, expectations and guidelines that are essential l for all safe and highly reliable companies, as we aim to be. For this reason, as part of the PROCURA project, a wide range of actions have been undertaken o reinforce certain individual, group and organisational behaviours. (Author).

  19. Repository operational criteria comparative analysis

    International Nuclear Information System (INIS)

    Hageman, J.P.; Chowdhury, A.H.

    1994-06-01

    The objective of the ''Repository Operational Criteria (ROC) Feasibility Studies'' (or ROC task) was to conduct comprehensive and integrated analyses of repository design, construction, and operations criteria in 10 CFR Part 60 regulations considering the interfaces among the components of the regulations and impacts of any potential changes to those regulations. The ROC task addresses regulatory criteria and uncertainties related to the preclosure aspects of the geologic repository. Those parts of 10 CFR Part 60 that require routine guidance or minor changes to the rule were addressed in Hageman and Chowdhury, 1992. The ROC task shows a possible need for further regulatory clarity, by major changes to the rule, related to the design bases and siting of a geologic repository operations area and radiological emergency planning in order to assure defense-in-depth. The analyses, presented in this report, resulted in the development and refinement of regulatory concepts and their supporting rationale for recommendations for potential major changes to 10 CFR Pan 0 regulations

  20. Compressive laser ranging.

    Science.gov (United States)

    Babbitt, Wm Randall; Barber, Zeb W; Renner, Christoffer

    2011-12-15

    Compressive sampling has been previously proposed as a technique for sampling radar returns and determining sparse range profiles with a reduced number of measurements compared to conventional techniques. By employing modulation on both transmission and reception, compressive sensing in ranging is extended to the direct measurement of range profiles without intermediate measurement of the return waveform. This compressive ranging approach enables the use of pseudorandom binary transmit waveforms and return modulation, along with low-bandwidth optical detectors to yield high-resolution ranging information. A proof-of-concept experiment is presented. With currently available compact, off-the-shelf electronics and photonics, such as high data rate binary pattern generators and high-bandwidth digital optical modulators, compressive laser ranging can readily achieve subcentimeter resolution in a compact, lightweight package.