WorldWideScience

Sample records for safety modeling requirements

  1. Specification of advanced safety modeling requirements (Rev. 0)

    International Nuclear Information System (INIS)

    Fanning, T. H.; Tautges, T. J.

    2008-01-01

    The U.S. Department of Energy's Global Nuclear Energy Partnership has lead to renewed interest in liquid-metal-cooled fast reactors for the purpose of closing the nuclear fuel cycle and making more efficient use of future repository capacity. However, the U.S. has not designed or constructed a fast reactor in nearly 30 years. Accurate, high-fidelity, whole-plant dynamics safety simulations will play a crucial role by providing confidence that component and system designs will satisfy established design limits and safety margins under a wide variety of operational, design basis, and beyond design basis transient conditions. Current modeling capabilities for fast reactor safety analyses have resulted from several hundred person-years of code development effort supported by experimental validation. The broad spectrum of mechanistic and phenomenological models that have been developed represent an enormous amount of institutional knowledge that needs to be maintained. Complicating this, the existing code architectures for safety modeling evolved from programming practices of the 1970s. This has lead to monolithic applications with interdependent data models which require significant knowledge of the complexities of the entire code in order for each component to be maintained. In order to develop an advanced fast reactor safety modeling capability, the limitations of the existing code architecture must be overcome while preserving the capabilities that already exist. To accomplish this, a set of advanced safety modeling requirements is defined, based on modern programming practices, that focuses on modular development within a flexible coupling framework. An approach for integrating the existing capabilities of the SAS4A/SASSYS-1 fast reactor safety analysis code into the SHARP framework is provided in order to preserve existing capabilities while providing a smooth transition to advanced modeling capabilities. In doing this, the advanced fast reactor safety models will

  2. Specification of advanced safety modeling requirements (Rev. 0).

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, T. H.; Tautges, T. J.

    2008-06-30

    The U.S. Department of Energy's Global Nuclear Energy Partnership has lead to renewed interest in liquid-metal-cooled fast reactors for the purpose of closing the nuclear fuel cycle and making more efficient use of future repository capacity. However, the U.S. has not designed or constructed a fast reactor in nearly 30 years. Accurate, high-fidelity, whole-plant dynamics safety simulations will play a crucial role by providing confidence that component and system designs will satisfy established design limits and safety margins under a wide variety of operational, design basis, and beyond design basis transient conditions. Current modeling capabilities for fast reactor safety analyses have resulted from several hundred person-years of code development effort supported by experimental validation. The broad spectrum of mechanistic and phenomenological models that have been developed represent an enormous amount of institutional knowledge that needs to be maintained. Complicating this, the existing code architectures for safety modeling evolved from programming practices of the 1970s. This has lead to monolithic applications with interdependent data models which require significant knowledge of the complexities of the entire code in order for each component to be maintained. In order to develop an advanced fast reactor safety modeling capability, the limitations of the existing code architecture must be overcome while preserving the capabilities that already exist. To accomplish this, a set of advanced safety modeling requirements is defined, based on modern programming practices, that focuses on modular development within a flexible coupling framework. An approach for integrating the existing capabilities of the SAS4A/SASSYS-1 fast reactor safety analysis code into the SHARP framework is provided in order to preserve existing capabilities while providing a smooth transition to advanced modeling capabilities. In doing this, the advanced fast reactor safety models

  3. Modeling of requirement specification for safety critical real time computer system using formal mathematical specifications

    International Nuclear Information System (INIS)

    Sankar, Bindu; Sasidhar Rao, B.; Ilango Sambasivam, S.; Swaminathan, P.

    2002-01-01

    Full text: Real time computer systems are increasingly used for safety critical supervision and control of nuclear reactors. Typical application areas are supervision of reactor core against coolant flow blockage, supervision of clad hot spot, supervision of undesirable power excursion, power control and control logic for fuel handling systems. The most frequent cause of fault in safety critical real time computer system is traced to fuzziness in requirement specification. To ensure the specified safety, it is necessary to model the requirement specification of safety critical real time computer systems using formal mathematical methods. Modeling eliminates the fuzziness in the requirement specification and also helps to prepare the verification and validation schemes. Test data can be easily designed from the model of the requirement specification. Z and B are the popular languages used for modeling the requirement specification. A typical safety critical real time computer system for supervising the reactor core of prototype fast breeder reactor (PFBR) against flow blockage is taken as case study. Modeling techniques and the actual model are explained in detail. The advantages of modeling for ensuring the safety are summarized

  4. Safety Culture: A Requirement for New Business Models — Lessons Learned from Other High Risk Industries

    International Nuclear Information System (INIS)

    Kecklund, L.

    2016-01-01

    Technical development and changes on global markets affects all high risk industries creating opportunities as well as risks related to the achievement of safety and business goals. Changes in legal and regulatory frameworks as well as in market demands create a need for major changes. Several high risk industries are facing a situation where they have to develop new business models. Within the transportation domain, e.g., aviation and railways, there is a growing concern related to how the new business models may affects safety issues. New business models in aviation and railways include extensive use of outsourcing and subcontractors to reduce costs resulting in, e.g., negative changes in working conditions, work hours, employment conditions and high turnover rates. The energy sector also faces pressures to create new business models for transition to renewable energy production to comply with new legal and regulatory requirements and to make best use of new reactor designs. In addition, large scale phase out and decommissioning of nuclear facilities have to be managed by the nuclear industry. Some negative effects of new business models have already arisen within the transportation domain, e.g., the negative effects of extensive outsourcing and subcontractor use. In the railway domain the infrastructure manager is required by European and national regulations to assure that all subcontractors are working according to the requirements in the infrastructure managers SMS (Safety Management System). More than ten levels of subcontracts can be working in a major infrastructure project making the system highly complex and thus difficult to control. In the aviation domain, tightly coupled interacting computer networks supplying airport services, as well as air traffic control, are managed and maintained by several different companies creating numerous interfaces which must be managed by the SMS. There are examples where a business model with several low

  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. Safety of Research Reactors. Safety Requirements

    International Nuclear Information System (INIS)

    2010-01-01

    The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

  7. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

    Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...

  8. Modeling the Non-functional Requirements in the Context of Usability, Performance, Safety and Security

    OpenAIRE

    Sadiq, Mazhar

    2007-01-01

    Requirement engineering is the most significant part of the software development life cycle. Until now great emphasis has been put on the maturity of the functional requirements. But with the passage of time it reveals that the success of software development does not only pertain to the functional requirements rather non-functional requirements should also be taken into consideration. Among the non-functional requirements usability, performance, safety and security are considered important. ...

  9. 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.)

  10. The Nuremberg Code subverts human health and safety by requiring animal modeling

    Directory of Open Access Journals (Sweden)

    Greek Ray

    2012-07-01

    Full Text Available Abstract Background The requirement that animals be used in research and testing in order to protect humans was formalized in the Nuremberg Code and subsequent national and international laws, codes, and declarations. Discussion We review the history of these requirements and contrast what was known via science about animal models then with what is known now. We further analyze the predictive value of animal models when used as test subjects for human response to drugs and disease. We explore the use of animals for models in toxicity testing as an example of the problem with using animal models. Summary We conclude that the requirements for animal testing found in the Nuremberg Code were based on scientifically outdated principles, compromised by people with a vested interest in animal experimentation, serve no useful function, increase the cost of drug development, and prevent otherwise safe and efficacious drugs and therapies from being implemented.

  11. Traceability of Software Safety Requirements in Legacy Safety Critical Systems

    Science.gov (United States)

    Hill, Janice L.

    2007-01-01

    How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

  12. Software safety analysis on the model specified by NuSCR and SMV input language at requirements phase of software development life cycle using SMV

    International Nuclear Information System (INIS)

    Koh, Kwang Yong; Seong, Poong Hyun

    2005-01-01

    Safety-critical software process is composed of development process, verification and validation (V and V) process and safety analysis process. Safety analysis process has been often treated as an additional process and not found in a conventional software process. But software safety analysis (SSA) is required if software is applied to a safety system, and the SSA shall be performed independently for the safety software through software development life cycle (SDLC). Of all the phases in software development, requirements engineering is generally considered to play the most critical role in determining the overall software quality. NASA data demonstrate that nearly 75% of failures found in operational software were caused by errors in the requirements. The verification process in requirements phase checks the correctness of software requirements specification, and the safety analysis process analyzes the safety-related properties in detail. In this paper, the method for safety analysis at requirements phase of software development life cycle using symbolic model verifier (SMV) is proposed. Hazard is discovered by hazard analysis and in other to use SMV for the safety analysis, the safety-related properties are expressed by computation tree logic (CTL)

  13. Safety of Nuclear Power Plants: Design. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  14. Strategy of severe accident physical modeling in view of recent requirements to safety analysis

    International Nuclear Information System (INIS)

    Bolshov, L.A.

    1994-01-01

    Nuclear power destiny in various states including Russia is not free from questions. Where there is plenty of non-expensive natural gas or coal in a country, the competition of nuclear power with other power sources is especially intense. Until one considers the economic efficiency or environmental impact of the normally operating plant, the estimate of the proponents favorite choice may be rather optimistic in many cases. As soon as safety aspects of nuclear power are concerned it is necessary to answer very significant questions about the dangers resulting from severe accidents. TMI and, to a greater extent, Chernobyl, demonstrated the other aspect of the severe accident problem. It serves no purpose to dwell upon the inadequate reaction of the population on the radiation problem. It is of little use to try to prove that the health consequences of the Chernobyl or some other radiation accident are substantially overestimated. To make an advance one must substantially reduce the severe accident risk. Besides that is is necessary to give a convincing proof that such a reduction has really been made

  15. Phase Two Feasibility Study for Software Safety Requirements Analysis Using Model Checking

    Science.gov (United States)

    Turgeon, Gregory; Price, Petra

    2010-01-01

    A feasibility study was performed on a representative aerospace system to determine the following: (1) the benefits and limitations to using SCADE , a commercially available tool for model checking, in comparison to using a proprietary tool that was studied previously [1] and (2) metrics for performing the model checking and for assessing the findings. This study was performed independently of the development task by a group unfamiliar with the system, providing a fresh, external perspective free from development bias.

  16. Safety of magnetic fusion facilities: Requirements

    International Nuclear Information System (INIS)

    1996-05-01

    This Standard identifies safety requirements for magnetic fusion facilities. Safety functions are used to define outcomes that must be achieved to ensure that exposures to radiation, hazardous materials, or other hazards are maintained within acceptable limits. Requirements applicable to magnetic fusion facilities have been derived from Federal law, policy, and other documents. In addition to specific safety requirements, broad direction is given in the form of safety principles that are to be implemented and within which safety can be achieved

  17. Safety of Research Reactors. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This Safety Requirements publication establishes requirements for all main areas of safety for research reactors, with particular emphasis on requirements for design and operation. It explains the safety objectives and concepts that form the basis for safety and safety assessment for all stages in the lifetime of a research reactor. Technical and administrative requirements for the safety of new research reactors are established in accordance with these objectives and concepts, and they are to be applied to the extent practicable for existing research reactors. The safety requirements established in this publication for the management of safety and regulatory supervision apply to site evaluation, design, manufacturing, construction, commissioning, operation (including utilization and modification), and planning for decommissioning of research reactors (including critical assemblies and subcritical assemblies). The publication is intended for use by regulatory bodies and other organizations with responsibilities in these areas and in safety analysis, verification and review, and the provision of technical support.

  18. A total safety management model

    International Nuclear Information System (INIS)

    Obadia, I.J.; Vidal, M.C.R.; Melo, P.F.F.F.

    2002-01-01

    In nuclear organizations, quality and safety are inextricably linked. Therefore, the search for excellence means reaching excellence in nuclear safety. The International Atomic Energy Agency, IAEA, developed, after the Chernobyl accident, the organizational approach for improving nuclear safety based on the safety culture, which requires a framework necessary to provide modifications in personnel attitudes and behaviors in situations related to safety. This work presents a Total Safety Management Model, based on the Model of Excellence of the Brazilian Quality Award and on the safety culture approach, which represents an alternative to this framework. The Model is currently under validation at the Nuclear Engineering Institute, in Rio de Janeiro, Brazil, and the results of its initial safety culture self assessment are also presented and discussed. (author)

  19. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  20. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2003-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  1. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations; to be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources; and to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  2. Leadership and Management for Safety. General Safety Requirements (Spanish Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    his Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

  3. Leadership and Management for Safety. General Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

  4. Leadership and Management for Safety. General Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

  5. Leadership and Management for Safety. General Safety Requirements (Chinese Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factors, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations and other organizations concerned with facilities and activities that give rise to radiation risks.

  6. Leadership and Management for Safety. General Safety Requirements

    International Nuclear Information System (INIS)

    2016-01-01

    This Safety Requirements publication establishes requirements that support Principle 3 of the Fundamental Safety Principles in relation to establishing, sustaining and continuously improving leadership and management for safety and an integrated management system. It emphasizes that leadership for safety, management for safety, an effective management system and a systemic approach (i.e. an approach in which interactions between technical, human and organizational factors are duly considered) are all essential to the specification and application of adequate safety measures and to the fostering of a strong safety culture. Leadership and an effective management system will integrate safety, health, environmental, security, quality, human-and-organizational factor, societal and economic elements. The management system will ensure the fostering of a strong safety culture, regular assessment of performance and the application of lessons from experience. The publication is intended for use by regulatory bodies, operating organizations (registrants and licensees) and other organizations concerned with facilities and activities that give rise to radiation risks

  7. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  8. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  9. Requirements of safety and reliability

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1977-01-01

    The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the findings derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essential for accident analyses, and the determination of the loads occuring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig./HP) [de

  10. Disposal of Radioactive Waste. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2011-01-01

    This publication establishes requirements applicable to all types of radioactive waste disposal facility. It is linked to the fundamental safety principles for each disposal option and establishes a set of strategic requirements that must be in place before facilities are developed. Consideration is also given to the safety of existing facilities developed prior to the establishment of present day standards. The requirements will be complemented by Safety Guides that will provide guidance on good practice for meeting the requirements for different types of waste disposal facility. Contents: 1. Introduction; 2. Protection of people and the environment; 3. Safety requirements for planning for the disposal of radioactive waste; 4. Requirements for the development, operation and closure of a disposal facility; 5. Assurance of safety; 6. Existing disposal facilities; Appendices.

  11. Radiation safety requirements for radionuclide laboratories

    International Nuclear Information System (INIS)

    1993-01-01

    In accordance with the section 26 of the Finnish Radiation Act (592/91) the safety requirements to be taken into account in planning laboratories and other premises, which affect safety in the use of radioactive materials, are confirmed by the Finnish Centre for Radiation and Nuclear Safety. The guide specifies the requirements for laboratories and storage rooms in which radioactive materials are used or stored as unsealed sources. There are also some general instructions concerning work procedures in a radionuclide laboratory

  12. Safety design requirements for safety systems and components of JSFR

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

    2011-01-01

    Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

  13. Site evaluation for nuclear installations. Safety requirements

    International Nuclear Information System (INIS)

    2003-01-01

    This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Siting, which was issued in 1988 as Safety Series No. 50-C-S (Rev. 1). It takes account of developments relating to site evaluations for nuclear installations since the Code on Siting was last revised. These developments include the issuing of the Safety Fundamentals publication on The Safety of Nuclear Installations, and the revision of various safety standards and other publications relating to safety. Requirements for site evaluation are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear installations. It is recognized that there are steady advances in technology and scientific knowledge, in nuclear safety and in what is considered adequate protection. Safety requirements change with these advances and this publication reflects the present consensus among States. This Safety Requirements publication was prepared under the IAEA programme on safety standards for nuclear installations. It establishes requirements and provides criteria for ensuring safety in site evaluation for nuclear installations. The Safety Guides on site evaluation listed in the references provide recommendations on how to meet the requirements established in this Safety Requirements publication. The objective of this publication is to establish the requirements for the elements of a site evaluation for a nuclear installation so as to characterize fully the site specific conditions pertinent to the safety of a nuclear installation. The purpose is to establish requirements for criteria, to be applied as appropriate to site and site-installation interaction in operational states and accident conditions, including those that could lead to emergency measures for: (a) Defining the extent of information on a proposed site to be presented by the applicant; (b) Evaluating a proposed site to ensure that the site

  14. Safety of nuclear fuel cycle facilities. Safety requirements

    International Nuclear Information System (INIS)

    2008-01-01

    This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific reference include aspects of nuclear fuel generation, storage, reprocessing and disposal. Contents: 1. Introduction; 2. The safety objective, concepts and safety principles; 3. Legal framework and regulatory supervision; 4. The management system and verification of safety; 5. Siting of the facility; 6. Design of the facility; 7. Construction of the facility; 8. Commissioning of the facility; 9. Operation of the facility; 10. Decommissioning of the facility; Appendix I: Requirements specific to uranium fuel fabrication facilities; Appendix II: Requirements specific to mixed oxide fuel fabrication facilities; Appendix III: Requirements specific to conversion facilities and enrichment facilities

  15. Meeting the maglev system's safety requirements

    Energy Technology Data Exchange (ETDEWEB)

    Pierick, K

    1983-12-01

    The author shows how the safety requirements of the maglev track system derive from the general legal conditions for the safety of tracked transport. It is described how their compliance beyond the so-called ''development-accompanying'' and ''acceptance-preparatory'' safety work can be assured for the Transrapid test layout (TVE) now building in Emsland and also for later application as public transport system in Germany within the meaning of the General Railway Act.

  16. Public requirement to demonstrate safety

    International Nuclear Information System (INIS)

    Green, P.

    1991-01-01

    To many working within Government or industry, public concern over the disposal of radioactive waste is misplaced and has arisen out of an irrational and unscientific fear of technology, or even science in general. Members of the public, it is argued, are concerned because they do not understand the size of the risk in question. From the industry's point of view, the risk arising from the disposal of radioactive waste is ''negligible when compared to other everyday risks of life. Furthermore, any public exposure that may arise, either soon after closure of a facility or in the far future would comply with internationally accepted safety standards. In this context, the continuing concern over disposal of radioactive waste is viewed as evidence of the irrational and unscientific attitude of the public. The assessment and regulation of risk from waste disposal therefore is presented as a purely scientific question. Some of these issues are examined and public concern is shown not to be irrational but to be based upon legitimate questions over current waste management policy. An important question is not just ''how safe is safe, but who decides and how?''. (Author)

  17. Modelling blood safety

    NARCIS (Netherlands)

    Janssen, M.P.

    2010-01-01

    This thesis describes the development and application of methods and models to support decision making on safety measures aimed at preventing the transmission of infections by blood donors. Safety measures refer to screening tests for blood donors, quarantine periods for blood plasma, or methods for

  18. Generic Safety Requirements for Developing Safe Insulin Pump Software

    Science.gov (United States)

    Zhang, Yi; Jetley, Raoul; Jones, Paul L; Ray, Arnab

    2011-01-01

    Background The authors previously introduced a highly abstract generic insulin infusion pump (GIIP) model that identified common features and hazards shared by most insulin pumps on the market. The aim of this article is to extend our previous work on the GIIP model by articulating safety requirements that address the identified GIIP hazards. These safety requirements can be validated by manufacturers, and may ultimately serve as a safety reference for insulin pump software. Together, these two publications can serve as a basis for discussing insulin pump safety in the diabetes community. Methods In our previous work, we established a generic insulin pump architecture that abstracts functions common to many insulin pumps currently on the market and near-future pump designs. We then carried out a preliminary hazard analysis based on this architecture that included consultations with many domain experts. Further consultation with domain experts resulted in the safety requirements used in the modeling work presented in this article. Results Generic safety requirements for the GIIP model are presented, as appropriate, in parameterized format to accommodate clinical practices or specific insulin pump criteria important to safe device performance. Conclusions We believe that there is considerable value in having the diabetes, academic, and manufacturing communities consider and discuss these generic safety requirements. We hope that the communities will extend and revise them, make them more representative and comprehensive, experiment with them, and use them as a means for assessing the safety of insulin pump software designs. One potential use of these requirements is to integrate them into model-based engineering (MBE) software development methods. We believe, based on our experiences, that implementing safety requirements using MBE methods holds promise in reducing design/implementation flaws in insulin pump development and evolutionary processes, therefore improving

  19. Canister Storage Building (CSB) Technical Safety Requirements

    International Nuclear Information System (INIS)

    KRAHN, D.E.

    2000-01-01

    The purpose of this section is to explain the meaning of logical connectors with specific examples. Logical connectors are used in Technical Safety Requirements (TSRs) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TSRs are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings

  20. Safety requirements applicable to the SMART design

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Kim, Wee Kyong; Kim, Hho Jung

    1999-01-01

    The 330 MW thermal power of integral reactor, named SMART (System integrated Modular Advanced ReacTor), is under development at KAERI for seawater desalination application and electricity generation. The final product of nuclear desalination plant (NDP) is electricity and fresh water. Thus, in addition to the protection of the public around the plant facility from the possible release of radioactive materials, the fresh water should be prevented from radioactivity contamination. In this study, to ensure the safety of SMART reactor in the early stage of design development, the safety requirements applicable to the SMART design were investigated, based on the current regulatory requirements for the existing NPPs and the advanced light water reactor (LWR) designs. The interface requirements related to the desalination facility were also investigated, based on the recent IAEA research activities pertaining to the NDP. As a result, it was found that the current regulatory requirements and guidance for the existing NPPs and advanced LWR designs are applicable to the SMART design and its safety evaluation. However, the safety requirements related to the SMART-specific design and the desalination plant are needed to develop in the future to assure the safety of the SMART reactor

  1. TWRS safety SSCs: Requirements and characteristics

    International Nuclear Information System (INIS)

    Smith-Fewell, M.A.

    1997-01-01

    Safety Systems, Structures, and Components (SSCs) have been identified from hazard and accident analyses. These analyses were performed to support the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR) and Basis for Interim Operation (BID). The text identifies and evaluates the SSCs and their supporting SSCs to show that they either prevent the occurrence of the accident or mitigate the consequences of the accident to below the acceptance guidelines. The requirements for the SSCs to fulfill these tasks are described

  2. Aviation Safety Simulation Model

    Science.gov (United States)

    Houser, Scott; Yackovetsky, Robert (Technical Monitor)

    2001-01-01

    The Aviation Safety Simulation Model is a software tool that enables users to configure a terrain, a flight path, and an aircraft and simulate the aircraft's flight along the path. The simulation monitors the aircraft's proximity to terrain obstructions, and reports when the aircraft violates accepted minimum distances from an obstruction. This model design facilitates future enhancements to address other flight safety issues, particularly air and runway traffic scenarios. This report shows the user how to build a simulation scenario and run it. It also explains the model's output.

  3. Geological disposal of radioactive waste. Safety requirements

    International Nuclear Information System (INIS)

    2006-01-01

    This Safety Requirements publication is concerned with providing protection to people and the environment from the hazards associated with waste management activities related to disposal, i.e. hazards that could arise during the operating period and following closure. It sets out the protection objectives and criteria for geological disposal and establishes the requirements that must be met to ensure the safety of this disposal option, consistent with the established principles of safety for radioactive waste management. It is intended for use by those involved in radioactive waste management and in making decisions in relation to the development, operation and closure of geological disposal facilities, especially those concerned with the related regulatory aspects. This publication contains 1. Introduction; 2. Protection of human health and the environment; 3. The safety requirements for geological disposal; 4. Requirements for the development, operation and closure of geological disposal facilities; Appendix: Assurance of compliance with the safety objective and criteria; Annex I: Geological disposal and the principles of radioactive waste management; Annex II: Principles of radioactive waste management

  4. Fire safety requirements for electrical cables towards nuclear reactor safety

    International Nuclear Information System (INIS)

    Raju, M.R.

    2002-01-01

    Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

  5. Safety of Nuclear Fuel Cycle Facilities. Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific relevance include aspects of nuclear fuel generation, storage, reprocessing and disposal

  6. Nuclear fuels with high burnup: safety requirements

    International Nuclear Information System (INIS)

    Phuc Tran Dai

    2016-01-01

    Vietnam authorities foresees to build 3 reactors from Russian design (VVER AES 2006) by 2030. In order to prepare the preliminary report on safety analysis the Vietnamese Agency for Radioprotection and Safety has launched an investigation on the behaviour of nuclear fuels at high burnups (up to 60 GWj/tU) that will be those of the new plants. This study deals mainly with the behaviour of the fuel assemblies in case of loss of coolant (LOCA). It appears that for an average burnup of 50 GWj/tU and for the advanced design of the fuel assembly (cladding and materials) safety requirements are fulfilled. For an average burnup of 60 GWj/tU, a list of issues remains to be assessed, among which the impact of clad bursting or the hydrogen embrittlement of the advanced zirconium alloys. (A.C.)

  7. Safety requirements for the Pu carriers

    International Nuclear Information System (INIS)

    Mishima, H.

    1993-01-01

    Ministry of Transport of Japan has now set about studying requirements for Pu carriers to ensure safety. It was first studied what the basic concept of safe carriage of Pu should be, and the basic ideas have been worked out. Next the requirements for the Pu carriers were studied based on the above. There are at present no international requirements of construction and equipment for the nuclear-material carriers, but MOT of Japan has so far required special construction and equipment for the nuclear-material carriers which carry a large amount of radioactive material, such as spent fuel or low level radioactive waste, corresponding to the level of the respective potential hazard. The requirements of construction and equipment of the Pu carriers have been established considering the difference in heat generation between Pu and spent fuel, physical protection, and so forth, in addition to the above basic concept. (J.P.N.)

  8. Nuclear safety review requirements for launch approval

    International Nuclear Information System (INIS)

    Sholtis, J.A. Jr.; Winchester, R.O.

    1992-01-01

    Use of nuclear power systems in space requires approval which is preceded by extensive safety analysis and review. This careful study allows an informed risk-benefit decision at the highest level of our government. This paper describes the process as it has historically been applied to U.S. isotopic power systems. The Ulysses mission, launched in October 1990, is used to illustrate the process. Expected variations to deal with reactor-power systems are explained

  9. Risk based limits for Operational Safety Requirements

    International Nuclear Information System (INIS)

    Cappucci, A.J. Jr.

    1993-01-01

    OSR limits are designed to protect the assumptions made in the facility safety analysis in order to preserve the safety envelope during facility operation. Normally, limits are set based on ''worst case conditions'' without regard to the likelihood (frequency) of a credible event occurring. In special cases where the accident analyses are based on ''time at risk'' arguments, it may be desirable to control the time at which the facility is at risk. A methodology has been developed to use OSR limits to control the source terms and the times these source terms would be available, thus controlling the acceptable risk to a nuclear process facility. The methodology defines a new term ''gram-days''. This term represents the area under a source term (inventory) vs time curve which represents the risk to the facility. Using the concept of gram-days (normalized to one year) allows the use of an accounting scheme to control the risk under the inventory vs time curve. The methodology results in at least three OSR limits: (1) control of the maximum inventory or source term, (2) control of the maximum gram-days for the period based on a source term weighted average, and (3) control of the maximum gram-days at the individual source term levels. Basing OSR limits on risk based safety analysis is feasible, and a basis for development of risk based limits is defensible. However, monitoring inventories and the frequencies required to maintain facility operation within the safety envelope may be complex and time consuming

  10. Technical safety requirements control level verification

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  11. Technical safety requirements control level verification; TOPICAL

    International Nuclear Information System (INIS)

    STEWART, J.L.

    1999-01-01

    A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL

  12. Generic safety documentation model

    International Nuclear Information System (INIS)

    Mahn, J.A.

    1994-04-01

    This document is intended to be a resource for preparers of safety documentation for Sandia National Laboratories, New Mexico facilities. It provides standardized discussions of some topics that are generic to most, if not all, Sandia/NM facilities safety documents. The material provides a ''core'' upon which to develop facility-specific safety documentation. The use of the information in this document will reduce the cost of safety document preparation and improve consistency of information

  13. ITER plasma safety interface models and assessments

    International Nuclear Information System (INIS)

    Uckan, N.A.; Bartels, H-W.; Honda, T.; Amano, T.; Boucher, D.; Post, D.; Wesley, J.

    1996-01-01

    Physics models and requirements to be used as a basis for safety analysis studies are developed and physics results motivated by safety considerations are presented for the ITER design. Physics specifications are provided for enveloping plasma dynamic events for Category I (operational event), Category II (likely event), and Category III (unlikely event). A safety analysis code SAFALY has been developed to investigate plasma anomaly events. The plasma response to ex-vessel component failure and machine response to plasma transients are considered

  14. 78 FR 46560 - Pipeline Safety: Class Location Requirements

    Science.gov (United States)

    2013-08-01

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Part... class location requirements for gas transmission pipelines. Section 5 of the Pipeline Safety, Regulatory... and, with respect to gas transmission pipeline facilities, whether applying IMP requirements to...

  15. Model-Driven Development of Safety Architectures

    Science.gov (United States)

    Denney, Ewen; Pai, Ganesh; Whiteside, Iain

    2017-01-01

    We describe the use of model-driven development for safety assurance of a pioneering NASA flight operation involving a fleet of small unmanned aircraft systems (sUAS) flying beyond visual line of sight. The central idea is to develop a safety architecture that provides the basis for risk assessment and visualization within a safety case, the formal justification of acceptable safety required by the aviation regulatory authority. A safety architecture is composed from a collection of bow tie diagrams (BTDs), a practical approach to manage safety risk by linking the identified hazards to the appropriate mitigation measures. The safety justification for a given unmanned aircraft system (UAS) operation can have many related BTDs. In practice, however, each BTD is independently developed, which poses challenges with respect to incremental development, maintaining consistency across different safety artifacts when changes occur, and in extracting and presenting stakeholder specific information relevant for decision making. We show how a safety architecture reconciles the various BTDs of a system, and, collectively, provide an overarching picture of system safety, by considering them as views of a unified model. We also show how it enables model-driven development of BTDs, replete with validations, transformations, and a range of views. Our approach, which we have implemented in our toolset, AdvoCATE, is illustrated with a running example drawn from a real UAS safety case. The models and some of the innovations described here were instrumental in successfully obtaining regulatory flight approval.

  16. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  17. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  18. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  19. Disposal of Radioactive Waste. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2011-01-01

    The IAEA's Statute authorizes the Agency to 'establish or adopt... standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the

  20. Tank Farms Technical Safety Requirements. Volume 1 and 2

    International Nuclear Information System (INIS)

    CASH, R.J.

    2000-01-01

    The Technical Safety Requirements (TSRs) define the acceptable conditions, safe boundaries, basis thereof, and controls to ensure safe operation during authorized activities, for facilities within the scope of the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR)

  1. Tank Farms Technical Safety Requirements [VOL 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    CASH, R.J.

    2000-12-28

    The Technical Safety Requirements (TSRs) define the acceptable conditions, safe boundaries, basis thereof, and controls to ensure safe operation during authorized activities, for facilities within the scope of the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR).

  2. Regulatory Safety Requirements for Operating Nuclear Installations

    International Nuclear Information System (INIS)

    Gubela, W.

    2017-01-01

    The National Nuclear Regulator (NNR) is established in terms of the National Nuclear Regulator Act (Act No 47 of 1999) and its mandate and authority are conferred through sections 5 and 7 of this Act, setting out the NNR's objectives and functions, which include exercising regulatory control over siting, design, construction etc of nuclear installations through the granting of nuclear authorisations. The NNR's responsibilities embrace all those actions aimed at providing the public with confidence and assurance that the risks arising from the production of nuclear energy remain within acceptable safety limits -> Therefore: Set fundamental safety standards, conducting pro-active safety assessments, determining licence conditions and obtaining assurance of compliance. The promotional aspects of nuclear activities in South Africa are legislated by the Nuclear Energy Act (Act No 46 of 1999). The NNR approach to regulations of nuclear safety and security take into consideration, amongst others, the potential hazards associated with the facility or activity, safety related programmes, the importance of the authorisation holder's safety related processes as well as the need to exercise regulatory control over the technical aspects such as of the design and operation of a nuclear facility in ensuring nuclear safety and security. South Africa does not have national nuclear industry codes and standards. The NNR is therefore non-prescriptive as it comes to the use of industry codes and standards. Regulatory framework (current) provide for the protection of persons, property, and environment against nuclear damage, through Licensing Process: Safety standards; Safety assessment; Authorisation and conditions of authorisation; Public participation process; Compliance assurance; Enforcement

  3. The main requirements of the International Basic Safety Standards

    International Nuclear Information System (INIS)

    Webb, G.A.M.

    1998-01-01

    The main requirements of the new international basic safety standards are discussed, including such topics as health effects of ionizing radiations, the revision of basic safety standards, the requirements for radiation protection practices, the requirements for intervention,and the field of regulatory infrastructures. (A.K.)

  4. Site safety requirements for high level waste disposal

    International Nuclear Information System (INIS)

    Chen Weiming; Wang Ju

    2006-01-01

    This paper outlines the content, status and trend of site safety requirements of International Atomic Energy Agency, America, France, Sweden, Finland and Japan. Site safety requirements are usually represented as advantageous vis-a-vis disadvantagous conditions, and potential advantage vis-a-vis disadvantage conditions, respectively in aspects of geohydrology, geochemistry, lithology, climate and human intrusion etc. Study framework and steps of site safety requirements for China are discussed under the view of systems science. (authors)

  5. Waste Encapsulation and Storage Facility interim operational safety requirements

    CERN Document Server

    Covey, L I

    2000-01-01

    The Interim Operational Safety Requirements (IOSRs) for the Waste Encapsulation and Storage Facility (WESF) define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt and inspection of cesium and strontium capsules from private irradiators; decontamination of the capsules and equipment; surveillance of the stored capsules; and maintenance activities. Controls required for public safety, significant defense-in-depth, significant worker safety, and for maintaining radiological consequences below risk evaluation guidelines (EGs) are included.

  6. Supplement to safety analysis report. 306-W building operations safety requirement

    International Nuclear Information System (INIS)

    Richey, C.R.

    1979-08-01

    The operations safety requirements (OSRs) presented in this report define the conditions, safe boundaries, and management control needed for safely conducting operations with radioactive materials in the Pacific Northwest Laboratory (PNL) 306-W building. The safety requirements are organized in five sections. Safety limits are safety-related process variables that are observable and measurable. Limiting conditions cover: equipment and technical conditions and characteristics of the facility and operations necessary for continued safe operation. Surveillance requirements prescribe the requirements for checking systems and components that are essential to safety. Equipment design controls require that changes to process equipment and systems be independently checked and approved to assure that the changes will have no adverse effect on safety. Administrative controls describe and discuss the organization and administrative systems and procedures to be used for safe operation of the facility. Details of the implementation of the operations safety requirements are prescribed by internal PNL documents such as criticality safety specifications and radiation work procedures

  7. Safety research needs for Russian-designed reactors. Requirements situation

    International Nuclear Information System (INIS)

    Brown, R. Allan; Holmstrom, Heikki; Reocreux, Michel; Schulz, Helmut; Liesch, Klaus; Santarossa, Giampiero; Hayamizu, Yoshitaka; Asmolov, Vladimir; Bolshov, Leonid; Strizhov, Valerii; Bougaenko, Sergei; Nikitin, Yuri N.; Proklov, Vladimir; Potapov, Alexandre; Kinnersly, Stephen R.; Voronin, Leonid M.; Honekamp, John R.; Frescura, Gianni M.; Maki, Nobuo; Reig, Javier; ); Bekjord, Eric S.; Rosinger, Herbert E.

    1998-01-01

    integrity must be verified, and material property data bases extended. - VVER severe accident research should focus on validation of codes for accident management procedures, and on extension and qualification of an appropriate data base for materials properties and their interactions. - RBMK thermal-hydraulic research is needed to improve the technical basis for further development of RBMK safety criteria. - Assessment of the integrity of the RBMK primary coolant circuit, and especially the fuel channel, requires urgent research. Methods of assessing RBMK pressure boundary integrity must be verified, and material property data bases extended. - RBMK severe accident research should focus on prevention of accidents and Accident Management for cases of loss of heat sink and Beyond Design-Basis Loss-of-Coolant Accidents. For these purposes, simple physical models and parametric codes need development and should be systematically used in plant specific analysis. Recommendations; - A Safety Research Strategic Plan should be developed. Such a plan sets goals, defines products, and describes when and how work will be done, including determination of research priorities. - Key players, including regulators, operators, plant designers and researchers should be involved in developing and implementing this plan and its execution and applying the results. - International cooperation in safety research should be encouraged for purposes of improving quality, preventing technical isolation and cost sharing. - New approaches, such as technical fora for specific technical topics, should be established to make safety research information in OECD countries available to researchers working on the safety of Russian-designed reactors

  8. Discussion of important safety requirements for new nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Lin; Jia Xiang; Yan Tianwen; Li Wenhong; Li Chun

    2014-01-01

    This paper presents the analysis of several important safety requirements and improvement direction. Technical view of security goals on site safety evaluation, internal and external events fortification, serious accident prevention and mitigation, as well as the core, containment system and instrument control system design and engineering optimization, and etc are indicated. It will be useful for new plant design, construction and safety improvement. (authors)

  9. Philosophy and safety requirements for land-based nuclear installations

    International Nuclear Information System (INIS)

    Kellermann, Otto

    1978-01-01

    The main ideas of safety philosophy for land-based nuclear installations are presented together with their background of protection goals. Today's requirements for design and quality assurance are deductively shown. Finally a proposition is made for a new balancing of safety philosophy according to the high safety level that nuclear installations have reached

  10. Investigation on regulatory requirements for radiation safety management

    International Nuclear Information System (INIS)

    Han, Eun Ok; Choi, Yoon Seok; Cho, Dae Hyung

    2013-01-01

    NRC recognizes that efficient management of radiation safety plan is an important factor to achieve radiation safety service. In case of Korea, the contents to perform the actual radiation safety management are legally contained in radiation safety management reports based on the Nuclear Safety Act. It is to prioritize the importance of safety regulations in each sector in accordance with the current situation of radiation and radioactive isotopes-used industry and to provide a basis for deriving safety requirements and safety regulations system maintenance by the priority of radiation safety management regulations. It would be helpful to achieve regulations to conform to reality based on international standards if consistent safety requirements is developed for domestic users, national standards and international standards on the basis of the results of questions answered by radiation safety managers, who lead on-site radiation safety management, about the priority of important factors in radioactive sources use, sales, production, moving user companies, to check whether derived configuration requirements for radiation safety management are suitable for domestic status

  11. IAEA safety requirements for safety assessment of fuel cycle facilities and activities

    International Nuclear Information System (INIS)

    Jones, G.

    2013-01-01

    The IAEA's Statute authorises the Agency to establish standards of safety for protection of health and minimisation of danger to life and property. In that respect, the IAEA has established a Safety Fundamentals publication which contains ten safety principles for ensuring the protection of workers, the public and the environment from the harmful effects of ionising radiation. A number of these principles require safety assessments to be carried out as a means of evaluating compliance with safety requirements for all nuclear facilities and activities and to determine the measures that need to be taken to ensure safety. The safety assessments are required to be carried out and documented by the organisation responsible for operating the facility or conducting the activity, are to be independently verified and are to be submitted to the regulatory body as part of the licensing or authorisation process. In addition to the principles of the Safety Fundamentals, the IAEA establishes requirements that must be met to ensure the protection of people and the environment and which are governed by the principles in the Safety Fundamentals. The IAEA's Safety Requirements publication 'Safety Assessment for Facilities and Activities', establishes the safety requirements that need to be fulfilled in conducting and maintaining safety assessments for the lifetime of facilities and activities, with specific attention to defence in depth and the requirement for a graded approach to the application of these safety requirements across the wide range of fuel cycle facilities and activities. Requirements for independent verification of the safety assessment that needs to be carried out by the operating organisation, including the requirement for the safety assessment to be periodically reviewed and updated are also covered. For many fuel cycle facilities and activities, environmental impact assessments and non-radiological risk assessments will be required. The

  12. High-Speed Maglev Trains; German Safety Requirements

    Science.gov (United States)

    1991-12-31

    This document is a translation of technology-specific safety requirements developed : for the German Transrapid Maglev technology. These requirements were developed by a : working group composed of representatives of German Federal Railways (DB), Tes...

  13. Requirements Engineering for Software Integrity and Safety

    Science.gov (United States)

    Leveson, Nancy G.

    2002-01-01

    Requirements flaws are the most common cause of errors and software-related accidents in operational software. Most aerospace firms list requirements as one of their most important outstanding software development problems and all of the recent, NASA spacecraft losses related to software (including the highly publicized Mars Program failures) can be traced to requirements flaws. In light of these facts, it is surprising that relatively little research is devoted to requirements in contrast with other software engineering topics. The research proposed built on our previous work. including both criteria for determining whether a requirements specification is acceptably complete and a new approach to structuring system specifications called Intent Specifications. This grant was to fund basic research on how these ideas could be extended to leverage innovative approaches to the problems of (1) reducing the impact of changing requirements, (2) finding requirements specification flaws early through formal and informal analysis, and (3) avoiding common flaws entirely through appropriate requirements specification language design.

  14. Innovative organotypic in vitro models for safety assessment: aligning with regulatory requirements and understanding models of the heart, skin, and liver as paradigms.

    NARCIS (Netherlands)

    Pridgeon, Chris S; Schlott, Constanze; Wong, Min Wei; Heringa, Minne B; Heckel, Tobias; Leedale, Joe; Launay, Laurence; Gryshkova, Vitalina; Przyborski, Stefan; Bearon, Rachel N; Wilkinson, Emma L; Ansari, Tahera; Greenman, John; Hendriks, Delilah F G; Gibbs, Sue; Sidaway, James; Sison-Young, Rowena L; Walker, Paul; Cross, Mike J; Park, B Kevin; Goldring, Chris E P

    2018-01-01

    The development of improved, innovative models for the detection of toxicity of drugs, chemicals, or chemicals in cosmetics is crucial to efficiently bring new products safely to market in a cost-effective and timely manner. In addition, improvement in models to detect toxicity may reduce the

  15. The Canadian Nuclear Safety Commission's financial guarantee requirements

    International Nuclear Information System (INIS)

    Ferch, R.

    2006-01-01

    The Nuclear Safety and Control Act gives the Canadian Nuclear Safety Commission (CNSC) the legal authority to require licensees to provide financial guarantees in order to meet the purposes of the Act. CNSC policy and guidance with regard to financial guarantees is outlined, and the current status of financial guarantee requirements as applied to various CNSC licensees is described. (author)

  16. Model quality and safety studies

    DEFF Research Database (Denmark)

    Petersen, K.E.

    1997-01-01

    The paper describes the EC initiative on model quality assessment and emphasizes some of the problems encountered in the selection of data from field tests used in the evaluation process. Further, it discusses the impact of model uncertainties in safety studies of industrial plants. The model...... that most of these have never been through a procedure of evaluation, but nonetheless are used to assist in making decisions that may directly affect the safety of the public and the environment. As a major funder of European research on major industrial hazards, DGXII is conscious of the importance......-tain model is appropriate for use in solving a given problem. Further, the findings from the REDIPHEM project related to dense gas dispersion will be highlighted. Finally, the paper will discuss the need for model quality assessment in safety studies....

  17. Predisposal Management of Radioactive Waste. General Safety Requirements Pt. 5

    International Nuclear Information System (INIS)

    2010-01-01

    There are a large number of facilities and activities around the world in which radioactive material is produced, handled and stored. This Safety Requirements publication presents international consensus requirements for the management of radioactive waste prior to its disposal. It provides the safety imperatives on the basis of which facilities can be designed, operated and regulated. The publication is supported by a number of Safety Guides that provide up to date recommendations and guidance on best practices for management of particular types of radioactive waste, for storage of radioactive waste, for assuring safety by developing safety cases and supporting safety assessments, and for applying appropriate management systems. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Responsibilities associated with the predisposal management of radioactive waste; 4. Steps in the predisposal management of radioactive waste; 5. Development and operation of predisposal radioactive waste management facilities and activities; Annex: Predisposal management of radioactive waste and the fundamental safety principles.

  18. Predisposal Management of Radioactive Waste. General Safety Requirements Pt. 5

    International Nuclear Information System (INIS)

    2009-01-01

    There are a large number of facilities and activities around the world in which radioactive material is produced, handled and stored. This Safety Requirements publication presents international consensus requirements for the management of radioactive waste prior to its disposal. It provides the safety imperatives on the basis of which facilities can be designed, operated and regulated. The publication is supported by a number of Safety Guides that provide up to date recommendations and guidance on best practices for management of particular types of radioactive waste, for storage of radioactive waste, for assuring safety by developing safety cases and supporting safety assessments, and for applying appropriate management systems. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Responsibilities associated with the predisposal management of radioactive waste; 4. Steps in the predisposal management of radioactive waste; 5. Development and operation of predisposal radioactive waste management facilities and activities; Annex: Predisposal management of radioactive waste and the fundamental safety principles.

  19. Analyzing Software Requirements Errors in Safety-Critical, Embedded Systems

    Science.gov (United States)

    Lutz, Robyn R.

    1993-01-01

    This paper analyzes the root causes of safety-related software errors in safety-critical, embedded systems. The results show that software errors identified as potentially hazardous to the system tend to be produced by different error mechanisms than non- safety-related software errors. Safety-related software errors are shown to arise most commonly from (1) discrepancies between the documented requirements specifications and the requirements needed for correct functioning of the system and (2) misunderstandings of the software's interface with the rest of the system. The paper uses these results to identify methods by which requirements errors can be prevented. The goal is to reduce safety-related software errors and to enhance the safety of complex, embedded systems.

  20. Operating safety requirements for the intermediate level liquid waste system

    International Nuclear Information System (INIS)

    1980-07-01

    The operation of the Intermediate Level Liquid Waste (ILW) System, which is described in the Final Safety Analysis, consists of two types of operations, namely: (1) the operation of a tank farm which involves the storage and transportation through pipelines of various radioactive liquids; and (2) concentration of the radioactive liquids by evaporation including rejection of the decontaminated condensate to the Waste Treatment Plant and retention of the concentrate. The following safety requirements in regard to these operations are presented: safety limits and limiting control settings; limiting conditions for operation; and surveillance requirements. Staffing requirements, reporting requirements, and steps to be taken in the event of an abnormal occurrence are also described

  1. Safety design guides for seismic requirements for CANDU 9

    International Nuclear Information System (INIS)

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

    1996-03-01

    This safety design guide for seismic requirements for CANDU 9 describes the seismic design philosophy, defines the applicable earthquakes and identifies the structures and systems requiring seismic qualification to ensure that the essential safety function can be adequately satisfied following earthquake. The detailed requirements for structures, systems and components which must be seismically qualified are specified in the Appendix. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 fig., (Author) .new

  2. Crewed Space Vehicle Battery Safety Requirements

    Science.gov (United States)

    Jeevarajan, Judith A.; Darcy, Eric C.

    2014-01-01

    This requirements document is applicable to all batteries on crewed spacecraft, including vehicle, payload, and crew equipment batteries. It defines the specific provisions required to design a battery that is safe for ground personnel and crew members to handle and/or operate during all applicable phases of crewed missions, safe for use in the enclosed environment of a crewed space vehicle, and safe for use in launch vehicles, as well as in unpressurized spaces adjacent to the habitable portion of a space vehicle. The required provisions encompass hazard controls, design evaluation, and verification. The extent of the hazard controls and verification required depends on the applicability and credibility of the hazard to the specific battery design and applicable missions under review. Evaluation of the design and verification program results shall be completed prior to certification for flight and ground operations. This requirements document is geared toward the designers of battery systems to be used in crewed vehicles, crew equipment, crew suits, or batteries to be used in crewed vehicle systems and payloads (or experiments). This requirements document also applies to ground handling and testing of flight batteries. Specific design and verification requirements for a battery are dependent upon the battery chemistry, capacity, complexity, charging, environment, and application. The variety of battery chemistries available, combined with the variety of battery-powered applications, results in each battery application having specific, unique requirements pertinent to the specific battery application. However, there are basic requirements for all battery designs and applications, which are listed in section 4. Section 5 includes a description of hazards and controls and also includes requirements.

  3. Operational and safety requirement of radiation facility

    International Nuclear Information System (INIS)

    Zulkafli Ghazali

    2007-01-01

    Gamma and electron irradiation facilities are the most common industrial sources of ionizing radiation. They have been used for medical, industrial and research purposes since the 1950s. Currently there are more than 160 gamma irradiation facilities and over 600 electron beam facilities in operation worldwide. These facilities are either used for the sterilization of medical and pharmaceutical products, the preservation of foodstuffs, polymer synthesis and modification, or the eradication of insect infestation. Irradiation with electron beam, gamma ray or ultra violet light can also destroy complex organic contaminants in both liquid and gaseous waste. EB systems are replacing traditional chemical sterilization methods in the medical supply industry. The ultra-violet curing facility, however, has found more industrial application in printing and furniture industries. Gamma and electron beam facilities produce very high dose rates during irradiation, and thus there is a potential of accidental exposure in the irradiation chamber which can be lethal within minutes. Although, the safety record of this industry has been relatively very good, there have been fatalities recorded in Italy (1975), Norway (1982), El Salvador (1989) and Israel (1990). Precautions against uncontrolled entry into irradiation chamber must therefore be taken. This is especially so in the case of gamma irradiation facilities those contain large amounts of radioactivity. If the mechanism for retracting the source is damaged, the source may remain exposed. This paper will, to certain extent, describe safety procedure and system being installed at ALURTRON, Nuclear Malaysia to eliminate accidental exposure of electron beam irradiation. (author)

  4. Traffic & safety statewide model and GIS modeling.

    Science.gov (United States)

    2012-07-01

    Several steps have been taken over the past two years to advance the Utah Department of Transportation (UDOT) safety initiative. Previous research projects began the development of a hierarchical Bayesian model to analyze crashes on Utah roadways. De...

  5. Radiation safety requirements for radionuclide laboratories

    International Nuclear Information System (INIS)

    2000-01-01

    The guide lays down the requirements for laboratories and storage rooms in which radioactive substances are used or stored as unsealed sources. In addition, some general instructions concerning work in radionuclide laboratories are set out

  6. Radiation safety requirements for radionuclide laboratories

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The guide lays down the requirements for laboratories and storage rooms in which radioactive substances are used or stored as unsealed sources. In addition, some general instructions concerning work in radionuclide laboratories are set out.

  7. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  8. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  9. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2016-01-01

    This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication

  10. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  11. Disposal of Radioactive Waste. Specific Safety Requirements (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Requirements publication applies to the disposal of radioactive waste of all types by means of emplacement in designed disposal facilities, subject to the necessary limitations and controls being placed on the disposal of the waste and on the development, operation and closure of facilities. The classification of radioactive waste is discussed. This Safety Requirements publication establishes requirements to provide assurance of the radiation safety of the disposal of radioactive waste, in the operation of a disposal facility and especially after its closure. The fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. This is achieved by setting requirements on the site selection and evaluation and design of a disposal facility, and on its construction, operation and closure, including organizational and regulatory requirements.

  12. OSHA safety requirements for hazardous chemicals in the workplace.

    Science.gov (United States)

    Dohms, J

    1992-01-01

    This article outlines the Occupational Safety and Health Administration (OSHA) requirements set forth by the Hazard Communication Standard, which has been in effect for the healthcare industry since 1987. Administrators who have not taken concrete steps to address employee health and safety issues relating to hazardous chemicals are encouraged to do so to avoid the potential of large fines for cited violations. While some states administer their own occupational safety and health programs, they must adopt standards and enforce requirements that are at least as effective as federal requirements.

  13. Design requirements of communication architecture of SMART safety system

    International Nuclear Information System (INIS)

    Park, H. Y.; Kim, D. H.; Sin, Y. C.; Lee, J. Y.

    2001-01-01

    To develop the communication network architecture of safety system of SMART, the evaluation elements for reliability and performance factors are extracted from commercial networks and classified the required-level by importance. A predictable determinacy, status and fixed based architecture, separation and isolation from other systems, high reliability, verification and validation are introduced as the essential requirements of safety system communication network. Based on the suggested requirements, optical cable, star topology, synchronous transmission, point-to-point physical link, connection-oriented logical link, MAC (medium access control) with fixed allocation are selected as the design elements. The proposed architecture will be applied as basic communication network architecture of SMART safety system

  14. Cold Vacuum Drying (CVD) Facility Technical Safety Requirements

    International Nuclear Information System (INIS)

    KRAHN, D.E.

    2000-01-01

    The Technical Safety Requirements (TSRs) for the Cold Vacuum Drying Facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt of multi-canister overpacks (MCOs) containing spent nuclear fuel. removal of free water from the MCOs using the cold vacuum drying process, and inerting and testing of the MCOs before transport to the Canister Storage Building. Controls required for public safety, significant defense in depth, significant worker safety, and for maintaining radiological and toxicological consequences below risk evaluation guidelines are included

  15. Nuclear safety requirements for operation licensing of Egyptian research reactors

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.

    2000-01-01

    From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

  16. [Safety culture: definition, models and design].

    Science.gov (United States)

    Pfaff, Holger; Hammer, Antje; Ernstmann, Nicole; Kowalski, Christoph; Ommen, Oliver

    2009-01-01

    Safety culture is a multi-dimensional phenomenon. Safety culture of a healthcare organization is high if it has a common stock in knowledge, values and symbols in regard to patients' safety. The article intends to define safety culture in the first step and, in the second step, demonstrate the effects of safety culture. We present the model of safety behaviour and show how safety culture can affect behaviour and produce safe behaviour. In the third step we will look at the causes of safety culture and present the safety-culture-model. The main hypothesis of this model is that the safety culture of a healthcare organization strongly depends on its communication culture and its social capital. Finally, we will investigate how the safety culture of a healthcare organization can be improved. Based on the safety culture model six measures to improve safety culture will be presented.

  17. A comparison of the difference of requirements between functional safety and nuclear safety controllers

    Energy Technology Data Exchange (ETDEWEB)

    Chen, C.K.; Lee, C.L.; Shyu, S.S. [Inst. of Nuclear Energy Research, Taoyuan, Taiwan (China)

    2014-07-01

    In order to establish self-reliant capabilities of nuclear I&C systems in Taiwan, Taiwan's Nuclear I&C System (TNICS) project had been established by Institute of Nuclear Energy Research (INER). A Triple Modular Redundant (TMR) safety controller (SCS-2000) has been completed and gone through the IEC 61508 Safety Integrity Level 3 (SIL3) certification of Functional Safety for industries. Based on the certification processes, the difference of requirements between Functional Safety and Nuclear Safety controllers in term of hardware and software are addressed in this study. Besides, the measures used to determine and verify the reliability of the safety control system design are presented. (author)

  18. Safety Design Approach for the Development of Safety Requirements for Design of Commercial HTGR

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Nishihara, Tetsuo; Yan, Xing; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-01-01

    The research committee on “Safety requirements for HTGR design” was established in 2013 under the Atomic Energy Society of Japan to develop the draft safety requirements for the design of commercial High Temperature Gas-cooled Reactors (HTGRs), which incorporate the HTGR safety features demonstrated using the High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the integration of the hydrogen production plants. The safety design approach for the commercial HTGRs which is a basement of the safety requirements is determined prior to the development of the safety requirements. The safety design approaches for the commercial HTGRs are to confine the radioactive materials within the coated fuel particles not only during normal operation but also during accident conditions, and the integrity of the coated fuel particles and other requiring physical barriers are protected by the inherent and passive safety features. This paper describes the main topics of the research committee, the safety design approaches and the safety functions of the commercial HTGRs determined in the research committee. (author)

  19. 41 CFR 128-1.8006 - Seismic Safety Program requirements.

    Science.gov (United States)

    2010-07-01

    ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Seismic Safety Program requirements. 128-1.8006 Section 128-1.8006 Public Contracts and Property Management Federal Property Management Regulations System (Continued) DEPARTMENT OF JUSTICE 1-INTRODUCTION 1.80-Seismic Safety Program...

  20. Safety assessment requirements for onsite transfers of radioactive material

    International Nuclear Information System (INIS)

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

    1992-05-01

    This document contains the requirements for developing a safety assessment document for an onsite package containing radioactive material. It also provides format and content guidance to establish uniformity in the safety assessment documentation and to ensure completeness of the information provided

  1. Macroscopic models for traffic safety.

    NARCIS (Netherlands)

    Oppe, S.

    1988-01-01

    Recently there has been an increased interest in the application of macroscopic models for the description of developments in traffic safety. A discussion was started on the causes of the sudden decrease in the number of fatal and injury accidents after 1974. Before that time these numbers had

  2. New requirements on safety of nuclear power plants according to the IAEA safety standards

    International Nuclear Information System (INIS)

    Misak, J.

    2005-01-01

    In this presentation author presents new requirements on safety of nuclear power plants according to the IAEA safety standards. It is concluded that: - New set of IAEA Safety Standards is close to completion: around 40 standards for NPPs; - Different interpretation of IAEA Safety Standards at present: best world practices instead of previous 'minimum common denominator'; - A number of safety improvements required for NPPs; - Requirements related to BDBAs and severe accidents are the most demanding due to degradation of barriers: hardware modifications and accident management; - Large variety between countries in implementation of accident management programmes: from minimum to major hardware modifications; -Distinction between existing and new NPPs is essential from the point of view of the requirements; WWER 440 reactors have potential to reflect IAEA Safety Standards for existing NPPs; relatively low reactor power offers broader possibilities

  3. Fuel Supply Shutdown Facility Interim Operational Safety Requirements

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2000-01-01

    The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

  4. Requirements of radiation protection and safety for nuclear medicine services

    International Nuclear Information System (INIS)

    1989-01-01

    The requirements of radiation protection and safety for nuclear medicine services are established. The norms is applied to activities related to the radiopharmaceuticals for therapeutics and 'in vivo' diagnostics purposes. (M.C.K.) [pt

  5. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEA’s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

  6. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEA’s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

  7. Requirements to be met by a safety philosophy

    International Nuclear Information System (INIS)

    Hahn, L.

    1990-01-01

    The author's assessment of the use of safety philosophies is that, since 'safety philosophers' still are not certain whether a safety philosophy ought to be applicable to just one, particular technology, or rather to a variety of different technologies, there is reason to state that the required ethical, philosophical and political foundations to build a safety philosophy on are still missing. And this, the author presumes, is one of the reasons why our society to a far extent is incapable of acting, faced not only with the nuclear issue, but also with the present and future ecological challenge. (orig./DG) [de

  8. Quality assurance requirements for the computer software and safety analyses

    International Nuclear Information System (INIS)

    Husarecek, J.

    1992-01-01

    The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref

  9. Simulation modeling and analysis in safety. II

    International Nuclear Information System (INIS)

    Ayoub, M.A.

    1981-01-01

    The paper introduces and illustrates simulation modeling as a viable approach for dealing with complex issues and decisions in safety and health. The author details two studies: evaluation of employee exposure to airborne radioactive materials and effectiveness of the safety organization. The first study seeks to define a policy to manage a facility used in testing employees for radiation contamination. An acceptable policy is one that would permit the testing of all employees as defined under regulatory requirements, while not exceeding available resources. The second study evaluates the relationship between safety performance and the characteristics of the organization, its management, its policy, and communication patterns among various functions and levels. Both studies use models where decisions are reached based on the prevailing conditions and occurrence of key events within the simulation environment. Finally, several problem areas suitable for simulation studies are highlighted. (Auth.)

  10. HTR-PM Safety requirement and Licensing experience

    International Nuclear Information System (INIS)

    Li Fu; Zhang Zuoyi; Dong Yujie; Wu Zongxin; Sun Yuliang

    2014-01-01

    HTR-PM is a 200MWe modular pebble bed high temperature reactor demonstration plant which is being built in Shidao Bay, Weihai, Shandong, China. The main design parameters of HTR-PM were fixed in 2006, the basic design was completed in 2008. The review of Preliminary Safety Analysis Report (PSAR) of HTR-PM was started in April 2008, completed in September 2009. In general, HTR- PM design complies with the current safety requirement for nuclear power plant in China, no special standards are developed for modular HTR. Anyway, Chinese Nuclear Safety Authority, together with the designers, developed some dedicated design criteria for key systems and components and published the guideline for the review of safety analysis report of HTR-PM, based on the experiences from licensing of HTR-10 and new development of nuclear safety. The probabilistic safety goal for HTR-PM was also defined by the safety authority. The review of HTR-PM PSAR lasted for one and a half years, with 3 dialogues meetings and 8 topics meetings, with more than 2000 worksheets and answer sheets. The heavily discussed topics during the PSAR review process included: the requirement for the sub-atmospheric ventilation system, the utilization of PSA in design process, the scope of beyond design basis accidents, the requirement for the qualification of TRISO coating particle fuel, and etc. Because of the characteristics of first of a kind for the demonstration plant, the safety authority emphasized the requirement for the experiment and validation, the PSAR was licensed with certain licensing conditions. The whole licensing process was under control, and was re-evaluated again after Fukushima accident to be shown that the design of HTR-PM complies with current safety requirement. This is a good example for how to license a new reactor. (author)

  11. Hazard Analysis and Safety Requirements for Small Drone Operations: To What Extent Do Popular Drones Embed Safety?

    Science.gov (United States)

    Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimihailidou, Maria Mikela

    2018-03-01

    Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this article presents: (1) a set of safety requirements generated from the application of the systems theoretic process analysis (STPA) technique on a generic small drone system; (2) a gap analysis between the set of safety requirements and the ones met by 19 popular drone models; (3) the extent of the differences between those models, their manufacturers, and the countries of origin; and (4) the association of drone prices with the extent they meet the requirements derived by STPA. The application of STPA resulted in 70 safety requirements distributed across the authority, manufacturer, end user, or drone automation levels. A gap analysis showed high dissimilarities regarding the extent to which the 19 drones meet the same safety requirements. Statistical results suggested a positive correlation between drone prices and the extent that the 19 drones studied herein met the safety requirements generated by STPA, and significant differences were identified among the manufacturers. This work complements the existing risk assessment frameworks for small drones, and contributes to the establishment of a commonly endorsed international risk analysis framework. Such a framework will support the development of a holistic and methodologically justified standardization scheme for small drone flights. © 2017 Society for Risk Analysis.

  12. Requirements to amend the main influence factors on the safety culture after fukushima accident

    International Nuclear Information System (INIS)

    Farcasiu, M.; Nitoi, M.

    2015-01-01

    The paper presents a general model that provides a framework for the safety culture assessment, creating the possibility to identify factors that can significantly influence the safety culture. The main safety culture influence factors (SCIF) used by model are the following: regulatory environment, organizational environment, worker characteristics, socio-political environment, national culture, organization history, business and technological characteristics. After the analysis of the deficiencies and weaknesses of SCIFc in evolution of the Fukushima accident, some issues that may become necessities and requirements to change and improve both the safety culture and safety of the nuclear installations were highlighted. For each influence factor were identified some requirements to amend. The results will emphasize the necesity of the human - technology - organization system assessment. Hence it was demonstrated that the safety culture results from the interaction of individuals with technology and with the organization. (authors)

  13. A review of the predictive modelling and data requirements for the long-term safety assessment of the deep disposal of low and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Broyd, T.W.

    1988-06-01

    This report considers the Her Majesty's Inspectorate of Pollution research and modelling requirements for a robust post-closure radiological risk assessment methodology applicable to the deep disposal of Low-Level Wastes and Intermediate-Level Wastes. Two disposal concepts have been envisaged: horizontal tunnels or galleries in a low permeability stratum of a sedimentary sequence located inland; vertical boreholes or shafts up to 15m diameter lined with concrete and of the order 500m to 1000m deep sunk into the seabed within territorial coastal waters of the United Kingdom. (author)

  14. International standardization of safety requirements for fast reactors

    International Nuclear Information System (INIS)

    2011-06-01

    Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

  15. Radiation safety requirements for training of users of diagnostic X ...

    African Journals Online (AJOL)

    Background. Globally, the aim of requirements regarding the use and ownership of diagnostic medical X-ray equipment is to limit radiation by abiding by the 'as low as reasonably achievable' (ALARA) principle. The ignorance of radiographers with regard to radiation safety requirements, however, is currently a cause of ...

  16. Safety Requirements and Modern Technical Requirements in Human Information Systems in Amman Hotels

    OpenAIRE

    Farouq Ahmad Alazzam; Sattam Rakan Allahawiah; Mohammad Nayef Alsarayreh; Kafa Hmoud Abdallah al Nawaiseh

    2015-01-01

    This study aimed to demonstrate the availability of Safety requirements and modern technical requirements in human information systems in Amman hotels. an the most important results of this study is the availability of security and safety requirements in human information systems In Amman hotels and The adequacy of the information that it provided .and show that all departments are not connected by appropriate and effective communication networks in adequate form . Also sophisticated operatin...

  17. Predisposal management of radioactive waste. General safety requirements. Pt. 5

    International Nuclear Information System (INIS)

    2009-01-01

    The objective of this Safety Requirements publication is to establish, the requirements that must be satisfied in the predisposal management of radioactive waste. This publication sets out the objectives, criteria and requirements for the protection of human health and the environment that apply to the siting, design, construction, commissioning, operation and shutdown of facilities for the predisposal management of radioactive waste, and the requirements that must be met to ensure the safety of such facilities and activities. This Safety Requirements publication applies to the predisposal management of radioactive waste of all types and covers all the steps in its management from its generation up to its disposal, including its processing (pretreatment, treatment and conditioning), storage and transport. Such waste may arise from the commissioning, operation and decommissioning of nuclear facilities; the use of radionuclides in medicine, industry, agriculture, research and education; the processing of materials that contain naturally occurring radionuclides; and the remediation of contaminated areas. The introduction of the document (Section 1) informs about its objective, scope and structure. The protection of human health and the environment is considered in Section 2 of this publication. Section 3 establishes requirements for the responsibilities associated with the predisposal management of radioactive waste. Requirements for the principal approaches to and the elements of the predisposal management of radioactive waste are established in Section 4. Section 5 establishes requirements for the safe development and operation of predisposal radioactive waste management facilities and safe conduct of activities. The Annex presents a discussion of the consistency of the safety requirements established in this publication with the fundamental safety principles

  18. An integrative model of organizational safety behavior.

    Science.gov (United States)

    Cui, Lin; Fan, Di; Fu, Gui; Zhu, Cherrie Jiuhua

    2013-06-01

    This study develops an integrative model of safety management based on social cognitive theory and the total safety culture triadic framework. The purpose of the model is to reveal the causal linkages between a hazardous environment, safety climate, and individual safety behaviors. Based on primary survey data from 209 front-line workers in one of the largest state-owned coal mining corporations in China, the model is tested using structural equation modeling techniques. An employee's perception of a hazardous environment is found to have a statistically significant impact on employee safety behaviors through a psychological process mediated by the perception of management commitment to safety and individual beliefs about safety. The integrative model developed here leads to a comprehensive solution that takes into consideration the environmental, organizational and employees' psychological and behavioral aspects of safety management. Copyright © 2013 National Safety Council and Elsevier Ltd. All rights reserved.

  19. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  20. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  1. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  2. The Management System for Facilities and Activities. Safety Requirements

    International Nuclear Information System (INIS)

    2011-01-01

    This publication establishes requirements for management systems that integrate safety, health, security, quality assurance and environmental objectives. A successful management system ensures that nuclear safety matters are not dealt with in isolation but are considered within the context of all these objectives. The aim of this publication is to assist Member States in establishing and implementing effective management systems that integrate all aspects of managing nuclear facilities and activities in a coherent manner. It details the planned and systematic actions necessary to provide adequate confidence that all these requirements are satisfied. Contents: 1. Introduction; 2. Management system; 3. Management responsibility; 4. Resource management; 5. Process implementation; 6. Measurement, assessment and improvement.

  3. Recommended general safety requirements for nuclear power plants

    International Nuclear Information System (INIS)

    1983-06-01

    This report presents recommendations for a set of general safety requirements that could form the basis for the licensing of nuclear power plants by the Atomic Energy Control Board. In addition to a number of recommended deterministic requirements the report includes criteria for the acceptability of the design of such plants based upon the calculated probability and consequence (in terms of predicted radiation dose to members of the public) of potential fault sequences. The report also contains a historical review of nuclear safety principles and practices in Canada

  4. Modeling requirements for in situ vitrification

    International Nuclear Information System (INIS)

    MacKinnon, R.J.; Mecham, D.C.; Hagrman, D.L.; Johnson, R.W.; Murray, P.E.; Slater, C.E.; Marwil, E.S.; Weaver, R.A.; Argyle, M.D.

    1991-11-01

    This document outlines the requirements for the model being developed at the INEL which will provide analytical support for the ISV technology assessment program. The model includes representations of the electric potential field, thermal transport with melting, gas and particulate release, vapor migration, off-gas combustion and process chemistry. The modeling objectives are to (1) help determine the safety of the process by assessing the air and surrounding soil radionuclide and chemical pollution hazards, the nuclear criticality hazard, and the explosion and fire hazards, (2) help determine the suitability of the ISV process for stabilizing the buried wastes involved, and (3) help design laboratory and field tests and interpret results therefrom

  5. Scale modelling in LMFBR safety

    International Nuclear Information System (INIS)

    Cagliostro, D.J.; Florence, A.L.; Abrahamson, G.R.

    1979-01-01

    This paper reviews scale modelling techniques used in studying the structural response of LMFBR vessels to HCDA loads. The geometric, material, and dynamic similarity parameters are presented and identified using the methods of dimensional analysis. Complete similarity of the structural response requires that each similarity parameter be the same in the model as in the prototype. The paper then focuses on the methods, limitations, and problems of duplicating these parameters in scale models and mentions an experimental technique for verifying the scaling. Geometric similarity requires that all linear dimensions of the prototype be reduced in proportion to the ratio of a characteristic dimension of the model to that of the prototype. The overall size of the model depends on the structural detail required, the size of instrumentation, and the costs of machining and assemblying the model. Material similarity requires that the ratio of the density, bulk modulus, and constitutive relations for the structure and fluid be the same in the model as in the prototype. A practical choice of a material for the model is one with the same density and stress-strain relationship as the operating temperature. Ni-200 and water are good simulant materials for the 304 SS vessel and the liquid sodium coolant, respectively. Scaling of the strain rate sensitivity and fracture toughness of materials is very difficult, but may not be required if these effects do not influence the structural response of the reactor components. Dynamic similarity requires that the characteristic pressure of a simulant source equal that of the prototype HCDA for geometrically similar volume changes. The energy source is calibrated in the geometry and environment in which it will be used to assure that heat transfer between high temperature loading sources and the coolant simulant and that non-equilibrium effects in two-phase sources are accounted for. For the geometry and flow conitions of interest, the

  6. Safety integrity requirements for computer based I ampersand C systems

    International Nuclear Information System (INIS)

    Thuy, N.N.Q.; Ficheux-Vapne, F.

    1997-01-01

    In order to take into account increasingly demanding functional requirements, many instrumentation and control (I ampersand C) systems in nuclear power plants are implemented with computers. In order to ensure the required safety integrity of such equipment, i.e., to ensure that they satisfactorily perform the required safety functions under all stated conditions and within stated periods of time, requirements applicable to these equipment and to their life cycle need to be expressed and followed. On the other hand, the experience of the last years has led EDF (Electricite de France) and its partners to consider three classes of systems and equipment, according to their importance to safety. In the EPR project (European Pressurized water Reactor), these classes are labeled E1A, E1B and E2. The objective of this paper is to present the outline of the work currently done in the framework of the ETC-I (EPR Technical Code for I ampersand C) regarding safety integrity requirements applicable to each of the three classes. 4 refs., 2 figs

  7. Requirements Modeling with Agent Programming

    Science.gov (United States)

    Dasgupta, Aniruddha; Krishna, Aneesh; Ghose, Aditya K.

    Agent-oriented conceptual modeling notations are highly effective in representing requirements from an intentional stance and answering questions such as what goals exist, how key actors depend on each other, and what alternatives must be considered. In this chapter, we review an approach to executing i* models by translating these into set of interacting agents implemented in the CASO language and suggest how we can perform reasoning with requirements modeled (both functional and non-functional) using i* models. In this chapter we particularly incorporate deliberation into the agent design. This allows us to benefit from the complementary representational capabilities of the two frameworks.

  8. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1, Revision 1 (Chinese Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication establishes requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  9. Fuel supply shutdown facility interim operational safety requirements

    International Nuclear Information System (INIS)

    Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

    1995-01-01

    These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

  10. Integrated Safety Culture Model and Application

    Institute of Scientific and Technical Information of China (English)

    汪磊; 孙瑞山; 刘汉辉

    2009-01-01

    A new safety culture model is constructed and is applied to analyze the correlations between safety culture and SMS. On the basis of previous typical definitions, models and theories of safety culture, an in-depth analysis on safety culture's structure, composing elements and their correlations was conducted. A new definition of safety culture was proposed from the perspective of sub-cuhure. 7 types of safety sub-culture, which are safety priority culture, standardizing culture, flexible culture, learning culture, teamwork culture, reporting culture and justice culture were defined later. Then integrated safety culture model (ISCM) was put forward based on the definition. The model divided safety culture into intrinsic latency level and extrinsic indication level and explained the potential relationship between safety sub-culture and all safety culture dimensions. Finally in the analyzing of safety culture and SMS, it concluded that positive safety culture is the basis of im-plementing SMS effectively and an advanced SMS will improve safety culture from all around.

  11. Safety related requirements on future nuclear power plants

    International Nuclear Information System (INIS)

    Niehaus, F.

    1991-01-01

    Nuclear power has the potential to significantly contribute to the future energy supply. However, this requires continuous improvements in nuclear safety. Technological advancements and implementation of safety culture will achieve a safety level for future reactors of the present generation of a probability of core-melt of less than 10 -5 per year, and less than 10 -6 per year for large releases of radioactive materials. There are older reactors which do not comply with present safety thinking. The paper reviews findings of a recent design review of WWER 440/230 plants. Advanced evolutionary designs might be capable of reducing the probability of significant off-site releases to less than 10 -7 per year. For such reactors there are inherent limitations to increase safety further due to the human element, complexity of design and capability of the containment function. Therefore, revolutionary designs are being explored with the aim of eliminating the potential for off-site releases. In this context it seems to be advisable to explore concepts where the ultimate safety barrier is the fuel itself. (orig.) [de

  12. UK experience of safety requirements for thermal reactor stations

    International Nuclear Information System (INIS)

    Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

    1977-01-01

    The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

  13. Safety and regulatory requirements of nuclear power plants

    International Nuclear Information System (INIS)

    Kumar, S.V.; Bhardwaj, S.A.

    2000-01-01

    A pre-requisite for a nuclear power program in any country is well established national safety and regulatory requirements. These have evolved for nuclear power plants in India with participation of the regulatory body, utility, research and development (R and D) organizations and educational institutions. Prevailing international practices provided a useful base to develop those applicable to specific system designs for nuclear power plants in India. Their effectiveness has been demonstrated in planned activities of building up the nuclear power program as well as with unplanned activities, like those due to safety related incidents etc. (author)

  14. Technical Safety Requirements for the Gamma Irradiation Facility (GIF)

    CERN Document Server

    Mahn, J A E M J G

    2003-01-01

    This document provides the Technical Safety Requirements (TSR) for the Sandia National Laboratories Gamma Irradiation Facility (GIF). The TSR is a compilation of requirements that define the conditions, the safe boundaries, and the administrative controls necessary to ensure the safe operation of a nuclear facility and to reduce the potential risk to the public and facility workers from uncontrolled releases of radioactive or other hazardous materials. These requirements constitute an agreement between DOE and Sandia National Laboratories management regarding the safe operation of the Gamma Irradiation Facility.

  15. Westinghouse Hanford Company safety analysis reports and technical safety requirements upgrade program

    International Nuclear Information System (INIS)

    Busche, D.M.

    1995-09-01

    During Fiscal Year 1992, the US Department of Energy, Richland Operations Office (RL) separately transmitted the following US Department of Energy (DOE) Orders to Westinghouse Hanford Company (WHC) for compliance: DOE 5480.21, ''Unreviewed Safety Questions,'' DOE 5480.22, ''Technical Safety Requirements,'' and DOE 5480.23, ''Nuclear Safety Analysis Reports.'' WHC has proceeded with its impact assessment and implementation process for the Orders. The Orders are closely-related and contain some requirements that are either identical, similar, or logically-related. Consequently, WHC has developed a strategy calling for an integrated implementation of the three Orders. The strategy is comprised of three primary objectives, namely: Obtain DOE approval of a single list of DOE-owned and WHC-managed Nuclear Facilities, Establish and/or upgrade the ''Safety Basis'' for each Nuclear Facility, and Establish a functional Unreviewed Safety Question (USQ) process to govern the management and preservation of the Safety Basis for each Nuclear Facility. WHC has developed policy-revision and facility-specific implementation plans to accomplish near-term tasks associated with the above strategic objectives. This plan, which as originally submitted in August 1993 and approved, provided an interpretation of the new DOE Nuclear Facility definition and an initial list of WHC-managed Nuclear Facilities. For each current existing Nuclear Facility, existing Safety Basis documents are identified and the plan/status is provided for the ISB. Plans for upgrading SARs and developing TSRs will be provided after issuance of the corresponding Rules

  16. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-09-15

    The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

  17. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Arabic Edition)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-09-15

    The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

  18. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Spanish Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered

  19. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (French Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered

  20. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Chinese Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered

  1. Development of photovoltaic array and module safety requirements

    Science.gov (United States)

    1982-01-01

    Safety requirements for photovoltaic module and panel designs and configurations likely to be used in residential, intermediate, and large-scale applications were identified and developed. The National Electrical Code and Building Codes were reviewed with respect to present provisions which may be considered to affect the design of photovoltaic modules. Limited testing, primarily in the roof fire resistance field was conducted. Additional studies and further investigations led to the development of a proposed standard for safety for flat-plate photovoltaic modules and panels. Additional work covered the initial investigation of conceptual approaches and temporary deployment, for concept verification purposes, of a differential dc ground-fault detection circuit suitable as a part of a photovoltaic array safety system.

  2. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  3. Regulatory requirements and administrative practice in safety of nuclear installations

    International Nuclear Information System (INIS)

    Servant, J.

    1977-01-01

    This paper reviews the current situation of the France regulatory rules and procedures dealing with the safety of the main nuclear facilities and, more broadly, the nuclear security. First, the author outlines the policy of the French administration which requires that the licensee responsible for an installation has to demonstrate that all possible measures are taken to ensure a sufficient level of safety, from the early stage of the project to the end of the operation of the plant. Thus, the administration performs the assessment on a case-by-case basis, of the safety of each installation before granting a nuclear license. On the other hand, the administration settles overall safety requirements for specific categories of installations or components, which determine the ultimate safety performances, but avoid, as far as possible, to detail the technical specifications to be applied in order to comply with these goals. This approach, which allows the designers and the licensees to rely upon sound codes and standards, gains the advantage of a great flexibility without imparing the nuclear safety. The author outlines the licensing progress for the main categories of installations: nuclear power plants of the PWR type, fast breeders, uranium isotope separation plants, and irradiated fuel processing plants. Emphasis is placed on the most noteworthy points: standardization of projects, specific risks of each site, problems of advanced type reactors, etc... The development of the technical regulations is presented with emphasis on the importance of an internationally concerned action within the nuclear international community. The second part of this paper describes the France operating experience of nuclear installations from the safety point of view. Especially, the author examines the technical and administrative utilization of data from safety significant incidents in reactors and plants, and the results of the control performed by the nuclear installations

  4. Directions for model building from asymptotic safety

    Science.gov (United States)

    Bond, Andrew D.; Hiller, Gudrun; Kowalska, Kamila; Litim, Daniel F.

    2017-08-01

    Building on recent advances in the understanding of gauge-Yukawa theories we explore possibilities to UV-complete the Standard Model in an asymptotically safe manner. Minimal extensions are based on a large flavor sector of additional fermions coupled to a scalar singlet matrix field. We find that asymptotic safety requires fermions in higher representations of SU(3) C × SU(2) L . Possible signatures at colliders are worked out and include R-hadron searches, diboson signatures and the evolution of the strong and weak coupling constants.

  5. Recommended safety objectives, principles and requirements for mini-reactors

    International Nuclear Information System (INIS)

    1991-05-01

    Canadian and international publications containing objectives, principles and requirements for the safety of nuclear facilities in general and nuclear power plants in particular have been reviewed for their relevance to mini-reactors. Most of the individual recommendations, sometimes with minor wording changes, are applicable to mini-reactors. However, some prescriptive requirements for the shutdown, emergency core cooling and containment systems of power reactors are considered inappropriate for mini-reactors. The Advisory Committee on Nuclear Safety favours a generally non-prescriptive approach whereby the applicant for a mini-reactor license is free to propose any means of satisfying the fundamental objectives, but must convince the regulatory agency to that effect. To do so, a probabilistic safety assessment (PSA) would be the favoured procedure. A generic PSA for all mini-reactors of the same design would be acceptable. Notwithstanding this non-prescriptive approach, the ACNS considers that it would be prudent to require the existence of at least one independent shutdown system and two physically independent locations from which the reactor can be shut down and the shutdown condition monitored, and to require provision for an assumed loss of integrity of the primary cooling system's boundary unless convincing arguments to the contrary are presented. The ACNS endorses in general the objectives and fundamental principles proposed by the interorganizational Small Reactor Criteria working group, and intends to review and comment on the documents on specific applications to be issued by that working group

  6. DARHT: INTEGRATION OF AUTHORIZATION BASIS REQUIREMENTS AND WORKER SAFETY

    International Nuclear Information System (INIS)

    MC CLURE, D. A.; NELSON, C. A.; BOUDRIE, R. L.

    2001-01-01

    This document describes the results of consensus agreements reached by the DARHT Safety Planning Team during the development of the update of the DARHT Safety Analysis Document (SAD). The SAD is one of the Authorization Basis (AB) Documents required by the Department prior to granting approval to operate the DARHT Facility. The DARHT Safety Planning Team is lead by Mr. Joel A. Baca of the Department of Energy Albuquerque Operations Office (DOE/AL). Team membership is drawn from the Department of Energy Albuquerque Operations Office, the Department of Energy Los Alamos Area Office (DOE/LAAO), and several divisions of the Los Alamos National Laboratory. Revision 1 of the DARHT SAD had been written as part of the process for gaining approval to operate the Phase 1 (First Axis) Accelerator. Early in the planning stage for the required update of the SAD for the approval to operate both Phase 1 and Phase 2 (First Axis and Second Axis) DARHT Accelerator, it was discovered that a conflict existed between the Laboratory approach to describing the management of facility and worker safety

  7. Long term safety requirements and safety indicators for the assessment of underground radioactive waste repositories

    International Nuclear Information System (INIS)

    Vovk, Ivan

    1998-01-01

    This presentation defines: waste disposal, safety issues, risk estimation; describes the integrated waste disposal process including quality assurance program. Related to actinides inventory it shows the main results of calculated activity obtained by deterministic estimation. It includes the Radioactive Waste Safety Standards and requirements; features related to site, design and waste package characteristics, as technical long term safety criteria for radioactive waste disposal facilities. Fundamental concern regarding the safety of radioactive waste disposal systems is their radiological impact on human beings and the environment. Safety requirements and criteria for judging the level of safety of such systems have been developed and there is a consensus among the international community on their basis within the well-established system of radiological protection. So far, however, little experience has been gained in applying long term safety criteria to actual disposal systems; consequently, there is an international debate on the most appropriate nature and form of the criteria to be used, taking into account the uncertainties involved. Emerging from the debate is the increasing conviction that the combined use of a variety of indicators would be advantageous in addressing the issue of reasonable assurance in the different time frames involved and in supporting the safety case for any particular repository concept. Indicators including risk, dose, radionuclide concentration, transit time, toxicity indices, fluxes at different points within the system, and barrier performance have all been identified as potentially relevant. Dose and risk are the indicators generally seen as most fundamental, as they seek directly to describe the radiological impact of a disposal system, and these are the ones that have been incorporated into most national standards to date. There are, however, certain problems in applying them. Application of a variety of different indicators

  8. GENERAL CONSIDERATIONS ON REGULATIONS AND SAFETY REQUIREMENTS FOR QUADRICYCLES

    Directory of Open Access Journals (Sweden)

    Ana Pavlovic

    2015-12-01

    Full Text Available In recent years, a new class of compact vehicles has been emerging and wide-spreading all around Europe: the quadricycle. These four-wheeled motor vehicles, originally derived from motorcycles, are a small and fuel-efficient mean of transportation used in rural or urban areas as an alternative to motorbikes or city cars. In some countries, they are also endorsed by local authorities and institutions which support small and environmentally-friendly vehicles. In this paper, several general considerations on quadricycles will be provided including the vehicle classification, evolution of regulations (as homologation, driver licence, emissions, etc, technical characteristics, safety requirements, most relevant investigations, and other additional useful information (e.g. references, links. It represents an important and actual topic of investigation for designers and manufacturers considering that the new EU regulation on the approval and market surveillance of quadricycles will soon enter in force providing conclusive requirements for functional safety environmental protection of these promising vehicles.

  9. Status of safety issues at licensed power plants: TMI action plan requirements, unresolved safety issues, generic safety issues

    International Nuclear Information System (INIS)

    1991-12-01

    As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program was established whereby an annual NUREG report would be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was compiled and reported in three NUREG volumes. Volume 1, published in March 1991, addressed the status of of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). This annual NUREG report combines these volumes into a single report and provides updated information as of September 30, 1991. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. This report is to provide a comprehensive description of the implementation and verification status of TMI Action Plan Requirements, safety issues designated as USIs, and GSIs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. An additional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

  10. Responsibility for the Violation of Ecological Safety Requirements

    Science.gov (United States)

    Selivanovskaya, J. I.; Gilmutdinova, I.

    2018-01-01

    The article deals with the problems of responsibility for the violation of ecological safety requirements from the point of view of sustainable development of the state. Such types of responsibility as property, disciplinary, financial, administrative and criminal responsibility in the area are analysed. Suggestions on the improvement of legislation are put forward. Among other things it is suggested to introduce criminal sanctions against legal bodies (enterprises) for ecological crimes with punishments in the form of fines, suspension or discontinuation of activities.

  11. Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

    International Nuclear Information System (INIS)

    2009-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

  12. Safety assessment for facilities and activities. General safety requirements. Pt. 4

    International Nuclear Information System (INIS)

    2009-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF 6 ; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

  13. Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

    International Nuclear Information System (INIS)

    2010-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

  14. Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

    International Nuclear Information System (INIS)

    2009-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation.? read more The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are

  15. Safety requirements and safety experience of nuclear facilities in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Schnurer, H.L.

    1977-01-01

    Peaceful use of nuclear energy within the F.R.G. is rapidly growing. The Energy Programme of the Federal Government forecasts a capacity of up to 50.000 MW in 1985. Whereas most of this capacity will be of the LWR-Type, other activities are related to LMFBR - and HTGR - development, nuclear ships, and facilities of the nuclear fuel cycle. Safety of nuclear energy is the pacemaker for the realization of nuclear programmes and projects. Due to a very high population - and industrialisation density, safety has the priority before economical aspects. Safety requirements are therefore extremely stringent, which will be shown for the legal, the technical as well as for the organizational area. They apply for each nuclear facility, its site and the nuclear energy system as a whole. Regulatory procedures differ from many other countries, assigning executive power to state authorities, which are supervised by the Federal Government. Another particularity of the regulatory process is the large scope of involvement of independent experts within the licensing procedures. The developement of national safety requirements in different countries generates a necessity to collaborate and harmonize safety and radiation protection measures, at least for facilities in border areas, to adopt international standards and to assist nuclear developing countries. However, different nationally, regional or local situations might raise problems. Safety experience with nuclear facilities can be concluded from the positive construction and operation experience, including also a few accidents and incidents and the conclusions, which have been drawn for the respective factilities and others of similar design. Another tool for safety assessments will be risk analyses, which are under development by German experts. Final, a scope of future problems and developments shows, that safety of nuclear installations - which has reached a high performance - nevertheless imposes further tasks to be solved

  16. 47 CFR 80.305 - Watch requirements of the Communications Act and the Safety Convention.

    Science.gov (United States)

    2010-10-01

    ... and the Safety Convention. 80.305 Section 80.305 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) SAFETY AND SPECIAL RADIO SERVICES STATIONS IN THE MARITIME SERVICES Safety Watch Requirements and Procedures Ship Station Safety Watches § 80.305 Watch requirements of the Communications Act and the Safety...

  17. Hazard analysis & safety requirements for small drone operations : to what extent do popular drones embed safety?

    NARCIS (Netherlands)

    Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimichailidou, Maria Mikela

    2018-01-01

    Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this paper presents: (1) a set of safety

  18. EPR meets the next generation PWR safety requirements

    International Nuclear Information System (INIS)

    Bouteille, Francois; Czech, Juergen; Sloan, Sandra

    2006-01-01

    At the origin was the common decision in 1989 of Framatome and Siemens to cooperate to design a Nuclear Island which meets the future needs of utilities. EDF and a group of main German Utilities joined this effort in 1991 and from that point were completely involved in the progress of the work. Compliance of the EPR with the European Utility Requirements (EUR) was verified to ensure a large acceptability of the design by other participating utilities. In addition, the entire process was backed up to the end of 1998 by the French and the German Safety Authorities which engaged into a long-lasting cooperation to define common requirements applicable to future Nuclear Power Plants. Upon signature of the Olkiluoto 3 contract, STUK, the Finnish safety and radiation authority, began reviewing the design of the EPR. Upon the favorable recommendation of STUK, the Finnish government delivered a Construction License for the Olkiluoto 3 NPP on February 17, 2005. Following the positive conclusion of the political debate in France with regard to nuclear energy, EDF will also submit a request to start the construction of an EPR on the Flamanville site. In the US, the first steps in view of a Design Certification by the NRC have been taken. These three independent decisions make the EPR the leading first generation 3+ design under construction. Important safety functions are assured by separate systems in a straightforward operating mode. Four separate, redundant trains for all safety systems are installed in four separate layout division for which a strict separation is ensured so that common mode failure, for example due to internal hazards, can be ruled out. A reduction in common mode failure potential is also obtained by design rules ensuring the systematic application of functional diversity. A four train-redundancy for the major safety systems provides flexibility in adapting the design to maintenance requirements, thus contributing to reduce the outage duration. Additional

  19. Safety requirements for long term operation of NPPs

    International Nuclear Information System (INIS)

    Houdre, T.; Osouf, N.; Juvin, J.-C.

    2012-01-01

    In the future, the reactors operating at present will run alongside reactors of the EPR type or their equivalent, designed for a significantly higher level of safety. This raises the question of the acceptability of continued operation of reactors beyond 40 years when there is an available technology that is safer. Two objectives are therefore imperative. First, a re-evaluation of the safety level in the light of that required of EPR type reactors or their equivalent is necessary, with proposals to bring about significant and relevant improvements to the reactors. R and D work in France and elsewhere is already indicating orientations that could lead to answers, and improvements that would provide significant reductions in release in case of severe accident are being studied. Second, strict compliance of the reactors with the applicable regulations must be demonstrated. At the same time, ageing and obsolescence of the equipment will have to be managed. Where these two points are concerned, ASN expects far-reaching proposals from the licensee. With a view to a request for continued operation beyond 40 years, ASN has referred the matter to the Advisory Committee for nuclear reactors which will meet at the end of 2011 to establish the safety requirements for reactors at their fourth ten-yearly outage. (author)

  20. 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)

  1. Technical Safety Requirements for the Waste Storage Facilities May 2014

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-04-16

    This document contains the Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Building 693 (B693) Yard Area of the Decontamination and Waste Treatment Facility (DWTF) at LLNL. The TSRs constitute requirements for safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analyses for the Waste Storage Facilities (DSA) (LLNL 2011). The analysis presented therein concluded that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts of waste from other DOE facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities.

  2. Technical Safety Requirements for the Waste Storage Facilities May 2014

    International Nuclear Information System (INIS)

    Laycak, D. T.

    2014-01-01

    This document contains the Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Building 693 (B693) Yard Area of the Decontamination and Waste Treatment Facility (DWTF) at LLNL. The TSRs constitute requirements for safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analyses for the Waste Storage Facilities (DSA) (LLNL 2011). The analysis presented therein concluded that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts of waste from other DOE facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities.

  3. Safety Cultural Competency Modeling in Nuclear Organizations

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sa Kil; Oh, Yeon Ju; Luo, Meiling; Lee, Yong Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear safety cultural competency model should be supplemented through a bottom-up approach such as behavioral event interview. The developed model, however, is meaningful for determining what should be dealt for enhancing safety cultural competency of nuclear organizations. The more details of the developing process, results, and applications will be introduced later. Organizational culture include safety culture in terms of its organizational characteristics.

  4. Requirements on the provisional safety analyses and technical comparison of safety measures

    International Nuclear Information System (INIS)

    2010-04-01

    decide on the provision of a design license for a repository site for SMA and another one for HAA, or for a common site for both SMA and HAA. The present report concerns the second step and recapitulates the assertions of SGT on the provisional safety analyses and on the safety technical comparison. It establishes the specific requirements of the Swiss Federal Nuclear Safety Inspectorate (ENSI) on provisional safety and the safety technical comparison. Further, it defines the extent and content of the safety technical documentation necessary for step 2

  5. [Storage of plant protection products in farms: minimum safety requirements].

    Science.gov (United States)

    Dutto, Moreno; Alfonzo, Santo; Rubbiani, Maristella

    2012-01-01

    Failure to comply with requirements for proper storage and use of pesticides in farms can be extremely hazardous and the risk of accidents involving farm workers, other persons and even animals is high. There are still wide differences in the interpretation of the concept of "securing or making safe", by workers in this sector. One of the critical points detected, particularly in the fruit sector, is the establishment of an adequate storage site for plant protection products. The definition of "safe storage of pesticides" is still unclear despite the recent enactment of Legislative Decree 81/2008 regulating health and work safety in Italy. In addition, there are no national guidelines setting clear minimum criteria for storage of plant protection products in farms. The authors, on the basis of their professional experience and through analysis of recent legislation, establish certain minimum safety standards for storage of pesticides in farms.

  6. Safety-related requirements for photovoltaic modules and arrays

    Science.gov (United States)

    Levins, A.; Smoot, A.; Wagner, R.

    1984-01-01

    Safety requirements for photovoltaic module and panel designs and configurations for residential, intermediate, and large scale applications are investigated. Concepts for safety systems, where each system is a collection of subsystems which together address the total anticipated hazard situation, are described. Descriptions of hardware, and system usefulness and viability are included. A comparison of these systems, as against the provisions of the 1984 National Electrical Code covering photovoltaic systems is made. A discussion of the Underwriters Laboratory UL investigation of the photovoltaic module evaluated to the provisions of the proposed UL standard for plat plate photovoltaic modules and panels is included. Grounding systems, their basis and nature, and the advantages and disadvantages of each are described. The meaning of frame grounding, circuit groundings, and the type of circuit ground are covered.

  7. Evaluation and qualification of novel control techniques with safety requirements

    International Nuclear Information System (INIS)

    Gossner, S.; Wach, D.

    1985-01-01

    The paper discusses the questions related to the assessment and qualification of new I and C-systems. The tasks of nuclear power plant I and Cs as well as the efficiency of the new techniques are reflected. Problems with application of new I and Cs and the state of application in Germany and abroad are addressed. Starting from the essential differencies between conventional and new I and C-systems it is evaluated, if and in which way existing safety requirements can be met and to what extent new requirements need to be formulated. An overall concept has to be developed comprising the definition of graded requirement profiles for design and qualification. Associated qualification procedures and tools have to be adapted, developed and tuned upon each other. (orig./HP) [de

  8. Legal and governmental infrastructure for nuclear, radiation, radioactive waste and transport safety. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    This publication establishes requirements for legal and governmental responsibilities in respect of the safety of nuclear facilities, the safe use of sources of ionizing radiation, radiation protection, the safe management of radioactive waste and the safe transport of radioactive material. Thus, it covers development of the legal framework for establishing a regulatory body and other actions to achieve effective regulatory control of facilities and activities. Other responsibilities are also covered, such as those for developing the necessary support for safety, involvement in securing third party liability and emergency preparedness

  9. Legal and governmental infrastructure for nuclear, radiation, radioactive waste and transport safety. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    This publication establishes requirements for legal and governmental responsibilities in respect of the safety of nuclear facilities, the safe use of sources of ionizing radiation, radiation protection, the safe management of radioactive waste and the safe transport of radioactive material. Thus, it covers development of the legal framework for establishing a regulatory body and other actions to achieve effective regulatory control of facilities and activities. Other responsibilities are also covered, such as those for developing the necessary support for safety, involvement in securing third party liability and emergency preparedness

  10. The actual development of European aviation safety requirements in aviation medicine: prospects of future EASA requirements.

    Science.gov (United States)

    Siedenburg, J

    2009-04-01

    Common Rules for Aviation Safety had been developed under the aegis of the Joint Aviation Authorities in the 1990s. In 2002 the Basic Regulation 1592/2002 was the founding document of a new entity, the European Aviation Safety Agency. Areas of activity were Certification and Maintenance of aircraft. On 18 March the new Basic Regulation 216/2008, repealing the original Basic Regulation was published and applicable from 08 April on. The included Essential Requirements extended the competencies of EASA inter alia to Pilot Licensing and Flight Operations. The future aeromedical requirements will be included as Annex II in another Implementing Regulation on Personnel Licensing. The detailed provisions will be published as guidance material. The proposals for these provisions have been published on 05 June 2008 as NPA 2008- 17c. After public consultation, processing of comments and final adoption the new proposals may be applicable form the second half of 2009 on. A transition period of four year will apply. Whereas the provisions are based on Joint Aviation Requirement-Flight Crew Licensing (JAR-FCL) 3, a new Light Aircraft Pilot Licence (LAPL) project and the details of the associated medical certification regarding general practitioners will be something new in aviation medicine. This paper consists of 6 sections. The introduction outlines the idea of international aviation safety. The second section describes the development of the Joint Aviation Authorities (JAA), the first step to common rules for aviation safety in Europe. The third section encompasses a major change as next step: the foundation of the European Aviation Safety Agency (EASA) and the development of its rules. In the following section provides an outline of the new medical requirements. Section five emphasizes the new concept of a Leisure Pilot Licence. The last section gives an outlook on ongoing rulemaking activities and the opportunities of the public to participate in them.

  11. Safety requirements and feedback of commonly used material handling equipment

    International Nuclear Information System (INIS)

    Pathak, M.K.

    2009-01-01

    Different types of cranes, hoists, chain pulley blocks are the most commonly used material handling equipment in industry along with attachments like chains, wire rope slings, d-shackles, etc. These equipment are used at work for transferring loads from one place to another and attachments are used for anchoring, fixing or supporting the load. Selection of the correct equipment, identification of the equipment planning of material handling operation, examination/testing of the equipment, education and training of the persons engaged in operation of the material handling equipment can reduce the risks to safety of people in workplace. Different safety systems like boom angle indicator, overload tripping device, limit switches, etc. should be available in the cranes for their safe use. Safety requirement for safe operation of material handling equipment with emphasis on different cranes and attachments particularly wire rope slings and chain slings have been brought out in this paper. An attempt has also been made to bring out common nature of deficiencies observed during regulatory inspection carried out by AERB. (author)

  12. 78 FR 65427 - Pipeline Safety: Reminder of Requirements for Liquefied Petroleum Gas and Utility Liquefied...

    Science.gov (United States)

    2013-10-31

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No. PHMSA-2013-0097] Pipeline Safety: Reminder of Requirements for Liquefied Petroleum Gas and Utility Liquefied Petroleum Gas Pipeline Systems AGENCY: Pipeline and Hazardous Materials Safety Administration...

  13. Supervision of nuclear safety - IAEA requirements, accepted solutions, trends

    International Nuclear Information System (INIS)

    Jurkowski, M.

    2007-01-01

    Ten principles of the nuclear safety, based on the IAEA's standards are presented. Convention on Nuclear Safety recommends for nuclear safety landscape, the control transparency, culture safety, legal framework and knowledge preservation. Examples of solutions accepted in France, Finland, and Czech Republic are discussed. New trends in safety fundamentals and Integration Regulatory Review are presented

  14. Technical Safety Requirements for the B695 Segment

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-09-11

    This document contains Technical Safety Requirements (TSRs) for the Radioactive and Hazardous Waste Management (RHWM) Division's B695 Segment of the Decontamination and Waste Treatment Facility (DWTF) at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the B695 Segment. The TSRs are derived from the Documented Safety Analysis (DSA) for the B695 Segment (LLNL 2007). The analysis presented there determined that the B695 Segment is a low-chemical hazard, Hazard Category 3, nonreactor nuclear facility. The TSRs consist primarily of inventory limits as well as controls to preserve the underlying assumptions in the hazard analyses. Furthermore, appropriate commitments to safety programs are presented in the administrative controls section of the TSRs. The B695 Segment (B695 and the west portion of B696) is a waste treatment and storage facility located in the northeast quadrant of the LLNL main site. The approximate area and boundary of the B695 Segment are shown in the B695 Segment DSA. Activities typically conducted in the B695 Segment include container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. B695 is used to store and treat radioactive, mixed, and hazardous waste, and it also contains equipment used in conjunction with waste processing operations to treat various liquid and solid wastes. The portion of the building called Building 696 Solid Waste Processing Area (SWPA), also referred to as B696S in this report, is used primarily to manage solid radioactive, mixed, and hazardous waste. Operations specific to the SWPA include sorting and segregating waste, lab-packing, sampling, and crushing empty drums that previously contained waste. Furthermore, a Waste Packaging Unit will be permitted to treat hazardous and mixed waste. RHWM generally processes LLW with no, or extremely low, concentrations of transuranics (i.e., much less than 100 n

  15. Technical Safety Requirements for the B695 Segment

    International Nuclear Information System (INIS)

    Laycak, D.

    2008-01-01

    This document contains Technical Safety Requirements (TSRs) for the Radioactive and Hazardous Waste Management (RHWM) Division's B695 Segment of the Decontamination and Waste Treatment Facility (DWTF) at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the B695 Segment. The TSRs are derived from the Documented Safety Analysis (DSA) for the B695 Segment (LLNL 2007). The analysis presented there determined that the B695 Segment is a low-chemical hazard, Hazard Category 3, nonreactor nuclear facility. The TSRs consist primarily of inventory limits as well as controls to preserve the underlying assumptions in the hazard analyses. Furthermore, appropriate commitments to safety programs are presented in the administrative controls section of the TSRs. The B695 Segment (B695 and the west portion of B696) is a waste treatment and storage facility located in the northeast quadrant of the LLNL main site. The approximate area and boundary of the B695 Segment are shown in the B695 Segment DSA. Activities typically conducted in the B695 Segment include container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. B695 is used to store and treat radioactive, mixed, and hazardous waste, and it also contains equipment used in conjunction with waste processing operations to treat various liquid and solid wastes. The portion of the building called Building 696 Solid Waste Processing Area (SWPA), also referred to as B696S in this report, is used primarily to manage solid radioactive, mixed, and hazardous waste. Operations specific to the SWPA include sorting and segregating waste, lab-packing, sampling, and crushing empty drums that previously contained waste. Furthermore, a Waste Packaging Unit will be permitted to treat hazardous and mixed waste. RHWM generally processes LLW with no, or extremely low, concentrations of transuranics (i.e., much less than 100 n

  16. Plasma-safety assessment model and safety analyses of ITER

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

    2001-01-01

    A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

  17. Decommissioning of Facilities. General Safety Requirements. Pt. 6

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-15

    Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning.

  18. Decommissioning of Facilities. General Safety Requirements. Pt. 6 (Spanish Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning.

  19. Decommissioning of Facilities. General Safety Requirements. Pt. 6 (Russian Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning

  20. Application of Safety Maturity Model and 4P-4C Model in Safety Culture Assessment

    International Nuclear Information System (INIS)

    Choi, K. S.; Lee, Y. E.; Ha, J. T.; Chang, H. S.; Kam, S. C.

    2010-01-01

    Korean government and utility have made efforts to enhance the nuclear safety culture and the development of quantitative index of safety culture was promoted for past several years. Quantitative index of safety culture and the past efforts to understand safety culture need insight into the concept of culture. This paper aims to apply new method of measuring nuclear safety culture through the review of approaches of evaluating safety culture in non-nuclear industries. Scoring table has been developed based on new models and example of result of interviews evaluating the nuclear safety culture is also shown

  1. Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. General Safety Requirements. Pt. 3 (Chinese Edition)

    International Nuclear Information System (INIS)

    2014-01-01

    This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

  2. Radiation protection and safety of radiation sources: International basic safety standards. General safety requirements. Pt. 3 (French Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

  3. Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. General Safety Requirements. Pt. 3 (Arabic Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

  4. Standard model for safety analysis report of fuel fabrication plants

    International Nuclear Information System (INIS)

    1980-09-01

    A standard model for a safety analysis report of fuel fabrication plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.) [pt

  5. Standard model for safety analysis report of fuel reprocessing plants

    International Nuclear Information System (INIS)

    1979-12-01

    A standard model for a safety analysis report of fuel reprocessing plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.) [pt

  6. Diversity requirements for safety critical software-based automation systems

    International Nuclear Information System (INIS)

    Korhonen, J.; Pulkkinen, U.; Haapanen, P.

    1998-03-01

    System vendors nowadays propose software-based systems even for the most critical safety functions in nuclear power plants. Due to the nature and mechanisms of influence of software faults new methods are needed for the safety and reliability evaluation of these systems. In the research project 'Programmable automation systems in nuclear power plants (OHA)' various safety assessment methods and tools for software based systems are developed and evaluated. This report first discusses the (common cause) failure mechanisms in software-based systems, then defines fault-tolerant system architectures to avoid common cause failures, then studies the various alternatives to apply diversity and their influence on system reliability. Finally, a method for the assessment of diversity is described. Other recently published reports in OHA-report series handles the statistical reliability assessment of software based (STUK-YTO-TR 119), usage models in reliability assessment of software-based systems (STUK-YTO-TR 128) and handling of programmable automation in plant PSA-studies (STUK-YTO-TR 129)

  7. MSSV Modeling for Wolsong-1 Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Bok Ja; Choi, Chul Jin; Kim, Seoung Rae [KEPCO EandC, Daejeon (Korea, Republic of)

    2010-10-15

    The main steam safety valves (MSSVs) are installed on the main steam line to prevent the overpressurization of the system. MSSVs are held in closed position by spring force and the valves pop open by internal force when the main steam pressure increases to open set pressure. If the overpressure condition is relieved, the valves begin to close. For the safety analysis of anticipated accident condition, the safety systems are modeled conservatively to simulate the accident condition more severe. MSSVs are also modeled conservatively for the analysis of over-pressurization accidents. In this paper, the pressure transient is analyzed at over-pressurization condition to evaluate the conservatism for MSSV models

  8. 42 CFR 3.210 - Required disclosure of patient safety work product to the Secretary.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Required disclosure of patient safety work product... HUMAN SERVICES GENERAL PROVISIONS PATIENT SAFETY ORGANIZATIONS AND PATIENT SAFETY WORK PRODUCT Confidentiality and Privilege Protections of Patient Safety Work Product § 3.210 Required disclosure of patient...

  9. Technical safety requirements for the Annular Core Research Reactor Facility (ACRRF)

    International Nuclear Information System (INIS)

    Boldt, K.R.; Morris, F.M.; Talley, D.G.; McCrory, F.M.

    1998-01-01

    The Technical Safety Requirements (TSR) document is prepared and issued in compliance with DOE Order 5480.22, Technical Safety Requirements. The bases for the TSR are established in the ACRRF Safety Analysis Report issued in compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports. The TSR identifies the operational conditions, boundaries, and administrative controls for the safe operation of the facility

  10. An optimization model for improving highway safety

    Directory of Open Access Journals (Sweden)

    Promothes Saha

    2016-12-01

    Full Text Available This paper developed a traffic safety management system (TSMS for improving safety on county paved roads in Wyoming. TSMS is a strategic and systematic process to improve safety of roadway network. When funding is limited, it is important to identify the best combination of safety improvement projects to provide the most benefits to society in terms of crash reduction. The factors included in the proposed optimization model are annual safety budget, roadway inventory, roadway functional classification, historical crashes, safety improvement countermeasures, cost and crash reduction factors (CRFs associated with safety improvement countermeasures, and average daily traffics (ADTs. This paper demonstrated how the proposed model can identify the best combination of safety improvement projects to maximize the safety benefits in terms of reducing overall crash frequency. Although the proposed methodology was implemented on the county paved road network of Wyoming, it could be easily modified for potential implementation on the Wyoming state highway system. Other states can also benefit by implementing a similar program within their jurisdictions.

  11. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition); Bezopasnost' atomnykh ehlektrostantsij: proektirovanie. Konkretnye trebovaniya bezopasnosti

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  12. System theory and safety models in Swedish, UK, Dutch and Australian road safety strategies.

    Science.gov (United States)

    Hughes, B P; Anund, A; Falkmer, T

    2015-01-01

    Road safety strategies represent interventions on a complex social technical system level. An understanding of a theoretical basis and description is required for strategies to be structured and developed. Road safety strategies are described as systems, but have not been related to the theory, principles and basis by which systems have been developed and analysed. Recently, road safety strategies, which have been employed for many years in different countries, have moved to a 'vision zero', or 'safe system' style. The aim of this study was to analyse the successful Swedish, United Kingdom and Dutch road safety strategies against the older, and newer, Australian road safety strategies, with respect to their foundations in system theory and safety models. Analysis of the strategies against these foundations could indicate potential improvements. The content of four modern cases of road safety strategy was compared against each other, reviewed against scientific systems theory and reviewed against types of safety model. The strategies contained substantial similarities, but were different in terms of fundamental constructs and principles, with limited theoretical basis. The results indicate that the modern strategies do not include essential aspects of systems theory that describe relationships and interdependencies between key components. The description of these strategies as systems is therefore not well founded and deserves further development. Copyright © 2014 Elsevier Ltd. All rights reserved.

  13. Technical Safety Requirements for the Waste Storage Facilities

    International Nuclear Information System (INIS)

    Larson, H L

    2007-01-01

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 612 (A612) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2006). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., drum crushing, size reduction, and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A612 is located in the southeast quadrant of LLNL. The A612 fenceline is approximately 220 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A612 and the DWTF Storage Area are subdivided into various facilities and storage

  14. Technical Safety Requirements for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Larson, H L

    2007-09-07

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 612 (A612) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2006). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., drum crushing, size reduction, and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A612 is located in the southeast quadrant of LLNL. The A612 fenceline is approximately 220 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A612 and the DWTF Storage Area are subdivided into various facilities and storage

  15. Safety requirements and radiological protection for ore installations

    International Nuclear Information System (INIS)

    2003-06-01

    This norm establishes the safety and radiological protection requirements for mining installations which manipulates, process and storing ores, raw materials, steriles, slags and wastes containing radionuclides of the uranium and thorium natural series, simultaneously or separated, and which can cause undue exposures to the public and workers, at anytime of the functioning or pos operational stage. This norm applies to the mining installations activities, suspended or which have ceased their activities before the issue date of this norm, destined to the mining, physical, chemical and metallurgical processing, and the industrialization of raw materials and residues containing associated radionuclides from the natural series of uranium and thorium, including the stages of implantation, operation and decommissioning of the installation

  16. NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

    International Nuclear Information System (INIS)

    Marshall, A.C.; Lee, J.H.; McCulloch, W.H.; Sawyer, J.C. Jr.; Bari, R.A.; Brown, N.W.; Cullingford, H.S.; Hardy, A.C.; Remp, K.; Sholtis, J.A.

    1992-01-01

    An Interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top- level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed

  17. Model uncertainty in safety assessment

    International Nuclear Information System (INIS)

    Pulkkinen, U.; Huovinen, T.

    1996-01-01

    The uncertainty analyses are an essential part of any risk assessment. Usually the uncertainties of reliability model parameter values are described by probability distributions and the uncertainty is propagated through the whole risk model. In addition to the parameter uncertainties, the assumptions behind the risk models may be based on insufficient experimental observations and the models themselves may not be exact descriptions of the phenomena under analysis. The description and quantification of this type of uncertainty, model uncertainty, is the topic of this report. The model uncertainty is characterized and some approaches to model and quantify it are discussed. The emphasis is on so called mixture models, which have been applied in PSAs. Some of the possible disadvantages of the mixture model are addressed. In addition to quantitative analyses, also qualitative analysis is discussed shortly. To illustrate the models, two simple case studies on failure intensity and human error modeling are described. In both examples, the analysis is based on simple mixture models, which are observed to apply in PSA analyses. (orig.) (36 refs., 6 figs., 2 tabs.)

  18. Model uncertainty in safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Pulkkinen, U; Huovinen, T [VTT Automation, Espoo (Finland). Industrial Automation

    1996-01-01

    The uncertainty analyses are an essential part of any risk assessment. Usually the uncertainties of reliability model parameter values are described by probability distributions and the uncertainty is propagated through the whole risk model. In addition to the parameter uncertainties, the assumptions behind the risk models may be based on insufficient experimental observations and the models themselves may not be exact descriptions of the phenomena under analysis. The description and quantification of this type of uncertainty, model uncertainty, is the topic of this report. The model uncertainty is characterized and some approaches to model and quantify it are discussed. The emphasis is on so called mixture models, which have been applied in PSAs. Some of the possible disadvantages of the mixture model are addressed. In addition to quantitative analyses, also qualitative analysis is discussed shortly. To illustrate the models, two simple case studies on failure intensity and human error modeling are described. In both examples, the analysis is based on simple mixture models, which are observed to apply in PSA analyses. (orig.) (36 refs., 6 figs., 2 tabs.).

  19. Regulatory framework and safety requirements for new (gen III) reactors

    International Nuclear Information System (INIS)

    Mourlon, Sophie

    2014-01-01

    Sophie Mourlon, ASN Deputy General Director, described the international process to enhance safety between local safety authorities through organizations such as WENRA. Then she explained to the participants the regulatory issues for the next generation of NPPs

  20. Special safety requirements applied to Brazilian nuclear power plant

    International Nuclear Information System (INIS)

    Lepecki, W.P.S.; Hamel, H.J.E.; Koenig, N.; Vieira, P.C.R.; Fritzsche, J.C.

    1981-01-01

    Some safety aspects of the Angra 2 and 3 nuclear power plants are presented. An analysis of the civil and mechanical project of these nuclear power plant having in view a safety analysis is done. (E.G.) [pt

  1. Time series modeling in traffic safety research.

    Science.gov (United States)

    Lavrenz, Steven M; Vlahogianni, Eleni I; Gkritza, Konstantina; Ke, Yue

    2018-08-01

    The use of statistical models for analyzing traffic safety (crash) data has been well-established. However, time series techniques have traditionally been underrepresented in the corresponding literature, due to challenges in data collection, along with a limited knowledge of proper methodology. In recent years, new types of high-resolution traffic safety data, especially in measuring driver behavior, have made time series modeling techniques an increasingly salient topic of study. Yet there remains a dearth of information to guide analysts in their use. This paper provides an overview of the state of the art in using time series models in traffic safety research, and discusses some of the fundamental techniques and considerations in classic time series modeling. It also presents ongoing and future opportunities for expanding the use of time series models, and explores newer modeling techniques, including computational intelligence models, which hold promise in effectively handling ever-larger data sets. The information contained herein is meant to guide safety researchers in understanding this broad area of transportation data analysis, and provide a framework for understanding safety trends that can influence policy-making. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. Technical Safety Requirements for the Waste Storage Facilities

    International Nuclear Information System (INIS)

    Laycak, D.T.

    2010-01-01

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2009). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas, consisting

  3. Technical Safety Requirements for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D T

    2008-06-16

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the 'Documented Safety Analysis for the Waste Storage Facilities' (DSA) (LLNL 2008). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas

  4. Regulator Loss Functions and Hierarchical Modeling for Safety Decision Making.

    Science.gov (United States)

    Hatfield, Laura A; Baugh, Christine M; Azzone, Vanessa; Normand, Sharon-Lise T

    2017-07-01

    Regulators must act to protect the public when evidence indicates safety problems with medical devices. This requires complex tradeoffs among risks and benefits, which conventional safety surveillance methods do not incorporate. To combine explicit regulator loss functions with statistical evidence on medical device safety signals to improve decision making. In the Hospital Cost and Utilization Project National Inpatient Sample, we select pediatric inpatient admissions and identify adverse medical device events (AMDEs). We fit hierarchical Bayesian models to the annual hospital-level AMDE rates, accounting for patient and hospital characteristics. These models produce expected AMDE rates (a safety target), against which we compare the observed rates in a test year to compute a safety signal. We specify a set of loss functions that quantify the costs and benefits of each action as a function of the safety signal. We integrate the loss functions over the posterior distribution of the safety signal to obtain the posterior (Bayes) risk; the preferred action has the smallest Bayes risk. Using simulation and an analysis of AMDE data, we compare our minimum-risk decisions to a conventional Z score approach for classifying safety signals. The 2 rules produced different actions for nearly half of hospitals (45%). In the simulation, decisions that minimize Bayes risk outperform Z score-based decisions, even when the loss functions or hierarchical models are misspecified. Our method is sensitive to the choice of loss functions; eliciting quantitative inputs to the loss functions from regulators is challenging. A decision-theoretic approach to acting on safety signals is potentially promising but requires careful specification of loss functions in consultation with subject matter experts.

  5. NASA's Aviation Safety and Modeling Project

    Science.gov (United States)

    Chidester, Thomas R.; Statler, Irving C.

    2006-01-01

    The Aviation Safety Monitoring and Modeling (ASMM) Project of NASA's Aviation Safety program is cultivating sources of data and developing automated computer hardware and software to facilitate efficient, comprehensive, and accurate analyses of the data collected from large, heterogeneous databases throughout the national aviation system. The ASMM addresses the need to provide means for increasing safety by enabling the identification and correcting of predisposing conditions that could lead to accidents or to incidents that pose aviation risks. A major component of the ASMM Project is the Aviation Performance Measuring System (APMS), which is developing the next generation of software tools for analyzing and interpreting flight data.

  6. Modeling patient safety incidents knowledge with the Categorial Structure method.

    Science.gov (United States)

    Souvignet, Julien; Bousquet, Cédric; Lewalle, Pierre; Trombert-Paviot, Béatrice; Rodrigues, Jean Marie

    2011-01-01

    Following the WHO initiative named World Alliance for Patient Safety (PS) launched in 2004 a conceptual framework developed by PS national reporting experts has summarized the knowledge available. As a second step, the Department of Public Health of the University of Saint Etienne team elaborated a Categorial Structure (a semi formal structure not related to an upper level ontology) identifying the elements of the semantic structure underpinning the broad concepts contained in the framework for patient safety. This knowledge engineering method has been developed to enable modeling patient safety information as a prerequisite for subsequent full ontology development. The present article describes the semantic dissection of the concepts, the elicitation of the ontology requirements and the domain constraints of the conceptual framework. This ontology includes 134 concepts and 25 distinct relations and will serve as basis for an Information Model for Patient Safety.

  7. Investigational new drug safety reporting requirements for human drug and biological products and safety reporting requirements for bioavailability and bioequivalence studies in humans. Final rule.

    Science.gov (United States)

    2010-09-29

    The Food and Drug Administration (FDA) is amending its regulations governing safety reporting requirements for human drug and biological products subject to an investigational new drug application (IND). The final rule codifies the agency's expectations for timely review, evaluation, and submission of relevant and useful safety information and implements internationally harmonized definitions and reporting standards. The revisions will improve the utility of IND safety reports, reduce the number of reports that do not contribute in a meaningful way to the developing safety profile of the drug, expedite FDA's review of critical safety information, better protect human subjects enrolled in clinical trials, subject bioavailability and bioequivalence studies to safety reporting requirements, promote a consistent approach to safety reporting internationally, and enable the agency to better protect and promote public health.

  8. What Isn't Working and New Requirements. The Need to Harmonize Safety and Security Requirements

    International Nuclear Information System (INIS)

    Flory, D.

    2011-01-01

    The year 2011 marks the 50th anniversary of the first IAEA regulations governing the transport of radioactive material. However transport safety at the IAEA obviously predates this, since the regulations took time to develop. In 1957, GC. 1/1 already states: 'The Agency should undertake studies with a view to the establishment of regulations relating to the international transportation of radioactive materials. ...'. And goes further: 'The transport of radioisotopes and radiation sources has brought to light many problems and involves the need for uniform packaging and shipping regulations ... facilitate the acceptance of such materials by sea and air carriers'. This conference reiterates the challenge given then through the sub-title 'The next fifty years - Creating a Safe, Secure and Sustainable Framework'. Looking back, we can see that the sustainable framework was a goal in 1957, where radioactive material could be transported should it be desired. Since these early days we have added to safety the need to ensure security. However we still see the same calls today to eradicate denial of shipment, which might suggest we have not progressed. But the picture today is very different - we have today well established requirements for safe transport of radioactive material, and the recommendations for security in transport are coming of age for all radioactive materials. The outstanding issue would seem to be harmonisation, not just between safety and security in IAEA documents, but also harmonisation between Member States.

  9. JET-ISX-B beryllium limiter experiment safety analysis report and operational safety requirements

    International Nuclear Information System (INIS)

    Edmonds, P.H.

    1985-09-01

    An experiment to evaluate the suitability of beryllium as a limiter material has been completed on the ISX-B tokamak. The experiment consisted of two phases: (1) the initial operation and characterization in the ISX experiment, and a period of continued operation to the specified surface fluence (10 22 atoms/cm 2 ) of hydrogen ions; and (2) the disassembly, decontamination, or disposal of the ISX facility. During these two phases of the project, the possibility existed for beryllium and/or beryllium oxide powder to be produced inside the vacuum vessel. Beryllium dust is a highly toxic material, and extensive precautions are required to prevent the release of the beryllium into the experimental work area and to prevent the contamination of personnel working on the device. Details of the health hazards associated with beryllium and the appropriate precautions are presented. Also described in appendixes to this report are the various operational safety requirements for the project

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

    International Nuclear Information System (INIS)

    Hasnaoui, C.; Parent, G.

    2000-01-01

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

  11. Model for safety reports including descriptive examples

    International Nuclear Information System (INIS)

    1995-12-01

    Several safety reports will be produced in the process of planning and constructing the system for disposal of high-level radioactive waste in Sweden. The present report gives a model, with detailed examples, of how these reports should be organized and what steps they should include. In the near future safety reports will deal with the encapsulation plant and the repository. Later reports will treat operation of the handling systems and the repository

  12. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

  13. Model-based testing for software safety

    NARCIS (Netherlands)

    Gurbuz, Havva Gulay; Tekinerdogan, Bedir

    2017-01-01

    Testing safety-critical systems is crucial since a failure or malfunction may result in death or serious injuries to people, equipment, or environment. An important challenge in testing is the derivation of test cases that can identify the potential faults. Model-based testing adopts models of a

  14. Does unbelted safety requirement affect protection for belted occupants?

    Science.gov (United States)

    Hu, Jingwen; Klinich, Kathleen D; Manary, Miriam A; Flannagan, Carol A C; Narayanaswamy, Prabha; Reed, Matthew P; Andreen, Margaret; Neal, Mark; Lin, Chin-Hsu

    2017-05-29

    Federal regulations in the United States require vehicles to meet occupant performance requirements with unbelted test dummies. Removing the test requirements with unbelted occupants might encourage the deployment of seat belt interlocks and allow restraint optimization to focus on belted occupants. The objective of this study is to compare the performance of restraint systems optimized for belted-only occupants with those optimized for both belted and unbelted occupants using computer simulations and field crash data analyses. In this study, 2 validated finite element (FE) vehicle/occupant models (a midsize sedan and a midsize SUV) were selected. Restraint design optimizations under standardized crash conditions (U.S.-NCAP and FMVSS 208) with and without unbelted requirements were conducted using Hybrid III (HIII) small female and midsize male anthropomorphic test devices (ATDs) in both vehicles on both driver and right front passenger positions. A total of 10 to 12 design parameters were varied in each optimization using a combination of response surface method (RSM) and genetic algorithm. To evaluate the field performance of restraints optimized with and without unbelted requirements, 55 frontal crash conditions covering a greater variety of crash types than those in the standardized crashes were selected. A total of 1,760 FE simulations were conducted for the field performance evaluation. Frontal crashes in the NASS-CDS database from 2002 to 2012 were used to develop injury risk curves and to provide the baseline performance of current restraint system and estimate the injury risk change by removing the unbelted requirement. Unbelted requirements do not affect the optimal seat belt and airbag design parameters in 3 out of 4 vehicle/occupant position conditions, except for the SUV passenger side. Overall, compared to the optimal designs with unbelted requirements, optimal designs without unbelted requirements generated the same or lower total injury risks for

  15. Data concentrator requirements for a safety parameter display system

    International Nuclear Information System (INIS)

    Brewer, C.R.

    1983-01-01

    To comply with NUREG 0696 several nuclear plants are being fitted with new facilities and data systems; specifically a Technical Support Center (TSC), Operational Support Center (OSC), Emergency Operational Facility (EOF), and Backup Safety Parameter Display System (SPDS), Emergency Response Computer System (ERCS) and Nuclear Data Link (NDL). The TSC, OSC, and EOF are physical locations while the SPDS, ERCS, and NDL are Systems. The SPDS and ERCS are usually separate and independent systems, however, they may share a common front end data acquisition system that acquires and sends SPDS related data to both the SPDS and to the ERCS. In the situation just described an SPDS system must depend upon input data from a source that is SPDS host computer independent. To achieve this independence the front end data acquisition system may employ a concept of intelligent distributed processing. This concept essentially takes functional capabilities that were once found only in realtime host computers and distributes it to front end data acquisition systems. Thus by expanding the functionality of the data acquisition system in a manner that provides more capability, independence from the computer vendor, links to multiple computer systems, processing power and redundancy, the concept of a data concentrator evolved. This paper will define this new distributed functionality, and its related requirements. It will also examine different system configuration approaches

  16. Regulatory considerations for computational requirements for nuclear criticality safety

    International Nuclear Information System (INIS)

    Bidinger, G.H.

    1995-01-01

    As part of its safety mission, the U.S. Nuclear Regulatory Commission (NRC) approves the use of computational methods as part of the demonstration of nuclear criticality safety. While each NRC office has different criteria for accepting computational methods for nuclear criticality safety results, the Office of Nuclear Materials Safety and Safeguards (NMSS) approves the use of specific computational methods and methodologies for nuclear criticality safety analyses by specific companies (licensees or consultants). By contrast, the Office of Nuclear Reactor Regulation approves codes for general use. Historically, computational methods progressed from empirical methods to one-dimensional diffusion and discrete ordinates transport calculations and then to three-dimensional Monte Carlo transport calculations. With the advent of faster computational ability, three-dimensional diffusion and discrete ordinates transport calculations are gaining favor. With the proper user controls, NMSS has accepted any and all of these methods for demonstrations of nuclear criticality safety

  17. Requirement and prospect of nuclear data activities for nuclear safety

    International Nuclear Information System (INIS)

    Kimura, Itsuro

    2000-01-01

    Owing to continuous efforts by the members of JNDC (Japanese Nuclear Data Committee) and Nuclear Data Center in JAERI (Japan Atomic Energy Research Institute), several superb evaluated nuclear data files, such as JENDL, FP (fission product) yields and decay heat, have been compiled in Japan and opened to the world. However, they are seldom adopted in safety design and safety evaluation of light water reactors and are hardly found in related safety regulatory guidelines and standards except the decay heat. In this report, shown are a few examples of presently used nuclear data in the safety design and the safety evaluation of PWRs (pressurized water reactors) and so forth. And then, several procedures are recommended in order to enhance more utilization of Japanese evaluated nuclear data files for nuclear safety. (author)

  18. Conducting organizational safety reviews - requirements, methods and experience

    International Nuclear Information System (INIS)

    Reiman, T.; Oedewald, P.; Wahlstroem, B.; Rollenhagen, C.; Kahlbom, U.

    2008-03-01

    Organizational safety reviews are part of the safety management process of power plants. They are typically performed after major reorganizations, significant incidents or according to specified review programs. Organizational reviews can also be a part of a benchmarking between organizations that aims to improve work practices. Thus, they are important instruments in proactive safety management and safety culture. Most methods that have been used for organizational reviews are based more on practical considerations than a sound scientific theory of how various organizational or technical issues influence safety. Review practices and methods also vary considerably. The objective of this research is to promote understanding on approaches used in organizational safety reviews as well as to initiate discussion on criteria and methods of organizational assessment. The research identified a set of issues that need to be taken into account when planning and conducting organizational safety reviews. Examples of the issues are definition of appropriate criteria for evaluation, the expertise needed in the assessment and the organizational motivation for conducting the assessment. The study indicates that organizational safety assessments involve plenty of issues and situations where choices have to be made regarding what is considered valid information and a balance has to be struck between focus on various organizational phenomena. It is very important that these choices are based on a sound theoretical framework and that these choices can later be evaluated together with the assessment findings. The research concludes that at its best, the organizational safety reviews can be utilised as a source of information concerning the changing vulnerabilities and the actual safety performance of the organization. In order to do this, certain basic organizational phenomena and assessment issues have to be acknowledged and considered. The research concludes with recommendations on

  19. Conducting organizational safety reviews - requirements, methods and experience

    Energy Technology Data Exchange (ETDEWEB)

    Reiman, T.; Oedewald, P.; Wahlstroem, B. [Technical Research Centre of Finland, VTT (Finland); Rollenhagen, C. [Royal Institute of Technology, KTH, (Sweden); Kahlbom, U. [RiskPilot (Sweden)

    2008-03-15

    Organizational safety reviews are part of the safety management process of power plants. They are typically performed after major reorganizations, significant incidents or according to specified review programs. Organizational reviews can also be a part of a benchmarking between organizations that aims to improve work practices. Thus, they are important instruments in proactive safety management and safety culture. Most methods that have been used for organizational reviews are based more on practical considerations than a sound scientific theory of how various organizational or technical issues influence safety. Review practices and methods also vary considerably. The objective of this research is to promote understanding on approaches used in organizational safety reviews as well as to initiate discussion on criteria and methods of organizational assessment. The research identified a set of issues that need to be taken into account when planning and conducting organizational safety reviews. Examples of the issues are definition of appropriate criteria for evaluation, the expertise needed in the assessment and the organizational motivation for conducting the assessment. The study indicates that organizational safety assessments involve plenty of issues and situations where choices have to be made regarding what is considered valid information and a balance has to be struck between focus on various organizational phenomena. It is very important that these choices are based on a sound theoretical framework and that these choices can later be evaluated together with the assessment findings. The research concludes that at its best, the organizational safety reviews can be utilised as a source of information concerning the changing vulnerabilities and the actual safety performance of the organization. In order to do this, certain basic organizational phenomena and assessment issues have to be acknowledged and considered. The research concludes with recommendations on

  20. Discussion on several important safety requirements for the new nuclear power plant

    International Nuclear Information System (INIS)

    Yan Tianwen; Li Jigen; Zhang Lin; Feng Youcai; Jia Xiang; Li Wenhong

    2013-01-01

    Post the Fukushima nuclear accident, the Chinese government raised higher safety goals and safety requirements for the new nuclear power plant to be constructed. The paper expounded the important indicators of safety requirements and the aspects of safety modification that had been developed for the new NPPs. It also discussed and analyzed the main fields required by the new NPPs safety requirements in the safety goals, safety evaluation of sites, defenses of internal and external events, severe accident prevention and mitigation, design of reactor core, containment system and I and C system, and optimization of engineering measure, which gave some references to the design, construction and safety modifications of new NPPs in China. (authors)

  1. Regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning

    International Nuclear Information System (INIS)

    Lipar, M.

    1997-01-01

    A review of regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning is given. It contains licensing steps in Slovakia during commissioning; Status and methodology of Mochovce safety analysis report; Mochovce NPP safety enhancement program; Regulatory body policy towards Mochovce NPP safety enhancement; Recent development in Mochovce pre-operational safety enhancement program review and assessment process; Licensing steps in Slovakia during commissioning

  2. Legal requirements concerning the technical safety of nuclear installations

    International Nuclear Information System (INIS)

    Nolte, R.

    1984-01-01

    A short survey on nuclear risks and the nuclear safety conception is followed by the attempted clear definition of the semantic import of section 7, sub-section (2), No. 3 of the Atomic Energy Act. There are first beginnings of a concretization of the state-of-the-art in science and technology, i.e. all kinds of sub-legislative regulations such as the regulations of the Radiation Protection Ordinance which show scientific substance, guidelines issued by the Ministers, as well as codes for practice set up by various technical bodies and standardization associations, all of which are designed to compensate for this loop hole in the legislation. This study goes to examine to what extent administration and jurisdiction may take into account such codes of practice for the concretization of the legal requirements, and whether they are even binding on those executing the law. Only the respective regulations of the Radiation Protection Ordinance have a binding effect. All other guidelines and codes of practice are not legally binding per se, nor are they capable of being legally permitted by being referred to in terms of legal norms or by the self-commitment of those executing the law. Any attempt of using them, as the basis of a prime facie evidence or as an anticipating expertise, at least evidentarily for the concretization will have to fail owing to their evaluating character and to the fact that they may interfere in sociological conflict. An exception may be a case where a clear distinction can be made as to what extent the contents of such codes of practice is related to scientific and technological findings or to decisions based on evaluations. In such a case, a prima facil evicdence for the conformity of the regulation in question with the state-of-the-art in science and technology may be considered, which would easy the concretization of Art. 7 II Section 3 of the Atomic Law. (orig./HSCH) [de

  3. 77 FR 75439 - Guidances for Industry and Investigators on Safety Reporting Requirements for Investigational New...

    Science.gov (United States)

    2012-12-20

    ...] Guidances for Industry and Investigators on Safety Reporting Requirements for Investigational New Drug Applications and Bioavailability/Bioequivalence Studies, and a Small Entity Compliance Guide; Availability... Reporting Requirements for INDs and BA/BE Studies'' and ``Safety Reporting Requirements for INDs and BA/BE...

  4. 10 CFR 76.87 - Technical safety requirements.

    Science.gov (United States)

    2010-01-01

    ...: (1) Effects of natural phenomena; (2) Building and process ventilation and offgas; (3) Criticality...; (8) Environmental protection; (9) Packaging and transporting nuclear materials; (10) Accident analysis; (11) Chemical safety; (12) Sharing of facilities, structures, systems and components; (13...

  5. Safety evaluations required in the safety regulations for Monju and the validity confirmation of safety evaluation methods

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to perform the safety evaluations of the fast breeder reactor 'Monju' and to confirm the validity of the safety evaluation methods. In JFY 2012, the following results were obtained. As for the development of safety evaluation methods needed in the safety examination achieved for the reactor establishment permission, development of the analysis codes, such as a core damage analysis code, were carried out according to the plan. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

  6. Small nuclear reactor safety design requirements for autonomous operation

    International Nuclear Information System (INIS)

    Kozier, K.S.; Kupca, S.

    1997-01-01

    Small nuclear power reactors offer compelling safety advantages in terms of the limited consequences that can arise from major accident events and the enhanced ability to use reliable, passive means to eliminate their occurrence by design. Accordingly, for some small reactor designs featuring a high degree of safety autonomy, it may be-possible to delineate a ''safety envelope'' for a given set of reactor circumstances within which safe reactor operation can be guaranteed without outside intervention for time periods of practical significance (i.e., days or weeks). The capability to operate a small reactor without the need for highly skilled technical staff permanently present, but with continuous remote monitoring, would aid the economic case for small reactors, simplify their use in remote regions and enhance safety by limiting the potential for accidents initiated by inappropriate operator action. This paper considers some of the technical design options and issues associated with the use of small power reactors in an autonomous mode for limited periods. The focus is on systems that are suitable for a variety of applications, producing steam for electricity generation, district heating, water desalination and/or marine propulsion. Near-term prospects at low power levels favour the use of pressurized, light-water-cooled reactor designs, among which those having an integral core arrangement appear to offer cost and passive-safety advantages. Small integral pressurized water reactors have been studied in many countries, including the test operation of prototype systems. (author)

  7. 75 FR 60129 - Draft Guidance for Industry and Investigators on Safety Reporting Requirements for...

    Science.gov (United States)

    2010-09-29

    ...., Bldg. 51, rm. 2201, Silver Spring, MD 20993-0002; or the Office of Communication, Outreach, and...'s ability to review critical safety information, improve safety monitoring of human drug and..., will represent the Agency's current thinking on safety reporting requirements for INDs and BA/BE...

  8. 42 CFR 9.10 - Occupational Health and Safety Program (OHSP) and biosafety requirements.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Occupational Health and Safety Program (OHSP) and... SANCTUARY SYSTEM § 9.10 Occupational Health and Safety Program (OHSP) and biosafety requirements. (a) How are employee Occupational Health and Safety Program risks and concerns addressed? The sanctuary shall...

  9. Evaluation of safety, an unavoidable requirement in the applications of ionizing radiations

    International Nuclear Information System (INIS)

    Jova Sed, Luis Andres

    2013-01-01

    The safety assessments should be conducted as a means to evaluate compliance with safety requirements (and thus the application of fundamental safety principles) for all facilities and activities in order to determine the measures to be taken to ensure safety. It is an essential tool in decision making. For long time we have linked the safety assessment to nuclear facilities and not to all practices involving the use of ionizing radiation in daily life. However, the main purpose of the safety assessment is to determine if it has reached an appropriate level of safety for an installation or activity and if it has fulfilled the objectives of safety and basic safety criteria set by the designer, operating organization and the regulatory body under the protection and safety requirements set out in the International Basic safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. This paper presents some criteria and personal experiences with the new international recommendations on this subject and its practical application in the region and demonstrates the importance of this requirement. Reflects the need to train personnel of the operator and the regulatory body in the proportional application of this requirement in practice with ionizing radiation

  10. Nuclear safety culture evaluation model based on SSE-CMM

    International Nuclear Information System (INIS)

    Yang Xiaohua; Liu Zhenghai; Liu Zhiming; Wan Yaping; Peng Guojian

    2012-01-01

    Safety culture, which is of great significance to establish safety objectives, characterizes level of enterprise safety production and development. Traditional safety culture evaluation models emphasis on thinking and behavior of individual and organization, and pay attention to evaluation results while ignore process. Moreover, determining evaluation indicators lacks objective evidence. A novel multidimensional safety culture evaluation model, which has scientific and completeness, is addressed by building an preliminary mapping between safety culture and SSE-CMM's (Systems Security Engineering Capability Maturity Model) process area and generic practice. The model focuses on enterprise system security engineering process evaluation and provides new ideas and scientific evidences for the study of safety culture. (authors)

  11. Meeting the next generation PWR safety requirements: The EPR Reactor

    International Nuclear Information System (INIS)

    Salhi, Othman

    2008-01-01

    The development process pursued the harmonization of technical solutions and the integration of all the lessons learned from earlier nuclear plants built by both vendors. As far as safety more specifically is concerned, the basic choice for the EPR was to adopt an evolutionary approach based on experience feedback from the reactors built by Areva, which at the time already amounted to nearly 100. This philosophy makes today's Areva EPR the natural descendant of the most advanced French N4 and German Konvoi power reactors currently in operation. EPR design choices affecting safety were motivated by a continuous quest for higher levels of safety. A two-fold approach was followed: 1. improvement of the measures aimed at further reducing the already very low probability of core melt 2. incorporation of measures aimed at further limiting the consequences of a severe accident, in the knowledge that its probability of occurrence has been considerably reduced. Through its filiations with French N4 and German Konvoi power reactors, the EPR benefits from the uninterrupted, evolutionary innovation process that has supported the development of PWRs since their introduction into the market place. This is especially true for safety where the EPR brings a unique combination of both tried and tested and innovative features that further improve the prevention of severe accidents and their mitigation

  12. Environmental and Personal Safety: No Vision Required. Practice Report

    Science.gov (United States)

    Bozeman, Laura A.

    2004-01-01

    Personal safety is an important issue for all people, regardless of their physical capabilities. For people with visual impairments (that is, those who are blind or have low vision), real concerns exist regarding their vulnerability to crime and their greater risk of attack. With a nationwide increase in crime in the United States, "Three out of…

  13. 76 FR 5494 - Pipeline Safety: Mechanical Fitting Failure Reporting Requirements

    Science.gov (United States)

    2011-02-01

    ... style'' fittings ( provides no explanation or e.g. stab, nut follower, bolted). justification for the...-RELATED CONDITION REPORTS 0 1. The authority citation for part 191 continues to read as follows: Authority... OF NATURAL AND OTHER GAS BY PIPELINE: MINIMUM FEDERAL SAFETY STANDARDS 0 3. The authority citation...

  14. Preparation, review, and approval of implementation plans for nuclear safety requirements

    International Nuclear Information System (INIS)

    1994-10-01

    This standard describes an acceptable method to prepare, review, and approve implementation plans for DOE Nuclear Safety requirements. DOE requirements are identified in DOE Rules, Orders, Notices, Immediate Action Directives, and Manuals

  15. 45 CFR 98.41 - Health and safety requirements.

    Science.gov (United States)

    2010-10-01

    ... establishment of any new or additional requirements if existing requirements comply with the requirements of the... as grandparents, great grandparents, siblings (if living in a separate residence), aunts, and uncles... care providers” does not include grandparents, great grandparents, siblings (if such providers live in...

  16. Data requirements for integrated near field models

    International Nuclear Information System (INIS)

    Wilems, R.E.; Pearson, F.J. Jr.; Faust, C.R.; Brecher, A.

    1981-01-01

    The coupled nature of the various processes in the near field require that integrated models be employed to assess long term performance of the waste package and repository. The nature of the integrated near field models being compiled under the SCEPTER program are discussed. The interfaces between these near field models and far field models are described. Finally, near field data requirements are outlined in sufficient detail to indicate overall programmatic guidance for data gathering activities

  17. Safety and licensing requirements in the Republic of South Africa

    International Nuclear Information System (INIS)

    Simpson, D.M.; Langford, E.L.

    1986-01-01

    The principles for licensing of nuclear installations in South Africa are based on the control of mortality risk to the operators of an installation and the population resident in the vicinity of the site. This paper describes the development of this safety philosophy, and the nuclear licensing process used in this country. The structure of the nuclear regulatory function is briefly described, including the respective roles of the Atomic Energy Corporation, Licencing Branch and the Council for Nuclear Safety. The development of risk criteria and quantitative release magnitude-probability criteria for radioactive material is outlined. Tasks that have to be undertaken by a potential waste disposal site licensee before a site licence is issued are described. Once the facility is commissioned periodic monitoring procedures will have to be adopted throughout the lifetime of the facility. The scope of typical monitoring activities is outlined and the ongoing analyses to be performed and the records to be kept are discussed

  18. Development of High-Level Safety Requirements for a Pyroprocessing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seok Jun; Jo, Woo Jin; You, Gil Sung; Choung, Won Myung; Lee, Ho Hee; Kim, Hyun Min; Jeon, Hong Rae; Ku, Jeong Hoe; Lee, Hyo Jik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Korea Atomic Energy Research Institute (KAERI) has been developing a pyroproceesing technology to reduce the waste volume and recycle some elements. The pyroprocessing includes several treatment processes which are related with not only radiological and physical but also chemical and electrochemical properties. Thus, it is of importance to establish safety design requirements considering all the aspects of those properties for a reliable pyroprocessing facility. In this study, high-level requirements are presented in terms of not only radiation protection, nuclear criticality, fire protection, and seismic safety but also confinement and chemical safety for the unique characteristics of a pyroprocessing facility. Several high-level safety design requirements such as radiation protection, nuclear criticality, fire protection, seismic, confinement, and chemical processing were presented for a pyroprocessing facility. The requirements must fulfill domestic and international safety technology standards for a nuclear facility. Furthermore, additional requirements should be considered for the unique electrochemical treatments in a pyroprocessing facility.

  19. Model summary report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Vahlund, Fredrik

    2006-10-15

    This document is the model summary report for the safety assessment SR-Can. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SR-Can, a number of different computer codes are used. In order to better understand how these codes are related Assessment Model Flowcharts, AMFs, have been produced within the project. From these, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A large number of different computer codes are used in the assessment of which some are commercial while others are developed especially for the current assessment project. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: It must be demonstrated that the code is suitable for its purpose; It must be demonstrated that the code has been properly used; and, It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met.

  20. Model summary report for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Vahlund, Fredrik

    2006-10-01

    This document is the model summary report for the safety assessment SR-Can. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SR-Can, a number of different computer codes are used. In order to better understand how these codes are related Assessment Model Flowcharts, AMFs, have been produced within the project. From these, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A large number of different computer codes are used in the assessment of which some are commercial while others are developed especially for the current assessment project. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: It must be demonstrated that the code is suitable for its purpose; It must be demonstrated that the code has been properly used; and, It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

  1. Patient safety climate strength: a concept that requires more attention

    Science.gov (United States)

    Ginsburg, Liane; Gilin Oore, Debra

    2016-01-01

    Background When patient safety climate (PSC) surveys are used in healthcare, reporting typically focuses on PSC level (mean or per cent positive scores). This paper explores how an additional focus on PSC strength can enhance the utility of PSC survey data. Setting and participants 442 care providers from 24 emergency departments (EDs) across Canada. Methods We use anonymised data from the Can-PSCS PSC instrument collected in 2011 as part of the Qmentum accreditation programme. We examine differences in climate strength across EDs using the Rwg(j) and intraclass correlation coefficients measures of inter-rater agreement. Results Across the six survey dimensions, median Rwg(j) was sufficiently high to support shared climate perceptions (0.64–0.83), but varied widely across the 24 ED units. We provide an illustrative example showing vastly different climate strength (Rwg(j) range=0.17–0.86) for units with an equivalent level of PSC (eg, climate mean score=3). Conclusions Most PSC survey results focus solely on climate level. To facilitate improvement in PSC, we advocate a simple, holistic safety climate profile including three metrics: climate level (using mean or per cent positive climate scores), climate strength (using the Rwg(j), or SD as a proxy) and the shape of the distribution (using histograms to see the distribution of scores within units). In PSC research, we advocate paying attention to climate strength as an important variable in its own right. Focusing on PSC level and strength can further understanding of the extent to which PSC is a key variable in the domain of patient safety. PMID:26453636

  2. Construction safety program for the National Ignition Facility Appendix A: Safety Requirements

    International Nuclear Information System (INIS)

    Cerruti, S.J.

    1997-01-01

    These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and construction contractors/subcontractors. The General Safety and Health rules shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S ampersand H A-1 that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Safety Rules

  3. Nuclear power plant's safety and risk (requirements of safety and reliability)

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1977-01-01

    Starting out from the given safety objectives as they have evolved during the past few years and from the present legal and regulatory provisions for the construction and operation of nuclear power plants, the hazards involved in regular operation, accidents and emergency situations are discussed. In compliance with the positive safety balance of nuclear power plants in the FRG, special attention is focused on the preventive safety analysis within the frame of the nuclear licensing procedure. Reference is made to the beginnings of a comprehensive hazard concept for an unbiased plant assessment. Emergency situations are discussed from the point of view of general hazard comparisons. (orig.) [de

  4. Construction safety program for the National Ignition Facility Appendix A: Safety Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Cerruti, S.J.

    1997-01-14

    These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and construction contractors/subcontractors. The General Safety and Health rules shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S & H A-1 that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Safety Rules.

  5. PROBABILISTIC MODEL FOR AIRPORT RUNWAY SAFETY AREAS

    Directory of Open Access Journals (Sweden)

    Stanislav SZABO

    2017-06-01

    Full Text Available The Laboratory of Aviation Safety and Security at CTU in Prague has recently started a project aimed at runway protection zones. The probability of exceeding by a certain distance from the runway in common incident/accident scenarios (take-off/landing overrun/veer-off, landing undershoot is being identified relative to the runway for any airport. As a result, the size and position of safety areas around runways are defined for the chosen probability. The basis for probability calculation is a probabilistic model using statistics from more than 1400 real-world cases where jet airplanes have been involved over the last few decades. Other scientific studies have contributed to understanding the issue and supported the model’s application to different conditions.

  6. A Safety Management Model for FAR 141 Approved Flight Schools

    OpenAIRE

    Mendonca, Flavio A. C.; Carney, Thomas Q

    2017-01-01

    The Safety Management Annex (Annex 19), which became applicable in November 2013, consolidates safety management provisions previously contained in six other International Civil Aviation Organization (ICAO) Annexes, and will serve as a resource for overarching state safety management responsibilities. Through Annex 19, ICAO has required that its member states develop and implement safety management systems (SMS) to improve safety. This mandate includes an approved training organization that i...

  7. Enforcement handbook: Enforcement of DOE nuclear safety requirements

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    This Handbook provides detailed guidance and procedures to implement the General Statement of DOE Enforcement Policy (Enforcement Policy or Policy). A copy of this Enforcement Policy is included for ready reference in Appendix D. The guidance provided in this Handbook is qualified, however, by the admonishment to exercise discretion in determining the proper disposition of each potential enforcement action. As discussed in subsequent chapters, the Enforcement and Investigation Staff will apply a number of factors in assessing each potential enforcement situation. Enforcement sanctions are imposed in accordance with the Enforcement Policy for the purpose of promoting public and worker health and safety in the performance of activities at DOE facilities by DOE contractors (and their subcontractors and suppliers) who are indemnified under the Price-Anderson Amendments Act. These indemnified contractors, and their suppliers and subcontractors, will be referred to in this Handbook collectively as DOE contractors. It should be remembered that the purpose of the Department`s enforcement policy is to improve nuclear safety for the workers and the public, and this goal should be the prime consideration in exercising enforcement discretion.

  8. Enforcement handbook: Enforcement of DOE nuclear safety requirements

    International Nuclear Information System (INIS)

    1995-06-01

    This Handbook provides detailed guidance and procedures to implement the General Statement of DOE Enforcement Policy (Enforcement Policy or Policy). A copy of this Enforcement Policy is included for ready reference in Appendix D. The guidance provided in this Handbook is qualified, however, by the admonishment to exercise discretion in determining the proper disposition of each potential enforcement action. As discussed in subsequent chapters, the Enforcement and Investigation Staff will apply a number of factors in assessing each potential enforcement situation. Enforcement sanctions are imposed in accordance with the Enforcement Policy for the purpose of promoting public and worker health and safety in the performance of activities at DOE facilities by DOE contractors (and their subcontractors and suppliers) who are indemnified under the Price-Anderson Amendments Act. These indemnified contractors, and their suppliers and subcontractors, will be referred to in this Handbook collectively as DOE contractors. It should be remembered that the purpose of the Department's enforcement policy is to improve nuclear safety for the workers and the public, and this goal should be the prime consideration in exercising enforcement discretion

  9. Discussion on safety culture general contract model of consultation enterprises

    International Nuclear Information System (INIS)

    Dong Huimin; Zhang Hao

    2012-01-01

    With a high safety requirement, long construction period, a large amount of investment and many influencing factors of the preparation and implementation of project schedule, local nuclear power always is built through EPC. Safety level depends on EPC. Some measures should be taken for local consultation enterprises to improve situation of safety. Some suggestion as follows: safety culture should be received enough attention; management system should be established in according with requirement of safety culture; try to encourage employee involvement; to assess it in time; safety system should be entirely compatible with enterprises system. (authors)

  10. Status of safety issues at licensed power plants: TMI Action Plan requirements, unresolved safety issues, generic safety issues, other multiplant action issues

    International Nuclear Information System (INIS)

    1992-12-01

    This report is to provide a comprehensive description of the implementation and verification status of Three Mile Island (TMI) Action Plan requirements, safety issues designated as Unresolved Safety Issues (USIs), Generic Safety Issues(GSIs), and other Multiplant Actions (MPAs) that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. An additional purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

  11. GENERAL REQUIREMENTS FOR SIMULATION MODELS IN WASTE MANAGEMENT

    International Nuclear Information System (INIS)

    Miller, Ian; Kossik, Rick; Voss, Charlie

    2003-01-01

    Most waste management activities are decided upon and carried out in a public or semi-public arena, typically involving the waste management organization, one or more regulators, and often other stakeholders and members of the public. In these environments, simulation modeling can be a powerful tool in reaching a consensus on the best path forward, but only if the models that are developed are understood and accepted by all of the parties involved. These requirements for understanding and acceptance of the models constrain the appropriate software and model development procedures that are employed. This paper discusses requirements for both simulation software and for the models that are developed using the software. Requirements for the software include transparency, accessibility, flexibility, extensibility, quality assurance, ability to do discrete and/or continuous simulation, and efficiency. Requirements for the models that are developed include traceability, transparency, credibility/validity, and quality control. The paper discusses these requirements with specific reference to the requirements for performance assessment models that are used for predicting the long-term safety of waste disposal facilities, such as the proposed Yucca Mountain repository

  12. Romania - NPP PLiM Between Regulatory Requirement / Oversight and Operator Safety / Financial Interest

    International Nuclear Information System (INIS)

    Goicea, Lucian

    2012-01-01

    Cernavoda Unit 1 PLiM started in the first third of its design life, to develop as regulatory requirements of the components of standards and programmes and to benefit by earlier implementation of the measures for achieving maximum operating life. CNCAN regulatory present approach on the utility PLiM combines the regulatory requirements on management system, ageing management provisions of periodic safety review, detailed technical requirements of ageing programmes and different techniques focusing only on safety issues. (author)

  13. Forecast model of safety economy contribution rate of China

    Institute of Scientific and Technical Information of China (English)

    LIU Li-jun; SHI Shi-liang

    2005-01-01

    It is the rational and exact computation of the safety economy contribution rate that has the far-reaching realistic meaning to the improvement of society cognition to safety and the investment to the nation safety and the national macro-safety decision-makings. The accurate function between safety inputs and outputs was obtained through a founded econometric model. Then the forecasted safety economy contribution rate is 3.01% and the forecasted ratio between safety inputs and outputs is 1:1.81 in China in 2005. And the model accords with the practice of China and the results are satisfying.

  14. Correct safety requirements during the life cycle of heating plants; Korrekta saekerhetskrav under vaermeanlaeggningars livscykel

    Energy Technology Data Exchange (ETDEWEB)

    Tegehall, Jan; Hedberg, Johan [Swedish National Testing and Research Inst., Boraas (Sweden)

    2006-10-15

    The safety of old steam boilers or hot water generators is in principle based on electromechanical components which are generally easy to understand. The use of safety-PLC is a new and flexible way to design a safe system. A programmable system offers more degrees of freedom and consequently new problems may arise. As a result, new standards which use the Safety Integrity Level (SIL) concept for the level of safety have been elaborated. The goal is to define a way of working to handle requirements on safety in control systems of heat and power plants. SIL-requirements are relatively new within the domain and there is a need for guidance to be able to follow the requirements. The target of this report is the people who work with safety questions during new construction, reconstruction, or modification of furnace plants. In the work, the Pressure Equipment Directive, 97/23/EC, as well as standards which use the SIL concept have been studied. Additionally, standards for water-tube boilers have been studied. The focus has been on the safety systems (safety functions) which are used in water-tube boilers for heat and power plants; other systems, which are parts of these boilers, have not been considered. Guidance has been given for the aforementioned standards as well as safety requirements specification and risk analysis. An old hot water generator and a relatively new steam boiler have been used as case studies. The design principles and safety functions of the furnaces have been described. During the risk analysis important hazards were identified. A method for performing a risk analysis has been described and the appropriate content of a safety requirements specification has been defined. If a heat or power plant is constructed, modified, or reconstructed, a safety life cycle shall be followed. The purpose of the safety life cycle is to plan, describe, document, perform, check, test, and validate that everything is correctly done. The components of the safety

  15. Ferrocyanide Safety Program: Data requirements for the ferrocyanide safety issue developed through the data quality objectives (DQO) process

    International Nuclear Information System (INIS)

    Buck, J.W.; Anderson, C.M.; Pulsipher, B.A.; Toth, J.J.; Turner, P.J.; Cash, R.J.; Dukelow, G.T.; Meacham, J.E.

    1993-12-01

    This document records the data quality objectives (DQO) process applied to the Ferrocyanide Waste Tank Safety Issue at the Hanford Site by the Pacific Northwest Laboratory and Westinghouse Hanford Company. Specifically, the major recommendations and findings from this Ferrocyanide DQO process are presented so that decision makers can determine the type, quantity, and quality of data required for addressing tank safety issues. The decision logic diagrams and error tolerance equations also are provided. Finally, the document includes the DQO sample-size formulas for determining specific tank sampling requirements

  16. Information Management system of the safety regulatory requirements and guidance for the Korea next generation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Y. C. [LG-EDS Systems, Seoul (Korea, Republic of); Lee, J. H.; Lee, H. C.; Lee, J. S. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2000-05-01

    In order to achieve the safety of the Korea Next Generation Reactors (KNGR), the Korea Institute of Nuclear Safety has carried out the Safety and Regulatory Requirements and Guidance (SRRG) development program from 1992 such as establishment of the SRRG hierarchy, development of technical requirements and guidance, and consideration of new licensing system. The SRRG hierarchy for the KNGR was consisted of five tiers; Safety Objectives, Safety Principles, General Safety Criteria, Specific Safety Requirements and Safety Regulatory Guides. The developed SRRG have been compared the criteria in 10CFR and Reg. Guide in the U.S.A and the IAEA documents for assuring internationally acceptable level of the SRRG. To improve the efficiency and accuracy of SRRG development, the construction of database system was required in the course of development. Therefore, the Information Management System of SRRG for the KNGR has been developed which enables developers to quickly and accurately seek and systematically manage whole contexts of the SRRG, reference requirements, and current atomic energy regulation rules. Moreover, through homepage whose URL is 'http://kngr.kins.re.kr', the concerned persons and public can acquire the information related with SRRG and KNGR project, and post his/her thought to the opinion forum in the homepage.

  17. Information Management system of the safety regulatory requirements and guidance for the Korea next generation reactors

    International Nuclear Information System (INIS)

    Yun, Y. C.; Lee, J. H.; Lee, H. C.; Lee, J. S.

    2000-01-01

    In order to achieve the safety of the Korea Next Generation Reactors (KNGR), the Korea Institute of Nuclear Safety has carried out the Safety and Regulatory Requirements and Guidance (SRRG) development program from 1992 such as establishment of the SRRG hierarchy, development of technical requirements and guidance, and consideration of new licensing system. The SRRG hierarchy for the KNGR was consisted of five tiers; Safety Objectives, Safety Principles, General Safety Criteria, Specific Safety Requirements and Safety Regulatory Guides. The developed SRRG have been compared the criteria in 10CFR and Reg. Guide in the U.S.A and the IAEA documents for assuring internationally acceptable level of the SRRG. To improve the efficiency and accuracy of SRRG development, the construction of database system was required in the course of development. Therefore, the Information Management System of SRRG for the KNGR has been developed which enables developers to quickly and accurately seek and systematically manage whole contexts of the SRRG, reference requirements, and current atomic energy regulation rules. Moreover, through homepage whose URL is 'http://kngr.kins.re.kr', the concerned persons and public can acquire the information related with SRRG and KNGR project, and post his/her thought to the opinion forum in the homepage

  18. Heat transfer and friction correlations required to describe steam--water behavior in nuclear safety studies

    International Nuclear Information System (INIS)

    Solbrig, C.W.; McFadden, J.H.; Lyczkowski, R.W.; Hughes, E.D.

    1975-01-01

    The description of two-phase flow is important in nuclear safety studies. Recent two-phase flow descriptions are based upon unequal phase velocities and unequal phase temperatures (UVUT) theories with interphase interaction terms. These theories are more mechanistic than homogeneous theories and require more and different types of correlations than homogeneous theories. The UVUT theories require correlations (or models) which describe wall and interphase mass transfer, friction, momentum transfer, and heat transfer for all flow regimes and heat transfer regimes. A set of correlations is presented in this paper which can be used with UVUT theories. These correlations cover the complete range of parameters needed and in all cases are expected to yield reasonable numbers. (U.S.)

  19. Safety requirements expected to the prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    2014-11-01

    In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

  20. Model summary report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    Vahlund, Fredrik; Zetterstroem Evins, Lena; Lindgren, Maria

    2010-12-01

    This document is the model summary report for the safety assessment SR-Site. In the report, the quality assurance (QA) measures conducted for assessment codes are presented together with the chosen QA methodology. In the safety assessment project SR-Site, a large number of numerical models are used to analyse the system and to show compliance. In order to better understand how the different models interact and how information are transferred between the different models Assessment Model Flowcharts, AMFs, are used. From these, different modelling tasks can be identify and the computer codes used. As a large number of computer codes are used in the assessment the complexity of these differs to a large extent, some of the codes are commercial while others are developed especially for the assessment at hand. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined for all codes: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. - It must be described how data are transferred between the different computational tasks. Although the requirements are identical for all codes in the assessment, the measures used to show that the requirements are fulfilled will be different for different types of codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented together with a discussion on how the requirements are met

  1. Model summary report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Vahlund, Fredrik; Zetterstroem Evins, Lena (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)); Lindgren, Maria (Kemakta Konsult AB, Stockholm (Sweden))

    2010-12-15

    This document is the model summary report for the safety assessment SR-Site. In the report, the quality assurance (QA) measures conducted for assessment codes are presented together with the chosen QA methodology. In the safety assessment project SR-Site, a large number of numerical models are used to analyse the system and to show compliance. In order to better understand how the different models interact and how information are transferred between the different models Assessment Model Flowcharts, AMFs, are used. From these, different modelling tasks can be identify and the computer codes used. As a large number of computer codes are used in the assessment the complexity of these differs to a large extent, some of the codes are commercial while others are developed especially for the assessment at hand. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined for all codes: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. - It must be described how data are transferred between the different computational tasks. Although the requirements are identical for all codes in the assessment, the measures used to show that the requirements are fulfilled will be different for different types of codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented together with a discussion on how the requirements are met

  2. Risk and safety requirements for diagnostic and therapeutic procedures in allergology

    DEFF Research Database (Denmark)

    Kowalski, Marek L; Ansotegui, Ignacio; Aberer, Werner

    2016-01-01

    One of the major concerns in the practice of allergy is related to the safety of procedures for the diagnosis and treatment of allergic disease. Management (diagnosis and treatment) of hypersensitivity disorders involves often intentional exposure to potentially allergenic substances (during skin...... attempted to present general requirements necessary to assure the safety of these procedures. Following review of available literature a group of allergy experts within the World Allergy Organization (WAO), representing various continents and areas of allergy expertise, presents this report on risk...... associated with diagnostic and therapeutic procedures in allergology and proposes a consensus on safety requirements for performing procedures in allergy offices. Optimal safety measures including appropriate location, type and required time of supervision, availability of safety equipment, access...

  3. Current trends in codal requirements for safety in operation of nuclear power plants

    International Nuclear Information System (INIS)

    Srivasista, K.; Shah, Y.K.; Gupta, S.K.

    2006-01-01

    The Code of practice on safety in nuclear power plant operation states the requirements to be met during operation of a nuclear power plant for assuring safety. Among various stages of authorization, regulatory body issues authorization for operation of a nuclear power plant, monitors and enforces regulatory requirements. The responsible organization shall have overall responsibility and the plant management shall have the primary responsibility for ensuring safe and efficient operation of its nuclear power plants. A set of codal requirements covering technical and administrative aspects are mandatory for the plant management to implement to ensure that the nuclear power plant is operated in accordance with the design intent. Requirements on operating procedures and instructions establish operation and maintenance, inspection and testing of the plant in a planned and systematic way. The requirements on emergency preparedness programme establish with a reasonable assurance that, in the event of an emergency situation, appropriate measures can be taken to mitigate the consequences. Commissioning requirements verify performance criteria during commissioning to ensure that the design intent and QA requirements are met. Several modifications in systems important to safety required during operation of a nuclear power plant are regulated. However new operational codal requirements arising out of periodic safety review, operational experience feedback, life management, probabilistic safety assessment, physical security, safety convention and obligations and decommissioning are not covered in the present code of practice for safety in nuclear power plant operation. Codal provisions on 'Review by operating organization on aspects of design having implications on operability' are also required to be addressed. The merits in developing such a methodology include acceptance of the design by operating organization, ensuring maintainability, proper layout etc. in the new designs

  4. The safety of food products requires X-ray testing

    International Nuclear Information System (INIS)

    Lardiere, C.

    2017-01-01

    Food safety through standards and regulations imposes food products to be tested for the presence of alien elements. So far metal detectors have been used to detect metal parts, now they have been progressively replaced with X-ray scanners that allow the detection of a lot more contaminants. The improvement of algorithms for image processing combined with the availability of ever more powerful PC have led to the routine use of X-ray testing on industrial processes. Technological progress has made X-ray testing more efficient: while previously a power of 500 W was necessary to cross a 10 cm thickness of water, now only 100 W is necessary. The main advantage of X-ray testing is to be able to test food even packed in metal containers and to detect if the container is deformed. Another advantages is to be able to detect a lot of elements like pieces of glass, small stones or bits of bones. The minimal size to be detected is 0.5 mm for stainless steels and 2 mm for glass or bones. Usually metal detectors are set at the end of the production line just before packaging but in some cases they can be included in the line in order to protect the next machine that intervenes to process the food. (A.C.)

  5. Meteorological aspects of the reactor safety study requiring further study

    International Nuclear Information System (INIS)

    Slinn, W.G.N.

    1981-01-01

    Simple and approximate methods are used in a search for meteorological features that dominate estimates of reactor-accident consequences, and that require more accurate descriptions if consequence estimates are to be more realistic. By considering variations in the source term, it is seen that accidents involving containment-vessel failure dominate both the mean and variance of the distribution of consequences, although this conclusion is subject to uncertainties about plume rise. Research is recommended on the behavior of horizontal, sonic jets, with heat transfer to the ground, and especially during stable atmospheric conditions. Diffusion with fumigation and lofting require further study; use of K-theory and National Weather Service data should be vigorously pursued. Conditional upon an accident occurring, precipitation scavenging appears to dominate the variance of the consequences

  6. Usage of information safety requirements in improving tube bending process

    Science.gov (United States)

    Livshitz, I. I.; Kunakov, E.; Lontsikh, P. A.

    2018-05-01

    This article is devoted to an improvement of the technological process's analysis with the information security requirements implementation. The aim of this research is the competition increase analysis in aircraft industry enterprises due to the information technology implementation by the example of the tube bending technological process. The article analyzes tube bending kinds and current technique. In addition, a potential risks analysis in a tube bending technological process is carried out in terms of information security.

  7. 49 CFR 1106.3 - Actions for which Safety Integration Plan is required.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 8 2010-10-01 2010-10-01 false Actions for which Safety Integration Plan is required. 1106.3 Section 1106.3 Transportation Other Regulations Relating to Transportation (Continued... TRANSPORTATION BOARD CONSIDERATION OF SAFETY INTEGRATION PLANS IN CASES INVOLVING RAILROAD CONSOLIDATIONS...

  8. 45 CFR 1356.30 - Safety requirements for foster care and adoptive home providers.

    Science.gov (United States)

    2010-10-01

    ... licensing file for that foster or adoptive family must contain documentation which verifies that safety... 45 Public Welfare 4 2010-10-01 2010-10-01 false Safety requirements for foster care and adoptive... ON CHILDREN, YOUTH AND FAMILIES, FOSTER CARE MAINTENANCE PAYMENTS, ADOPTION ASSISTANCE, AND CHILD AND...

  9. A Requirements Analysis Model Based on QFD

    Institute of Scientific and Technical Information of China (English)

    TANG Zhi-wei; Nelson K.H.Tang

    2004-01-01

    The enterprise resource planning (ERP) system has emerged to offer an integrated IT solution and more and more enterprises are increasing by adopting this system and regarding it as an important innovation. However, there is already evidence of high failure risks in ERP project implementation, one major reason is poor analysis of the requirements for system implementation. In this paper, the importance of requirements analysis for ERP project implementation is highlighted, and a requirements analysis model by applying quality function deployment (QFD) is presented, which will support to conduct requirements analysis for ERP project.

  10. 78 FR 2797 - Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and Electric Vehicles

    Science.gov (United States)

    2013-01-14

    ... Sound Requirements for Hybrid and Electric Vehicles; Draft Environmental Assessment for Rulemaking To Establish Minimum Sound Requirements for Hybrid and Electric Vehicles; Proposed Rules #0;#0;Federal Register...-0148] RIN 2127-AK93 Federal Motor Vehicle Safety Standards; Minimum Sound Requirements for Hybrid and...

  11. Model-based human reliability analysis: prospects and requirements

    International Nuclear Information System (INIS)

    Mosleh, A.; Chang, Y.H.

    2004-01-01

    Major limitations of the conventional methods for human reliability analysis (HRA), particularly those developed for operator response analysis in probabilistic safety assessments (PSA) of nuclear power plants, are summarized as a motivation for the need and a basis for developing requirements for the next generation HRA methods. It is argued that a model-based approach that provides explicit cognitive causal links between operator behaviors and directly or indirectly measurable causal factors should be at the core of the advanced methods. An example of such causal model is briefly reviewed, where due to the model complexity and input requirements can only be currently implemented in a dynamic PSA environment. The computer simulation code developed for this purpose is also described briefly, together with current limitations in the models, data, and the computer implementation

  12. Defence-in-depth and development of safety requirements for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Carnino, A.; Gasparini, M.

    2002-01-01

    The paper addresses a general approach for the preparation of the design safety requirements using the IAEA Safety Objectives and the strategy of defence-in-depth. It proposes a general method (top-down approach) to prepare safety requirements for a given kind of reactor using the IAEA requirements for nuclear power plants as a starting point through a critical interpretation and application of the strategy of defence-in-depth. The IAEA has recently developed a general methodology for screening the defence-in-depth of nuclear power plants starting from the fundamental safety objectives as proposed in the IAEA Safety Fundamentals. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor. Currently the IAEA is preparing the technical basis for the development of safety requirements for Modular High Temperature Gas Reactors, with the aim of showing the viability of the method. A draft TECDOC has been prepared and circulated among several experts for comments. This paper is largely based on the content of the draft TECDOC. (authors)

  13. Lithuanian requirements for ageing management of systems and components important to safety of nuclear power plant

    International Nuclear Information System (INIS)

    Ramanauskiene, A.

    2000-01-01

    In this paper the Lithuanian requirements for ageing management of systems and components important to safety of Ignalina nuclear power plant (two RBMK-1500 water-cooled graphite moderated channel-type power reactors) are presented

  14. Modeling and Testing Legacy Data Consistency Requirements

    DEFF Research Database (Denmark)

    Nytun, J. P.; Jensen, Christian Søndergaard

    2003-01-01

    An increasing number of data sources are available on the Internet, many of which offer semantically overlapping data, but based on different schemas, or models. While it is often of interest to integrate such data sources, the lack of consistency among them makes this integration difficult....... This paper addresses the need for new techniques that enable the modeling and consistency checking for legacy data sources. Specifically, the paper contributes to the development of a framework that enables consistency testing of data coming from different types of data sources. The vehicle is UML and its...... accompanying XMI. The paper presents techniques for modeling consistency requirements using OCL and other UML modeling elements: it studies how models that describe the required consistencies among instances of legacy models can be designed in standard UML tools that support XMI. The paper also considers...

  15. An architecture model for communication of safety in public transportation

    NARCIS (Netherlands)

    Rajabalinejad, Mohammad; Horváth, Imre; Pernot, Jean-Paul; Rusák, Zoltan

    2016-01-01

    Safety in transportation is under the influence of the rising complexity, increasing demands for capacity and decreasing cost. Furthermore, the interdisciplinary environment of operation and altered safety regulations invite for a centralized (integrated) modelling/ communication approach. This

  16. “The Safety of Tourist” - Required Community Standard

    Directory of Open Access Journals (Sweden)

    Olguta Ludmila Benescu

    2010-06-01

    be forgotten which would attract tourists.The Romanian State, in order to have had an attractive and profitable tourism it must turn into a leading provider of safety.

  17. Early Engagement of Safety and Mission Assurance Expertise Using Systems Engineering Tools: A Risk-Based Approach to Early Identification of Safety and Assurance Requirements

    Science.gov (United States)

    Darpel, Scott; Beckman, Sean

    2016-01-01

    Decades of systems engineering practice have demonstrated that the earlier the identification of requirements occurs, the lower the chance that costly redesigns will needed later in the project life cycle. A better understanding of all requirements can also improve the likelihood of a design's success. Significant effort has been put into developing tools and practices that facilitate requirements determination, including those that are part of the model-based systems engineering (MBSE) paradigm. These efforts have yielded improvements in requirements definition, but have thus far focused on a design's performance needs. The identification of safety & mission assurance (S&MA) related requirements, in comparison, can occur after preliminary designs are already established, yielding forced redesigns. Engaging S&MA expertise at an earlier stage, facilitated by the use of MBSE tools, and focused on actual project risk, can yield the same type of design life cycle improvements that have been realized in technical and performance requirements.

  18. Dimensioning of emergency condensers in accordance with safety requirements

    Energy Technology Data Exchange (ETDEWEB)

    Palavecino, C [SIEMENS, Energieerzeugung, Offenbach (Germany)

    1996-12-01

    The emergency condensers are heat exchangers consisting of a parallel arrangement of horizontal U-tubes between two common heads. The tope header is connected via piping to the reactor vessel steam space, while the lower header is connected to the reactor vessel below the reactor vessel water level. The heat exchangers are located in a pool filled with cold water. The emergency condensers and the reactor vessel thus form a system of communicating pipes. At normal reactor water level, the emergency condensers are flooded with cold, non-flowing water. No heat transfer takes place in this condition. If there is a drop in the reactor water level, the heat exchanging surfaces are gradually uncovered and the incoming steam condenses on the cold surfaces. The cold condensate in returned to the reactor vessel. In this way, heat is removed from the reactor vessel and water simultaneously supplied to the reactor vessel. This means that the emergency condensers function as a heat removal system while at the same time serving as HP and LP coolant injection systems. The emergency condensers operate with the highest possible degree of passivity imaginable, namely through a drop in the reactor vessel water level alone, requiring neither control systems nor power supply. The design of the emergency condensers must meet the requirements dictated by the thermal and the hydraulic conditions. Taking into consideration a redundancy degree of N + 2, a specific thermal rating of 63 MW per emergency condenser results for a reactor with an output of 2778 MW. The total performance of the emergency condenser system in thus 252 MW, or 9.1% of reactor output. The probability of failure of the emergency condenser of Siemens SWR 1000 is approximately 10{sup -4} per demand, while that of the older emergency condenser designs is approximately 2 to 3 x 10{sup -3} per demand. (author). 7 figs, 2 tabs.

  19. Status of safety issues at licensed power plants: TMI Action Plan requirements; unresolved safety issues; generic safety issues; other multiplant action issues

    International Nuclear Information System (INIS)

    1993-12-01

    As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. This third annual NUREG report, Supplement 3, presents updated information as of September 30, 1993. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

  20. Status of safety issues at licensed power plants: TMI Action Plan requirements, unresolved safety issues, generic safety issues, other multiplant action issues. Supplement 4

    International Nuclear Information System (INIS)

    1994-12-01

    As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, the NRC established a program for publishing an annual report on the status of licensee implementation and NRC verification of safety issues in major NRC requirements areas. This information was initially compiled and reported in three NUREG-series volumes. Volume 1, published in March 1991, addressed the status of Three Mile Island (TMI) Action Plan Requirements. Volume 2, published in May 1991, addressed the status of unresolved safety issues (USIs). Volume 3, published in June 1991, addressed the implementation and verification status of generic safety issues (GSIs). The first annual supplement, which combined these volumes into a single report and presented updated information as of September 30, 1991, was published in December 1991. The second annual supplement, which provided updated information as of September 30, 1992, was published in December 1992. Supplement 2 also provided the status of licensee implementation and NRC verification of other multiplant action (MPA) issues not related to TMI Action Plan requirements, USIs, or GSIs. Supplement 3 gives status as of September 30, 1993. This annual report, Supplement 4, presents updated information as of September 30, 1994. This report gives a comprehensive description of the implementation and verification status of TMI Action Plan requirements, safety issues designated as USIs, GSIs, and other MPAs that have been resolved and involve implementation of an action or actions by licensees. This report makes the information available to other interested parties, including the public. Additionally, this report serves as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues until requirements are approved for imposition at licensed plants or until the NRC issues a request for action by licensees

  1. User requirements in the area of safety of innovative nuclear reactors and fuel cycle installations

    International Nuclear Information System (INIS)

    Kuczera, B.; Juhn, P.E.; Fukuda, K.; )

    2002-01-01

    Full text: Against the background of already existing IAEA and INSAC publications in the area of safety, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a set of user requirements for the safety of future nuclear installations has been established. Five top-level requirements are expected to apply to any type of innovative design. They should foster an increased level of safety that is transparent to and fully accepted by the general public. The approach to future reactor safety includes two complementary strategies: increased emphasis on inherent safety characteristics and enhancement of defense in depth. As compared to existing plants, the effectiveness of preventing measures should be highly enhanced, resulting in fewer mitigation measures. The targets and possible approaches of each of the five levels of defense developed for innovative reactor designs are outlined in the paper

  2. Eurosafe 2006 radioactive waste management: long term safety requirements and societal expectations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    The EUROSAFE Forum is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety, to share experiences, exchange technical and scientific opinions, and conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum 2006 focuses on 'Radioactive Waste Management: Long Term Safety Requirements and Societal Expectations' from the point of view of the authorities, TSOs and industry and presents the latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe. A high level of nuclear safety is a priority for Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining increasing recognition. This

  3. Eurosafe 2006 radioactive waste management: long term safety requirements and societal expectations

    International Nuclear Information System (INIS)

    2006-01-01

    The EUROSAFE Forum is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety, to share experiences, exchange technical and scientific opinions, and conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum 2006 focuses on 'Radioactive Waste Management: Long Term Safety Requirements and Societal Expectations' from the point of view of the authorities, TSOs and industry and presents the latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe. A high level of nuclear safety is a priority for Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining increasing recognition. This

  4. Eurosafe 2006 radioactive waste management: long term safety requirements and societal expectations

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    The EUROSAFE Forum is part of the EUROSAFE approach, which consists of two further elements: the EUROSAFE Tribune and the EUROSAFE web site. The general aim of EUROSAFE is to contribute to fostering the convergence of technical nuclear safety practices in a broad European context. This is done by providing technical safety and research organisations, safety authorities, power utilities, the rest of the industry and non-governmental organisations mainly from the European Union and East-European countries, and international organisations with a platform for the presentation of recent analyses and R and D in the field of nuclear safety, to share experiences, exchange technical and scientific opinions, and conduct debates on key issues in the fields of nuclear safety and radiation protection. The EUROSAFE Forum 2006 focuses on 'Radioactive Waste Management: Long Term Safety Requirements and Societal Expectations' from the point of view of the authorities, TSOs and industry and presents the latest work in nuclear installation safety and research, waste management, radiation safety as well as nuclear material and nuclear facilities security carried out by GRS, IRSN, AVN and their partners in the European Union, Switzerland and Eastern Europe. A high level of nuclear safety is a priority for Europe. The technical safety organisations play an important role in contributing to that objective through appropriate approaches to major safety issues as part of their assessments and research activities. The challenges to nuclear safety are international. Changes in underlying technologies such as instrumentation and control, the impact of electricity market deregulation, demands for improved safety and safety management, the ageing of nuclear facilities, waste management, maintaining and improving scientific and technical knowledge, and the need for greater transparency - these are all issues where the value of an international approach is gaining increasing recognition

  5. Statement on safety requirements concerning the long-term operation of the Muehleberg nuclear power station

    International Nuclear Information System (INIS)

    2012-12-01

    This report published by the Swiss Federal Nuclear Safety Inspectorate ENSI investigates the safety requirements with respect to the long-term operation of the Muehleberg nuclear power station in Switzerland. Relevant international requirements and Swiss legal stipulations concerning the long-term operation of the power station are stated. The management of aging processes is looked at. The regular verification of the integrity of various plant components such as containments, piping, steam generation system, etc. is looked at in detail. The state-of-the-art concerning deterministic accident analyses and refitting technology are discussed, as are automated safety systems. The applicable laws, decrees and guidelines are listed in appendices

  6. Meeting up-to-date safety requirements in the Russian NPP projects

    International Nuclear Information System (INIS)

    Tepkyan, G. O.; Yashkin, A. V.

    2014-01-01

    Safety features in Russian NPP designs are implemented by the combination of active and passive safety systems • Russian NPP designs are in compliance with up-to-date international and European safety requirements and refer to Generation III+ • Russian state-of-the-art designs have already implemented some design solutions, which take into account “post-Fukushima” requirements. Russian NPP design principles have been approved during the European discussions in spring 2012, including the IAEA extraordinary session addressed to Fukushima NPP accident

  7. Guide for reviewing safety analysis reports for packaging: Review of quality assurance requirements

    International Nuclear Information System (INIS)

    Moon, D.W.

    1988-10-01

    This review section describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging which are important to safety. The design effort, operation's plans, and quality assurance requirements should be integrated to achieve a system in which the independent QA program is not overly stringent and the application of QA requirements is commensurate with safety significance. The reviewer must verify that the applicant's QA section in the SARP contains package-specific QA information required by DOE Orders and federal regulations that demonstrate compliance. 8 refs

  8. Excellence and safety in surgery require excellent and safe tutoring.

    Science.gov (United States)

    Ferrarese, Alessia; Buccelli, Claudio; Addeo, Giuseppe; Capasso, Emanuele; Conti, Adelaide; Amato, Maurizio; Compagna, Rita; Niola, Massimo; Martino, Valter

    2016-01-01

    The surgical education in Italy has always been a very important issue. The aim of this article is to bring together the feedback of the definitions of the various components of the learning scheme and to evaluate the importance of the legal point. In March 2016 we performed a literature review. We have also examinated the internet pages of the Italian Department of Education, Health and Medical Order. In Italy the tutor had an unclear role from a legal point of view. He is the person who must be able to perform a specific procedure with expert technical and who must know how to stop the student if this is about to perform a dangerous maneuver. In Italy the ability to work for the trainee is limited in all reality, it depends on several factors including the increase of numbers of medical-legal disputes, the timing, the commitment it requires mentoring and a lack of mentors. In surgery, the problem is greater because of the increasingly of medico-legal implications that we are after surgical procedure. It would be necessary to define a role of the tutor in a regular protocol and a proper assessment of his performance.

  9. Safety sans Frontières: An International Safety Culture Model.

    Science.gov (United States)

    Reader, Tom W; Noort, Mark C; Shorrock, Steven; Kirwan, Barry

    2015-05-01

    The management of safety culture in international and culturally diverse organizations is a concern for many high-risk industries. Yet, research has primarily developed models of safety culture within Western countries, and there is a need to extend investigations of safety culture to global environments. We examined (i) whether safety culture can be reliably measured within a single industry operating across different cultural environments, and (ii) if there is an association between safety culture and national culture. The psychometric properties of a safety culture model developed for the air traffic management (ATM) industry were examined in 17 European countries from four culturally distinct regions of Europe (North, East, South, West). Participants were ATM operational staff (n = 5,176) and management staff (n = 1,230). Through employing multigroup confirmatory factor analysis, good psychometric properties of the model were established. This demonstrates, for the first time, that when safety culture models are tailored to a specific industry, they can operate consistently across national boundaries and occupational groups. Additionally, safety culture scores at both regional and national levels were associated with country-level data on Hofstede's five national culture dimensions (collectivism, power distance, uncertainty avoidance, masculinity, and long-term orientation). MANOVAs indicated safety culture to be most positive in Northern Europe, less so in Western and Eastern Europe, and least positive in Southern Europe. This indicates that national cultural traits may influence the development of organizational safety culture, with significant implications for safety culture theory and practice. © 2015 Society for Risk Analysis.

  10. A fuzzy model for exploiting customer requirements

    Directory of Open Access Journals (Sweden)

    Zahra Javadirad

    2016-09-01

    Full Text Available Nowadays, Quality function deployment (QFD is one of the total quality management tools, where customers’ views and requirements are perceived and using various techniques improves the production requirements and operations. The QFD department, after identification and analysis of the competitors, takes customers’ feedbacks to meet the customers’ demands for the products compared with the competitors. In this study, a comprehensive model for assessing the importance of the customer requirements in the products or services for an organization is proposed. The proposed study uses linguistic variables, as a more comprehensive approach, to increase the precision of the expression evaluations. The importance of these requirements specifies the strengths and weaknesses of the organization in meeting the requirements relative to competitors. The results of these experiments show that the proposed method performs better than the other methods.

  11. Introduction of the Amendment of IAEA Safety Requirements Reflected Lessons Learned from Fukushima Nuclear Accident

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang-Kyu; Ahn, Hyung-Joon; Kim, Sun-Hae; Cheong, Jae-Hak [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The following five Safety Requirements publications were amended: Governmental, Legal and Regulatory Framework for Safety (GSR Part 1, 2010), Site Evaluation for Nuclear Installations (NS-R-3, 2003), Safety of Nuclear Power Plants: Design (SSR-2/1, 2012), Safety of Nuclear Power Plants: Commissioning and Operation (SSR-2/2, 2011), and Safety Assessment for Facilities and Activities (GSR Part 4, 2009). Figure 1 shows IAEA Safety Standards Categories Major amendments of five Safety Requirements publications were introduced and analyzed in this study. The five IAEA safety requirements publications which are GSR Part 1 and 4, NS-R-3 and SSR-2/1 and 2, were amended to reflect the lesson learned from the Fukushima accident and other operating experiences. Specially, 36 provisions were modified and the new 29 provision with 1 requirement (No. 67: Emergency response facilities on the site) of the SSR-2/1 were established. Since the Fukushima accident happened, a new word, design extension conditions (DECs) which cover substantially the beyond design basis accidents (BDBA), including severe accident conditions, was created and more elaborated by the world nuclear experts. Design extension conditions could include conditions in events without significant fuel degradation and conditions with core melting. Figure 2 shows the range of the DECs. The amendment of the five IAEA safety requirements publications are focused at the prevention of initiating events, which would lead to the DECs, and mitigation of the consequences of DECs by the enhanced defense in depth principle. The following examples of the IAEA requirements to prevent the initiating events are: margins for withstanding external events; margins for avoiding cliff edge effects; safety assessment for multiple facilities or activities at a single site; safety assessment in cases where resources at a facility are shared; consideration of the potential occurrence of events in combination; establishing levels of hazard

  12. Modeling issues associated with production reactor safety assessment

    International Nuclear Information System (INIS)

    Stack, D.W.; Thomas, W.R.

    1990-01-01

    This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs

  13. Plutonium air transportable package Model PAT-1. Safety analysis report

    International Nuclear Information System (INIS)

    1978-02-01

    The document is a Safety Analysis Report for the Plutonium Air Transportable Package, Model PAT-1, which was developed by Sandia Laboratories under contract to the Nuclear Regulatory Commission (NRC). The document describes the engineering tests and evaluations that the NRC staff used as a basis to determine that the package design meets the requirements specified in the NRC ''Qualification Criteria to Certify a Package for Air Transport of Plutonium'' (NUREG-0360). By virtue of its ability to meet the NRC Qualification Criteria, the package design is capable of safely withstanding severe aircraft accidents. The document also includes engineering drawings and specifications for the package. 92 figs, 29 tables

  14. Lessons learned - development of the tritium facilities 5480.23 safety analysis report and technical safety requirements

    International Nuclear Information System (INIS)

    Cappucci, A.J. Jr.; Bowman, M.E.; Goff, L.

    1997-01-01

    A review was performed which identified open-quotes Lessons Learnedclose quotes from the development of the 5480.23 Tritium Safety Analysis Report (SAR) and the Technical Safety Requirements (TSR) for the Tritium Facilities (TF). The open-quotes Lessons Learnedclose quotes were based on an evaluation of the use of the SRS procedures, processes, and work practices which contributed to the success or lack thereof. This review also identified recommendations and suggestions for improving the development of SARs and TSRs at SRS. The 5480.23 SAR describes the site for the TF, the various process systems in the process buildings, a complete hazards and accident analysis of the most significant hazards affecting the nearby offsite population, and the selection of safety systems, structures, and components to protect both the public and site workers. It also provides descriptions of important programs and processes which add defense in depth to public and worker protection

  15. Integrating Safety and Mission Assurance into Systems Engineering Modeling Practices

    Science.gov (United States)

    Beckman, Sean; Darpel, Scott

    2015-01-01

    During the early development of products, flight, or experimental hardware, emphasis is often given to the identification of technical requirements, utilizing such tools as use case and activity diagrams. Designers and project teams focus on understanding physical and performance demands and challenges. It is typically only later, during the evaluation of preliminary designs that a first pass, if performed, is made to determine the process, safety, and mission quality assurance requirements. Evaluation early in the life cycle, though, can yield requirements that force a fundamental change in design. This paper discusses an alternate paradigm for using the concepts of use case or activity diagrams to identify safety hazard and mission quality assurance risks and concerns using the same systems engineering modeling tools being used to identify technical requirements. It contains two examples of how this process might be used in the development of a space flight experiment, and the design of a Human Powered Pizza Delivery Vehicle, along with the potential benefits to decrease development time, and provide stronger budget estimates.

  16. 12 CFR Appendix A to Part 1720 - Policy Guidance; Minimum Safety and Soundness Requirements

    Science.gov (United States)

    2010-01-01

    ..., DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT SAFETY AND SOUNDNESS SAFETY AND SOUNDNESS Pt. 1720, App. A... effectively and to model the effect of differing interest rate scenarios on the Enterprise's financial... are implemented effectively, and that the Enterprise's organization structure and assignment of...

  17. Requirements and international co-operation in nuclear safety for evolutionary light water reactors

    International Nuclear Information System (INIS)

    Carnino, A.

    1999-01-01

    The principles of safety are now well known and implemented world-wide, leading to a situation of harmonisation in accordance with the Convention on Nuclear Safety. Future reactors are expected not only to meet current requirements but to go beyond the safety level presently accepted. To this end, technical safety requirements, as defined by the IAEA document Safety Fundamentals, need be duly considered in the design, the risks to workers and population must be decreased, a stable, transparent and objective regulatory process, including an international harmonisation with respect to licensing of new reactors, must be developed, and the issue of public acceptance must be addressed. Well-performing existing installations are seen as a prerequisite for an improved public acceptability; there should be no major accidents, the results from safety performance indicators must be unquestionable, and compliance with internationally harmonised criteria is essential. Economical competitiveness is another factor that influences the acceptability; the costs for constructing the plant, for its operation and maintenance, for the fuel cycle, and for the final decommissioning are of paramount importance. Plant simplification, longer fuel cycles, life extension are appealing options, but safety will have first priority. The IAEA can play an important role in this field, by providing peer reviews by teams of international experts and assistance to Member States on the use of its safety standards. (author)

  18. Impact of New Radiation Safety Standards on Licensing Requirements of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Strohal, P.; Subasic, D.; Valcic, I.

    1996-01-01

    As the outcomes of the newly introduced safety philosophies, new and more strict safety design requirements for nuclear installation are expected to be introduced. New in-depth defence measures should be incorporated into the design and operation procedure for a nuclear installation, to compensate for potential failures in protection or safety measures. The new requirements will also apply to licensing of NPP's operation as well as to licensing of nuclear sites, especially for radioactive waste disposal sites. This paper intends to give an overview of possible impacts of new internationally agreed basic safety standards with respect to NPP and related technologies. Recently issued new basic safety standards for radiation protection are introducing some new safety principles which may have essential impact on future licensing requirements regarding nuclear power plants and radioactive waste installations. These new standards recognize exposures under normal conditions ('practices') and intervention conditions. The term interventions describes the human activities that seek to reduce the existing radiation exposure or existing likelihood of incurring exposure which is not part of a controlled practice. The other new development in safety standards is the introduction of so called potential exposure based on the experience gained from a number of radiation accidents. This exposure is not expected to be delivered with certainty but it may result from an accident at a source or owing to an event or sequence of events of a probabilistic nature, including equipment failures and operating errors. (author)

  19. Standard model for the safety analysis report of nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    1980-02-01

    This norm establishes the Standard Model for the Safety Analysis Report of Nuclear Fuel Reprocessing Plants, comprehending the presentation format, the detailing level of the minimum information required by the CNEN for evaluation the requests of Construction License or Operation Authorization, in accordance with the legislation in force. This regulation applies to the following basic reports: Preliminary Safety Analysis Report - PSAR, integrating part of the requirement of Construction License; and Final Safety Analysis Report (FSAR) which is the integrating part of the requirement for Operation Authorization

  20. Model-based safety architecture framework for complex systems

    NARCIS (Netherlands)

    Schuitemaker, Katja; Rajabali Nejad, Mohammadreza; Braakhuis, J.G.; Podofillini, Luca; Sudret, Bruno; Stojadinovic, Bozidar; Zio, Enrico; Kröger, Wolfgang

    2015-01-01

    The shift to transparency and rising need of the general public for safety, together with the increasing complexity and interdisciplinarity of modern safety-critical Systems of Systems (SoS) have resulted in a Model-Based Safety Architecture Framework (MBSAF) for capturing and sharing architectural

  1. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF...

  2. Occupational safety and health in the Universities: fulfilling the fundamental requirement of OSHA and AELA

    International Nuclear Information System (INIS)

    Ismail Bahari

    2000-01-01

    This paper discusses the result of a survey among the universities to looks at whether such basic similarities in requirements by both Acts actually help in fulfilling and integrating the fundamental requirement of OSHA, Malaysian Occupational Safety and Health Act and AELA, Malaysian Atomic Energy Licensing Act especially through self-regulation

  3. 78 FR 47015 - Software Requirement Specifications for Digital Computer Software Used in Safety Systems of...

    Science.gov (United States)

    2013-08-02

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0195] Software Requirement Specifications for Digital Computer Software Used in Safety Systems of Nuclear Power Plants AGENCY: Nuclear Regulatory Commission... issuing a revised regulatory guide (RG), revision 1 of RG 1.172, ``Software Requirement Specifications for...

  4. Safety requirements to the operation of hydropower plants; Sicherheit beim Betrieb von Wasserkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Lux, Reinhard [Berufsgenossenschaft Energie Textil Elektro Medienerzeugnisse (BG ETEM), Koeln (Germany)

    2011-07-01

    Employers have to take into account various safety and health requirements relating to the design, construction, operation and maintenance of hydropower plants. Especially the diversity of the hydropower plant components requires the consideration of different safety and health aspects. In 2011 the ''Fachausschuss Elektrotechnik'' (expert committee electro-technics) of the institution for statutory accident insurance and prevention presented a new ''BG-Information'' dealing with ''Safe methods operating hydropower plants''. The following article gives an introduction into the conception and the essential requirements of this new BG-Information. (orig.)

  5. Models of Human Information Requirements: "When Reasonable Aiding Systems Disagree"

    Science.gov (United States)

    Corker, Kevin; Pisanich, Gregory; Shafto, Michael (Technical Monitor)

    1994-01-01

    Aircraft flight management and Air Traffic Control (ATC) automation are under development to maximize the economy of flight and to increase the capacity of the terminal area airspace while maintaining levels of flight safety equal to or better than current system performance. These goals are being realized by the introduction of flight management automation aiding and operations support systems on the flight deck and by new developments of ATC aiding systems that seek to optimize scheduling of aircraft while potentially reducing required separation and accounting for weather and wake vortex turbulence. Aiding systems on both the flight deck and the ground operate through algorithmic functions on models of the aircraft and of the airspace. These models may differ from each other as a result of variations in their models of the immediate environment. The resultant flight operations or ATC commands may differ in their response requirements (e.g. different preferred descent speeds or descent initiation points). The human operators in the system must then interact with the automation to reconcile differences and resolve conflicts. We have developed a model of human performance including cognitive functions (decision-making, rule-based reasoning, procedural interruption recovery and forgetting) that supports analysis of the information requirements for resolution of flight aiding and ATC conflicts. The model represents multiple individuals in the flight crew and in ATC. The model is supported in simulation on a Silicon Graphics' workstation using Allegro Lisp. Design guidelines for aviation automation aiding systems have been developed using the model's specification of information and team procedural requirements. Empirical data on flight deck operations from full-mission flight simulation are provided to support the model's predictions. The paper describes the model, its development and implementation, the simulation test of the model predictions, and the empirical

  6. Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Requirements (Chinese Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication, jointly sponsored by the FAO, IAEA, ICAO, ILO, IMO, INTERPOL, OECD/NEA, PAHO, CTBTO, UNEP, OCHA, WHO and WMO, is the new edition establishing the requirements for preparedness and response for a nuclear or radiological emergency which takes into account the latest experience and developments in the area. It supersedes the previous edition of the Safety Requirements for emergency preparedness and response, Safety Standards Series No. GS-R-2, which was published in 2002. This publication establishes the requirements for ensuring an adequate level of preparedness and response for a nuclear or radiological emergency, irrespective of its cause. These Safety Requirements are intended to be used by governments, emergency response organizations, other authorities at the local, regional and national levels, operating organizations and the regulatory body as well as by relevant international organizations at the international level.

  7. Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication, jointly sponsored by the FAO, IAEA, ICAO, ILO, IMO, INTERPOL, OECD/NEA, PAHO, CTBTO, UNEP, OCHA, WHO and WMO, is the new edition establishing the requirements for preparedness and response for a nuclear or radiological emergency which takes into account the latest experience and developments in the area. It supersedes the previous edition of the Safety Requirements for emergency preparedness and response, Safety Standards Series No. GS-R-2, which was published in 2002. This publication establishes the requirements for ensuring an adequate level of preparedness and response for a nuclear or radiological emergency, irrespective of its cause. These Safety Requirements are intended to be used by governments, emergency response organizations, other authorities at the local, regional and national levels, operating organizations and the regulatory body as well as by relevant international organizations at the international level.

  8. Preparedness and Response for a Nuclear or Radiological Emergency. General Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication, jointly sponsored by the FAO, IAEA, ICAO, ILO, IMO, INTERPOL, OECD/NEA, PAHO, CTBTO, UNEP, OCHA, WHO and WMO, is the new edition establishing the requirements for preparedness and response for a nuclear or radiological emergency which takes into account the latest experience and developments in the area. It supersedes the previous edition of the Safety Requirements for emergency preparedness and response, Safety Standards Series No. GS-R-2, which was published in 2002. This publication establishes the requirements for ensuring an adequate level of preparedness and response for a nuclear or radiological emergency, irrespective of its cause. These Safety Requirements are intended to be used by governments, emergency response organizations, other authorities at the local, regional and national levels, operating organizations and the regulatory body as well as by relevant international organizations at the international level.

  9. Specification of safety requirements for waste packages with respect to practicable quality control measures

    International Nuclear Information System (INIS)

    Gruendler, D.; Wurtinger, W.

    1987-01-01

    Waste packages for disposal in a repository in the Federal Republic of Germany have to meet safety requirements derived from site specific safety analyses. The examination of the waste packages with regard to compliance with these requirements is the main objective of quality control measures. With respect to quality control the requirements have to be specified in a way that practicable control measures can be applied. This is dealt with for the quality control of the activity inventory and the quality control of the waste form. The paper discusses the determination of the activity of hard-to-measure radionuclides and the specification of safety related requirements for the waste form and the packaging using typical examples

  10. Criticality Safety Evaluation of Standard Criticality Safety Requirements #1-520 g Operations in PF-4

    Energy Technology Data Exchange (ETDEWEB)

    Yamanaka, Alan Joseph Jr. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-13

    Guidance has been requested from the Nuclear Criticality Safety Division (NCSD) regarding processes that involve 520 grams of fissionable material or less. This Level-3 evaluation was conducted and documented in accordance with NCS-AP-004 (Ref. 1), formerly NCS-GUIDE-01. This evaluation is being written as a generic evaluation for all operations that will be able to operate using a 520-gram mass limit. Implementation for specific operations will be performed using a Level 1 CSED, which will confirm and document that this CSED can be used for the specific operation as discussed in NCS-MEMO-17-007 (Ref. 2). This Level 3 CSED updates and supersedes the analysis performed in NCS-TECH-14-014 (Ref. 3).

  11. A review of models relevant to road safety.

    Science.gov (United States)

    Hughes, B P; Newstead, S; Anund, A; Shu, C C; Falkmer, T

    2015-01-01

    It is estimated that more than 1.2 million people die worldwide as a result of road traffic crashes and some 50 million are injured per annum. At present some Western countries' road safety strategies and countermeasures claim to have developed into 'Safe Systems' models to address the effects of road related crashes. Well-constructed models encourage effective strategies to improve road safety. This review aimed to identify and summarise concise descriptions, or 'models' of safety. The review covers information from a wide variety of fields and contexts including transport, occupational safety, food industry, education, construction and health. The information from 2620 candidate references were selected and summarised in 121 examples of different types of model and contents. The language of safety models and systems was found to be inconsistent. Each model provided additional information regarding style, purpose, complexity and diversity. In total, seven types of models were identified. The categorisation of models was done on a high level with a variation of details in each group and without a complete, simple and rational description. The models identified in this review are likely to be adaptable to road safety and some of them have previously been used. None of systems theory, safety management systems, the risk management approach, or safety culture was commonly or thoroughly applied to road safety. It is concluded that these approaches have the potential to reduce road trauma. Copyright © 2014 Elsevier Ltd. All rights reserved.

  12. Simulation modeling on the growth of firm's safety management capability

    Institute of Scientific and Technical Information of China (English)

    LIU Tie-zhong; LI Zhi-xiang

    2008-01-01

    Aiming to the deficiency of safety management measure, established simulation model about firm's safety management capability(FSMC) based on organizational learning theory. The system dynamics(SD) method was used, in which level and rate system, variable equation and system structure flow diagram was concluded. Simulation model was verified from two aspects: first, model's sensitivity to variable was tested from the gross of safety investment and the proportion of safety investment; second, variables dependency was checked up from the correlative variable of FSMC and organizational learning. The feasibility of simulation model is verified though these processes.

  13. Development of NPP Safety Requirements into Kenya's Grid Codes

    Energy Technology Data Exchange (ETDEWEB)

    Ndirangu, Nguni James; Koo, Chang Choong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand.

  14. Software Safety Analysis of Digital Protection System Requirements Using a Qualitative Formal Method

    International Nuclear Information System (INIS)

    Lee, Jang-Soo; Kwon, Kee-Choon; Cha, Sung-Deok

    2004-01-01

    The safety analysis of requirements is a key problem area in the development of software for the digital protection systems of a nuclear power plant. When specifying requirements for software of the digital protection systems and conducting safety analysis, engineers find that requirements are often known only in qualitative terms and that existing fault-tree analysis techniques provide little guidance on formulating and evaluating potential failure modes. A framework for the requirements engineering process is proposed that consists of a qualitative method for requirements specification, called the qualitative formal method (QFM), and a safety analysis method for the requirements based on causality information, called the causal requirements safety analysis (CRSA). CRSA is a technique that qualitatively evaluates causal relationships between software faults and physical hazards. This technique, extending the qualitative formal method process and utilizing information captured in the state trajectory, provides specific guidelines on how to identify failure modes and the relationship among them. The QFM and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital protection system example

  15. Development of NPP Safety Requirements into Kenya's Grid Codes

    International Nuclear Information System (INIS)

    Ndirangu, Nguni James; Koo, Chang Choong

    2015-01-01

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand

  16. Flightdeck Automation Problems (FLAP) Model for Safety Technology Portfolio Assessment

    Science.gov (United States)

    Ancel, Ersin; Shih, Ann T.

    2014-01-01

    NASA's Aviation Safety Program (AvSP) develops and advances methodologies and technologies to improve air transportation safety. The Safety Analysis and Integration Team (SAIT) conducts a safety technology portfolio assessment (PA) to analyze the program content, to examine the benefits and risks of products with respect to program goals, and to support programmatic decision making. The PA process includes systematic identification of current and future safety risks as well as tracking several quantitative and qualitative metrics to ensure the program goals are addressing prominent safety risks accurately and effectively. One of the metrics within the PA process involves using quantitative aviation safety models to gauge the impact of the safety products. This paper demonstrates the role of aviation safety modeling by providing model outputs and evaluating a sample of portfolio elements using the Flightdeck Automation Problems (FLAP) model. The model enables not only ranking of the quantitative relative risk reduction impact of all portfolio elements, but also highlighting the areas with high potential impact via sensitivity and gap analyses in support of the program office. Although the model outputs are preliminary and products are notional, the process shown in this paper is essential to a comprehensive PA of NASA's safety products in the current program and future programs/projects.

  17. Workshop on Program for Elimination of Requirements Marginal to Safety: Proceedings

    International Nuclear Information System (INIS)

    Dey, M.

    1993-09-01

    These are the proceedings of the Public Workshop on the US Nuclear Regulatory Commission's Program for Elimination of Requirements Marginal to Safety. The workshop was held at the Holiday Inn, Bethesda, on April 27 and 28, 1993. The purpose of the workshop was to provide an opportunity for public and industry input to the program. The workshop addressed the institutionalization of the program to review regulations with the purpose of eliminating those that are marginal. The objective is to avoid the dilution of safety efforts. One session was devoted to discussion of the framework for a performance-based regulatory approach. In addition, panelists and attendees discussed scope, schedules and status of specific regulatory items: containment leakage testing requirements, fire protection requirements, requirements for environmental qualification of electrical equipment, requests for information under 10CFR50.54(f), requirements for combustible gas control systems, and quality assurance requirements

  18. Workshop on Program for Elimination of Requirements Marginal to Safety: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Dey, M. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution; Arsenault, F.; Patterson, M.; Gaal, M. [SCIENTECH, Inc., Rockville, MD (United States)

    1993-09-01

    These are the proceedings of the Public Workshop on the US Nuclear Regulatory Commission`s Program for Elimination of Requirements Marginal to Safety. The workshop was held at the Holiday Inn, Bethesda, on April 27 and 28, 1993. The purpose of the workshop was to provide an opportunity for public and industry input to the program. The workshop addressed the institutionalization of the program to review regulations with the purpose of eliminating those that are marginal. The objective is to avoid the dilution of safety efforts. One session was devoted to discussion of the framework for a performance-based regulatory approach. In addition, panelists and attendees discussed scope, schedules and status of specific regulatory items: containment leakage testing requirements, fire protection requirements, requirements for environmental qualification of electrical equipment, requests for information under 10CFR50.54(f), requirements for combustible gas control systems, and quality assurance requirements.

  19. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  20. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  1. Design of plant safety model in plant enterprise engineering environment

    International Nuclear Information System (INIS)

    Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-01-01

    Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured

  2. A Safety and Health Guide for Vocational Educators. Incorporating Requirements of the Occupational Safety and Health Act of 1970, Relevant Pennsylvania Requirements with Particular Emphasis for Those Concerned with Cooperative Education and Work Study Programs. Volume 15. Number 1.

    Science.gov (United States)

    Wahl, Ray

    Intended as a guide for vocational educators to incorporate the requirements of the Occupational Safety and Health Act (1970) and the requirements of various Pennsylvania safety and health regulations with their cooperative vocational programs, the first chapter of this document presents the legal implications of these safety and health…

  3. What Isn’t Working and New Requirements. The Need to Harmonize Safety and Security Requirements

    International Nuclear Information System (INIS)

    Flory, D.

    2016-01-01

    This paper sets out the key issues for consideration at the transport conference. It will introduce each of the aspects of the framework for safe, secure and sustainable transport, building on the description of the existing situation presented in Session 1A. It will discuss purpose of the IAEA framework, and examine the scientific basis, the IAEA recommendations and requirements, the UN interface, the use of conventions, national implementation, industry compliance, communication and information, response and restoration. It will also look at the activities and related requirements outside of transport which could influence the transport frameworks either in a positive or negative manner. (author)

  4. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition); Seguridad de las centrales nucleares: Diseno. Requisitos de seguridad especificos

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  5. Microbial food safety - modeling and applications

    Science.gov (United States)

    Microbial food safety is a key issue for the food processing industry, and enhancing food safety is everyone’s responsibility from food producers to consumers. Financial losses to the economy due to foodborne illness are in the billions of dollars, annually. Foodborne illness can be caused by patho...

  6. Probabilistic safety analysis of DC power supply requirements for nuclear power plants. Technical report

    International Nuclear Information System (INIS)

    Baranowsky, P.W.; Kolaczkowski, A.M.; Fedele, M.A.

    1981-04-01

    A probabilistic safety assessment was performed as part of the Nuclear Regulatory Commission generic safety task A-30, Adequacy of Safety Related DC Power Supplies. Event and fault tree analysis techniques were used to determine the relative contribution of DC power related accident sequences to the total core damage probability due to shutdown cooling failures. It was found that a potentially large DC power contribution could be substantially reduced by augmenting the minimum design and operational requirements. Recommendations included (1) requiring DC power divisional independence, (2) improved test, maintenance, and surveillance, and (3) requiring core cooling capability be maintained following the loss of one DC power bus and a single failure in another system

  7. Validation of a functional model for integration of safety into process system design

    DEFF Research Database (Denmark)

    Wu, J.; Lind, M.; Zhang, X.

    2015-01-01

    with the process system functionalities as required for the intended safety applications. To provide the scientific rigor and facilitate the acceptance of qualitative modelling, this contribution focuses on developing a scientifically based validation method for functional models. The Multilevel Flow Modeling (MFM...

  8. Safety Case Development as an Information Modelling Problem

    Science.gov (United States)

    Lewis, Robert

    This paper considers the benefits from applying information modelling as the basis for creating an electronically-based safety case. It highlights the current difficulties of developing and managing large document-based safety cases for complex systems such as those found in Air Traffic Control systems. After a review of current tools and related literature on this subject, the paper proceeds to examine the many relationships between entities that can exist within a large safety case. The paper considers the benefits to both safety case writers and readers from the future development of an ideal safety case tool that is able to exploit these information models. The paper also introduces the idea that the safety case has formal relationships between entities that directly support the safety case argument using a methodology such as GSN, and informal relationships that provide links to direct and backing evidence and to supporting information.

  9. Validation of CFD models for hydrogen safety application

    International Nuclear Information System (INIS)

    Nikolaeva, Anna; Skibin, Alexander; Krutikov, Alexey; Golibrodo, Luka; Volkov, Vasiliy; Nechaev, Artem; Nadinskiy, Yuriy

    2015-01-01

    Most accidents involving hydrogen begin with its leakage and spreading in the air and spontaneous detonation, which is accompanied by fire or deflagration of hydrogen mixture with heat and /or shocks, which may cause harm to life and equipment. Outflow of hydrogen in a confined volume and its propagation in the volume is the worst option because of the impact of the insularity on the process of detonation. According to the safety requirements for handling hydrogen specialized systems (ventilation, sprinklers, burners etc.) are required for maintaining the hydrogen concentration less than the critical value, to eliminate the possibility of detonation and flame propagation. In this study, a simulation of helium propagation in a confined space with different methods of injection and ventilation of helium is presented, which is used as a safe replacement of hydrogen in experimental studies. Five experiments were simulated in the range from laminar to developed turbulent with different Froude numbers, which determine the regime of the helium outflow in the air. The processes of stratification and erosion of helium stratified layer were investigated. The study includes some results of OECD/NEA-PSI PANDA benchmark and some results of Gamelan project. An analysis of applicability of various turbulence models, which are used to close the system of equations of momentum transport, implemented in the commercial codes STAR CD, STAR CCM+, ANSYS CFX, was conducted for different mesh types (polyhedral and hexahedral). A comparison of computational studies results with experimental data showed a good agreement. In particular, for transition and turbulent regimes the error of the numerical results lies in the range from 5 to 15% for all turbulence models considered. This indicates applicability of the methods considered for some hydrogen safety problems. However, it should be noted that more validation research should be made to use CFD in Hydrogen safety applications with a wide

  10. Safety requirements laid down in the Atomic Energy Law and in the Law on Immission Control

    International Nuclear Information System (INIS)

    Hansmann, K.

    1981-01-01

    The paper deals with safety requirements relating to installations, laid down in the Atomic Energy Law and in the Law on Immission Control. Actually it is a matter of how the safety requirements of sect. 7 of the Atomic Energy Law can be compared with those laid down in the sections 5 and 6 of the Federal Act for the Protection Against Nuisances. In the process, three comparative levels are examined: 1. The normative conditions concerning the licencability of hazardous installations, 2. those demands that go way beyond that in order to reduce residual risks, and 3. the licensing authorities' scope of discretion. (orig./HP) [de

  11. Nuclear safety requirements for upgrading the National Repository for Radioactive Wastes-Baita Bihor

    International Nuclear Information System (INIS)

    Vladescu, Gabriela; Necula, Daniela

    2000-01-01

    The upgrading project of National Repository for Radioactive Wastes-Baita Bihor is based on the integrated concept of nuclear safety. Its ingredients are the following: A. The principles of nuclear safety regarding the management of radioactive wastes and radioprotection; B. Safety objectives for final disposal of low- and intermediate-level radioactive wastes; C. Safety criteria for final disposal of low- and intermediate-level radioactive wastes; D. Assessment of safety criteria fulfillment for final disposal of low- and intermediate-level radioactive wastes. Concerning the nuclear safety in radioactive waste management the following issues are considered: population health protection, preventing transfrontier contamination, future generation radiation protection, national legislation, control of radioactive waste production, interplay between radioactive waste production and management, radioactive waste repository safety. The safety criteria of final disposal of low- and intermediate-level radioactive wastes are discussed by taking into account the geological and hydrogeological configuration, the physico-chemical and geochemical characteristics, the tectonics and seismicity conditions, extreme climatic potential events at the mine location. Concerning the requirements upon the repository, the following aspects are analyzed: the impact on environment, the safety system reliability, the criticality control, the filling composition to prevent radioactive leakage, the repository final sealing, the surveillance. Concerning the radioactive waste, specific criteria taken into account are the radionuclide content, the chemical composition and stability, waste material endurance to heat and radiation. The waste packaging criteria discussed are the mechanical endurance, materials toughness and types as related to deterioration caused by handling, transportation, storing or accidents. Fulfillment of safety criteria is assessed by scenarios analyses and analyses of

  12. Safety and environmental requirements and design targets for TIBER-II

    International Nuclear Information System (INIS)

    Piet, S.J.

    1987-09-01

    A consistent set of safety and environmental requirements and design targets was proposed and adopted for the TIBER-II (Tokamak Ignition/Burn Experimental Reactor) design effort. TIBER-II is the most recent US version of a fusion experimental test reactor (ETR). These safety and environmental design targets were one contribution of the Fusion Safety Program in the TIBER-II design effort. The other contribution, safety analyses, is documented in the TIBER-II design report. The TIBER-II approach, described here, concentrated on logical development of, first, a complete and consistent set of safety and environmental requirements that are likely appropriate for an ETR, and, second, an initial set of design targets to guide TIBER-II. Because of limited time in the TIBER-II design effort, the iterative process only included one iteration - one set of targets and one design. Future ETR design efforts should therefore build on these design targets and the associated safety analyses. 29 refs., 5 figs., 3 tabs

  13. Models and methods for hot spot safety work

    DEFF Research Database (Denmark)

    Vistisen, Dorte

    2002-01-01

    Despite the fact that millions DKK each year are spent on improving roadsafety in Denmark, funds for traffic safety are limited. It is therefore vital to spend the resources as effectively as possible. This thesis is concerned with the area of traffic safety denoted "hot spot safety work", which...... is the task of improving road safety through alterations of the geometrical and environmental characteristics of the existing road network. The presently applied models and methods in hot spot safety work on the Danish road network were developed about two decades ago, when data was more limited and software...... and statistical methods less developed. The purpose of this thesis is to contribute to improving "State of the art" in Denmark. Basis for the systematic hot spot safety work are the models describing the variation in accident counts on the road network. In the thesis hierarchical models disaggregated on time...

  14. Modelling the effects of road traffic safety measures.

    Science.gov (United States)

    Lu, Meng

    2006-05-01

    A model is presented for assessing the effects of traffic safety measures, based on a breakdown of the process in underlying components of traffic safety (risk and consequence), and five (speed and conflict related) variables that influence these components, and are influenced by traffic safety measures. The relationships between measures, variables and components are modelled as coefficients. The focus is on probabilities rather than historical statistics, although in practice statistics may be needed to find values for the coefficients. The model may in general contribute to improve insight in the mechanisms between traffic safety measures and their safety effects. More specifically it allows comparative analysis of different types of measures by defining an effectiveness index, based on the coefficients. This index can be used to estimate absolute effects of advanced driver assistance systems (ADAS) related measures from absolute effects of substitutional (in terms of safety effects) infrastructure measures.

  15. General Approaches and Requirements on Safety and Security of Radioactive Materials Transport in Russian Federation

    International Nuclear Information System (INIS)

    Ershov, V.N.; Buchel'nikov, A.E.; Komarov, S.V.

    2016-01-01

    Development and implementation of safety and security requirements for transport of radioactive materials in the Russian Federation are addressed. At the outset it is worth noting that the transport safety requirements implemented are in full accordance with the IAEA's ''Regulations for the Safe Transport of Radioactive Material (2009 Edition)''. However, with respect to security requirements for radioactive material transport in some cases the Russian Federation requirements for nuclear material are more stringent compared to IAEA recommendations. The fundamental principles of safety and security of RM managements, recommended by IAEA documents (publications No. SF-1 and GOV/41/2001) are compared. Its correlation and differences concerning transport matters, the current level and the possibility of harmonization are analysed. In addition a reflection of the general approaches and concrete transport requirements is being evaluated. Problems of compliance assessment, including administrative and state control problems for safety and security provided at internal and international shipments are considered and compared. (author)

  16. Safety requirements in the design of research reactors: A Canadian perspective

    International Nuclear Information System (INIS)

    Lee, A.G.; Langman, V.J.

    2000-01-01

    In Canada, the formal development of safety requirements for the design of research reactors in general began under an inter-organizational Small Reactor Criteria Committee. This committee developed safety and licensing criteria for use by several small reactor projects in their licensing discussions with the Atomic Energy Control Board. The small reactor projects or facilities represented included the MAPLE-X10 reactor, the proposed SES-10 heating reactor and its prototype, the SDR reactor at the Whiteshell Laboratories, the Korea Multipurpose Research Reactor (a.k.a., HANARO) in Korea, the SCORE project, and the McMaster University Nuclear Reactor. The top level set of criteria which form a safety philosophy and serve as a framework for more detailed developments was presented at an IAEA Conference in 1989. AECL continued this work to develop safety principles and design criteria for new small reactors. The first major application of this work has been to the design, safety analysis and licensing of the MAPLE 1 and 2 reactors for the MDS Nordion Medical Isotope Reactor Project. This paper provides an overview of the safety principles and design criteria. Examples of an implementation of these safety principles and design criteria are drawn from the work to design the MAPLE 1 and 2 reactors. (author)

  17. International review on safety requirements for the prototype fast breeder reactor “Monju”

    International Nuclear Information System (INIS)

    2016-01-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  18. 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)

  19. International review on safety requirements for the prototype fast breeder reactor “Monju” (Translated document)

    International Nuclear Information System (INIS)

    2016-02-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  20. Preliminary Assessment of Operational Hazards and Safety Requirements for Airborne Trajectory Management (ABTM) Roadmap Applications

    Science.gov (United States)

    Cotton, William B.; Hilb, Robert; Koczo, Stefan, Jr.; Wing, David J.

    2016-01-01

    A set of five developmental steps building from the NASA TASAR (Traffic Aware Strategic Aircrew Requests) concept are described, each providing incrementally more efficiency and capacity benefits to airspace system users and service providers, culminating in a Full Airborne Trajectory Management capability. For each of these steps, the incremental Operational Hazards and Safety Requirements are identified for later use in future formal safety assessments intended to lead to certification and operational approval of the equipment and the associated procedures. Two established safety assessment methodologies that are compliant with the FAA's Safety Management System were used leading to Failure Effects Classifications (FEC) for each of the steps. The most likely FEC for the first three steps, Basic TASAR, Digital TASAR, and 4D TASAR, is "No effect". For step four, Strategic Airborne Trajectory Management, the likely FEC is "Minor". For Full Airborne Trajectory Management (Step 5), the most likely FEC is "Major".

  1. Multilevel model of safety climate for furniture industries.

    Science.gov (United States)

    Rodrigues, Matilde A; Arezes, Pedro M; Leão, Celina P

    2015-01-01

    Furniture companies can analyze their safety status using quantitative measures. However, the data needed are not always available and the number of accidents is under-reported. Safety climate scales may be an alternative. However, there are no validated Portuguese scales that account for the specific attributes of the furniture sector. The current study aims to develop and validate an instrument that uses a multilevel structure to measure the safety climate of the Portuguese furniture industry. The Safety Climate in Wood Industries (SCWI) model was developed and applied to the safety climate analysis using three different scales: organizational, group and individual. A multilevel exploratory factor analysis was performed to analyze the factorial structure. The studied companies' safety conditions were also analyzed. Different factorial structures were found between and within levels. In general, the results show the presence of a group-level safety climate. The scores of safety climates are directly and positively related to companies' safety conditions; the organizational scale is the one that best reflects the actual safety conditions. The SCWI instrument allows for the identification of different safety climates in groups that comprise the same furniture company and it seems to reflect those groups' safety conditions. The study also demonstrates the need for a multilevel analysis of the studied instrument.

  2. Job Demands-Control-Support model and employee safety performance.

    Science.gov (United States)

    Turner, Nick; Stride, Chris B; Carter, Angela J; McCaughey, Deirdre; Carroll, Anthony E

    2012-03-01

    The aim of this study was to explore whether work characteristics (job demands, job control, social support) comprising Karasek and Theorell's (1990) Job Demands-Control-Support framework predict employee safety performance (safety compliance and safety participation; Neal and Griffin, 2006). We used cross-sectional data of self-reported work characteristics and employee safety performance from 280 healthcare staff (doctors, nurses, and administrative staff) from Emergency Departments of seven hospitals in the United Kingdom. We analyzed these data using a structural equation model that simultaneously regressed safety compliance and safety participation on the main effects of each of the aforementioned work characteristics, their two-way interactions, and the three-way interaction among them, while controlling for demographic, occupational, and organizational characteristics. Social support was positively related to safety compliance, and both job control and the two-way interaction between job control and social support were positively related to safety participation. How work design is related to employee safety performance remains an important area for research and provides insight into how organizations can improve workplace safety. The current findings emphasize the importance of the co-worker in promoting both safety compliance and safety participation. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

  3. Identifying environmental safety and health requirements for an Environmental Restoration Management Contractor

    International Nuclear Information System (INIS)

    Beckman, W.H.; Cossel, S.C.; Alhadeff, N.; Lindamood, S.B.; Beers, J.A.

    1993-10-01

    The purpose of the Standards/Requirements Identification Program, developed partially in response to the Defense Nuclear Facilities Safety Board Recommendation 90-2, was to identify applicable requirements that established the Environmental Restoration Management Contractor's (ERMC) responsibilities and authorities under the Environmental Restoration Management Contract, determine the adequacy of these requirements, ascertain a baseline level of compliance with them, and implement a maintenance program that would keep the program current as requirements or compliance levels change. The resultant Standards/Requirements Identification Documents (S/RIDs) consolidate the applicable requirements. These documents govern the development of procedures and manuals to ensure compliance with the requirements. Twenty-four such documents, corresponding with each functional area identified at the site, are to be issued. These requirements are included in the contractor's management plan

  4. Application of model systems in nanobiotechnology safety

    International Nuclear Information System (INIS)

    Khalilov, R.I.; Aliev, E.Sh.; Khudaverdieva, S.R.

    2010-11-01

    Full text : Last 10-15 years the human civilization, as a result of fast development of biotechnology, cases of new and known illnesses and increase of danger of bioterrorism, collides with new biological dangers. Now, all necessity of actions for biology for prevention of possible dangers admits. Nanobiotechnological researches and offers on application of the scientific results reached in this area prevail of all others. And thus, in many cases or it is at all left outside of attention possible harmful effects of application in an expert of nanoparticles, or it is limited to researches on subcellular level. Adequate results can be received only in case of carrying out of such researches on organism level. Greater prospects in this area have the model systems consisting the culture of unicellular green seaweed, on which now we have been studying the ionizing radiation influence effects. It speaks that on behalf of such cultures we have simultaneously cellular, organism and population levels of the structural organization. Some optimal laboratory methods of maintenance and propagating of this unicellular green seaweed have already been developed. The way offered was a studying at cellular-organism level of the structural organization of effects of action on vital systems of nanoparticles (especially what are offered for application in pharmaceutics) with use of culture of unicellular green seaweed Chlamydomonas reinhardti. Genes of many enzymes of this eucariotic seaweed are established, and also its perspective value in biological synthesis of hydrogen is shown. Studying of negative effects of action of nanoparticles in an example of the object, many molecular features of which are investigated, will allow to establish borders of safety of all biosystems.

  5. Preparedness and response for a nuclear or radiological emergency. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Requirements publication establishes the requirements for an adequate level of preparedness and response for a nuclear or radiological emergency in any State. Their implementation is intended to minimize the consequences for people, property and the environment of any nuclear or radiological emergency. The fulfilment of these requirements will also contribute to the harmonization of arrangements in the event of a transnational emergency. These requirements are intended to be applied by authorities at the national level by means of adopting legislation, establishing regulations and assigning responsibilities. The requirements apply to all those practices and sources that have the potential for causing radiation exposure or environmental radioactive contamination warranting an emergency intervention and that are: (a) Used in a State that chooses to adopt the requirements or that requests any of the sponsoring organizations to provide for the application of the requirements. (B) Used by States with the assistance of the FAO, IAEA, ILO, PAHO, OCHA or WHO in compliance with applicable national rules and regulations. (C) Used by the IAEA or which involve the use of materials, services, equipment, facilities and non-published information made available by the IAEA or at its request or under its control or supervision. Or (d) Used under any bilateral or multilateral arrangement whereby the parties request the IAEA to provide for the application of the requirements. The requirements also apply to the off-site jurisdictions that may need to make an emergency intervention in a State that adopts the requirements. The types of practices and sources covered by these requirements include: fixed and mobile nuclear reactors. Facilities for the mining and processing of radioactive ores. Facilities for fuel reprocessing and other fuel cycle facilities. Facilities for the management of radioactive waste. The transport of radioactive material. Sources of radiation used in

  6. Economics of the specification 6M safety re-evaluation and regulatory requirements

    International Nuclear Information System (INIS)

    Hopper, C.M.

    1985-01-01

    The objective of this work was to examine the potential economic impact of the DOT Specification 6M criticality safety re-evaluation and regulatory requirements. The examination was based upon comparative analyses of current authorized fissile material load limits for the 6M, current Federal regulations (and interpretations) limiting the contents of Type B fissile material packages, limiting aggregates of fissile material packages, and recent proposed fissile material mass limits derived from specialized criticality safety analyses of the 6M package. The work examines influences on cost in transportation, handling, and storage of fissile materials. Depending upon facility throughput requirements (and assumed incremental costs of fissile material packaging, storage, and transport), operating, facility storage capacity, and transportation costs can be reduced significantly. As an example of the pricing algorithm application based upon reasonable cost influences, the magnitude of the first year cost reductions could extend beyond four times the cost of the packaging nuclear criticality safety re-evaluation. 1 tab

  7. Assessment of the impact of dipped guideways on urban rail transit systems: Ventilation and safety requirements

    Science.gov (United States)

    1982-01-01

    The ventilation and fire safety requirements for subway tunnels with dipped profiles between stations as compared to subway tunnels with level profiles were evaluated. This evaluation is based upon computer simulations of a train fire emergency condition. Each of the tunnel configurations evaluated was developed from characteristics that are representative of modern transit systems. The results of the study indicate that: (1) The level tunnel system required about 10% more station cooling than dipped tunnel systems in order to meet design requirements; and (2) The emergency ventilation requirements are greater with dipped tunnel systems than with level tunnel systems.

  8. Regulations for the safe transport of radioactive material, 2005 edition. Safety requirements

    International Nuclear Information System (INIS)

    2005-01-01

    This publication includes amendments to the 1996 Edition (As Amended 2003) arising from the second cycle of the biennial review and revision process, as agreed by the Transport Safety Standards Committee (TRANSSC) at its ninth meeting in March 2004, as endorsed by the Commission on Safety Standards at its meeting in June 2004 and as approved by the IAEA Board of Governors in November 2004. Although this publication is identified as a new edition, there are no changes that affect the administrative and approval requirements in Section VIII. The fields covered are General Provisions (radiation protection; emergency response; quality assurance; compliance assurance; non-compliance; special arrangement and training); Activity Limits and Materials Restrictions, Requirement and Controls for Transport , Requirements for Radioactive Materials and for Packagings and Packages, Test Procedures, Approval and Administrative Requirements

  9. Modelling safety of multistate systems with ageing components

    Energy Technology Data Exchange (ETDEWEB)

    Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna [Gdynia Maritime University, Department of Mathematics ul. Morska 81-87, Gdynia 81-225 Poland (Poland)

    2016-06-08

    An innovative approach to safety analysis of multistate ageing systems is presented. Basic notions of the ageing multistate systems safety analysis are introduced. The system components and the system multistate safety functions are defined. The mean values and variances of the multistate systems lifetimes in the safety state subsets and the mean values of their lifetimes in the particular safety states are defined. The multi-state system risk function and the moment of exceeding by the system the critical safety state are introduced. Applications of the proposed multistate system safety models to the evaluation and prediction of the safty characteristics of the consecutive “m out of n: F” is presented as well.

  10. Modelling safety of multistate systems with ageing components

    International Nuclear Information System (INIS)

    Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna

    2016-01-01

    An innovative approach to safety analysis of multistate ageing systems is presented. Basic notions of the ageing multistate systems safety analysis are introduced. The system components and the system multistate safety functions are defined. The mean values and variances of the multistate systems lifetimes in the safety state subsets and the mean values of their lifetimes in the particular safety states are defined. The multi-state system risk function and the moment of exceeding by the system the critical safety state are introduced. Applications of the proposed multistate system safety models to the evaluation and prediction of the safty characteristics of the consecutive “m out of n: F” is presented as well.

  11. Models and data requirements for human reliability analysis

    International Nuclear Information System (INIS)

    1989-03-01

    It has been widely recognised for many years that the safety of the nuclear power generation depends heavily on the human factors related to plant operation. This has been confirmed by the accidents at Three Mile Island and Chernobyl. Both these cases revealed how human actions can defeat engineered safeguards and the need for special operator training to cover the possibility of unexpected plant conditions. The importance of the human factor also stands out in the analysis of abnormal events and insights from probabilistic safety assessments (PSA's), which reveal a large proportion of cases having their origin in faulty operator performance. A consultants' meeting, organized jointly by the International Atomic Energy Agency (IAEA) and the International Institute for Applied Systems Analysis (IIASA) was held at IIASA in Laxenburg, Austria, December 7-11, 1987, with the aim of reviewing existing models used in Probabilistic Safety Assessment (PSA) for Human Reliability Analysis (HRA) and of identifying the data required. The report collects both the contributions offered by the members of the Expert Task Force and the findings of the extensive discussions that took place during the meeting. Refs, figs and tabs

  12. Assessment of modelling needs for safety analysis of current HTGR concepts

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Van Tuyle, G.J.

    1985-12-01

    In view of the recent shift in emphasis of the DOE/Industry HTGR development efforts to smaller modular designs it became necessary to review the modelling needs and the codes available to assess the safety performance of these new designs. This report provides a final assessment of the most urgent modelling needs, comparing these to the tools available, and outlining the most significant areas where further modelling is required. Plans to implement the required work are presented. 47 refs., 20 figs

  13. Modeling the factors affecting unsafe behavior in the construction industry from safety supervisors' perspective.

    Science.gov (United States)

    Khosravi, Yahya; Asilian-Mahabadi, Hassan; Hajizadeh, Ebrahim; Hassanzadeh-Rangi, Narmin; Bastani, Hamid; Khavanin, Ali; Mortazavi, Seyed Bagher

    2014-01-01

    There can be little doubt that the construction is the most hazardous industry in the worldwide. This study was designed to modeling the factors affecting unsafe behavior from the perspective of safety supervisors. The qualitative research was conducted to extract a conceptual model. A structural model was then developed based on a questionnaire survey (n=266) by two stage Structural Equation Model (SEM) approach. An excellent confirmed 12-factors structure explained about 62% of variances unsafe behavior in the construction industry. A good fit structural model indicated that safety climate factors were positively correlated with safety individual factors (Pconstruction workers' engagement in safe or unsafe behavior. In order to improve construction safety performance, more focus on the workplace condition is required.

  14. 76 FR 64 - Safety and Health Requirements Related to Camp Cars

    Science.gov (United States)

    2011-01-03

    .... Water uses such as personal oral hygiene, drinking, food washing, preparation, cooking, cleaning of the... of Sec. 228.325 to ensure that the food service is safe and sanitary. FRA will hold the railroad... proposed section sets forth requirements regarding the safety of heating, cooking, ventilation, air...

  15. Safety Design Requirements for The Interior Architecture of Scientific Research Laboratories

    International Nuclear Information System (INIS)

    ElDib, A.A.

    2014-01-01

    The paper discusses one of the primary objectives of interior architecture design of research laboratories (specially those using radioactive materials) where it should provide a safe, accessible environment for laboratory personnel to conduct their work. A secondary objective is to allow for maximum flexibility for safe research. Therefore, health and safety hazards must be anticipated and carefully evaluated so that protective measures can be incorporated into the interior architectural design of these facilities wherever possible. The interior architecture requirements discussed in this paper illustrate some of the basic health and safety design features required for new and remodeled laboratories.The paper discusses one of the primary objectives of interior architecture design of research laboratories (specially those using radioactive materials) where it should provide a safe, accessible environment for laboratory personnel to conduct their work. A secondary objective is to allow for maximum flexibility for safe research. Therefore, health and safety hazards must be anticipated and carefully evaluated so that protective measures can be incorporated into the interior architectural design of these facilities wherever possible. The interior architecture requirements discussed in this paper illustrate some of the basic health and safety design features required for new and remodeled laboratories.

  16. 49 CFR 214.507 - Required safety equipment for new on-track roadway maintenance machines.

    Science.gov (United States)

    2010-10-01

    ... maintenance machines. 214.507 Section 214.507 Transportation Other Regulations Relating to Transportation... Roadway Maintenance Machines and Hi-Rail Vehicles § 214.507 Required safety equipment for new on-track roadway maintenance machines. (a) Each new on-track roadway maintenance machine shall be equipped with: (1...

  17. Plant design and layout of the different buildings with respect to safety, operational and maintenance requirements

    International Nuclear Information System (INIS)

    Liebich, H.

    1981-01-01

    The descriptions and pictures in this lecture show that the arrangement of the buildings and the location of components and systems are based on proven ideas with the aim to fulfil safety, operational and maintenance requirements also from the point of view of plant layout. (orig.)

  18. 78 FR 28987 - Revisions to Transportation Safety Requirements and Harmonization With International Atomic...

    Science.gov (United States)

    2013-05-16

    .... 115, ``International Basic Safety Standards for Protection against Ionizing Radiation and for the... paragraph 107(f) of TS-R-1, which addresses non-radioactive solid objects with radioactive substances..., ``Radiation protection--Sealed radioactive sources-- General requirements and classification,'' Second Edition...

  19. A Review of Safety and Design Requirements of the Artificial Pancreas

    NARCIS (Netherlands)

    Blauw, Helga; Keith-Hynes, Patrick; Koops, Robin; DeVries, J. Hans

    2016-01-01

    As clinical studies with artificial pancreas systems for automated blood glucose control in patients with type 1 diabetes move to unsupervised real-life settings, product development will be a focus of companies over the coming years. Directions or requirements regarding safety in the design of an

  20. Structural Design Requirements and Factors of Safety for Spaceflight Hardware: For Human Spaceflight. Revision A

    Science.gov (United States)

    Bernstein, Karen S.; Kujala, Rod; Fogt, Vince; Romine, Paul

    2011-01-01

    This document establishes the structural requirements for human-rated spaceflight hardware including launch vehicles, spacecraft and payloads. These requirements are applicable to Government Furnished Equipment activities as well as all related contractor, subcontractor and commercial efforts. These requirements are not imposed on systems other than human-rated spacecraft, such as ground test articles, but may be tailored for use in specific cases where it is prudent to do so such as for personnel safety or when assets are at risk. The requirements in this document are focused on design rather than verification. Implementation of the requirements is expected to be described in a Structural Verification Plan (SVP), which should describe the verification of each structural item for the applicable requirements. The SVP may also document unique verifications that meet or exceed these requirements with NASA Technical Authority approval.

  1. Leader communication approaches and patient safety: An integrated model.

    Science.gov (United States)

    Mattson, Malin; Hellgren, Johnny; Göransson, Sara

    2015-06-01

    Leader communication is known to influence a number of employee behaviors. When it comes to the relationship between leader communication and safety, the evidence is more scarce and ambiguous. The aim of the present study is to investigate whether and in what way leader communication relates to safety outcomes. The study examines two leader communication approaches: leader safety priority communication and feedback to subordinates. These approaches were assumed to affect safety outcomes via different employee behaviors. Questionnaire data, collected from 221 employees at two hospital wards, were analyzed using structural equation modeling. The two examined communication approaches were both positively related to safety outcomes, although leader safety priority communication was mediated by employee compliance and feedback communication by organizational citizenship behaviors. The findings suggest that leader communication plays a vital role in improving organizational and patient safety and that different communication approaches seem to positively affect different but equally essential employee safety behaviors. The results highlights the necessity for leaders to engage in one-way communication of safety values as well as in more relational feedback communication with their subordinates in order to enhance patient safety. Copyright © 2015 Elsevier Ltd. and National Safety Council. Published by Elsevier Ltd. All rights reserved.

  2. A Review of Safety and Design Requirements of the Artificial Pancreas.

    Science.gov (United States)

    Blauw, Helga; Keith-Hynes, Patrick; Koops, Robin; DeVries, J Hans

    2016-11-01

    As clinical studies with artificial pancreas systems for automated blood glucose control in patients with type 1 diabetes move to unsupervised real-life settings, product development will be a focus of companies over the coming years. Directions or requirements regarding safety in the design of an artificial pancreas are, however, lacking. This review aims to provide an overview and discussion of safety and design requirements of the artificial pancreas. We performed a structured literature search based on three search components-type 1 diabetes, artificial pancreas, and safety or design-and extended the discussion with our own experiences in developing artificial pancreas systems. The main hazards of the artificial pancreas are over- and under-dosing of insulin and, in case of a bi-hormonal system, of glucagon or other hormones. For each component of an artificial pancreas and for the complete system we identified safety issues related to these hazards and proposed control measures. Prerequisites that enable the control algorithms to provide safe closed-loop control are accurate and reliable input of glucose values, assured hormone delivery and an efficient user interface. In addition, the system configuration has important implications for safety, as close cooperation and data exchange between the different components is essential.

  3. Considerations on the Application of the IAEA Safety Requirements for the Design of Nuclear Power Plants

    International Nuclear Information System (INIS)

    2016-05-01

    Revised to take into consideration findings from the Fukushima Daiichi nuclear power plant accident, IAEA Safety Standards Series No. SSR-2/1 (Rev. 1), Safety of Nuclear Power Plants: Design, has introduced some new concepts with respect to the earlier safety standard published in the year 2000. The preparation of SSR-2/1 (Rev. 1) was carried out with constant and intense involvement of IAEA Member States, but some new requirements, because of the novelty of the concepts introduced and the complexity of the issues, are not always interpreted in a unique way. The IAEA is confident that a complete clarification and a full understanding of the new requirements will be available when the supporting safety guides for design and safety assessment of nuclear power plants are prepared. The IAEA expects that the effort devoted to the preparation of this publication, which received input and comments from several Member States and experts, will also facilitate and harmonize the preparation or revision of these supporting standards

  4. The art of regression modeling in road safety

    CERN Document Server

    Hauer, Ezra

    2015-01-01

    This unique book explains how to fashion useful regression models from commonly available data to erect models essential for evidence-based road safety management and research. Composed from techniques and best practices presented over many years of lectures and workshops, The Art of Regression Modeling in Road Safety illustrates that fruitful modeling cannot be done without substantive knowledge about the modeled phenomenon. Class-tested in courses and workshops across North America, the book is ideal for professionals, researchers, university professors, and graduate students with an interest in, or responsibilities related to, road safety. This book also: · Presents for the first time a powerful analytical tool for road safety researchers and practitioners · Includes problems and solutions in each chapter as well as data and spreadsheets for running models and PowerPoint presentation slides · Features pedagogy well-suited for graduate courses and workshops including problems, solutions, and PowerPoint p...

  5. Development and application of a living probabilistic safety assessment tool: Multi-objective multi-dimensional optimization of surveillance requirements in NPPs considering their ageing

    International Nuclear Information System (INIS)

    Kančev, Duško; Čepin, Marko; Gjorgiev, Blaže

    2014-01-01

    The benefits of utilizing the probabilistic safety assessment towards improvement of nuclear power plant safety are presented in this paper. Namely, a nuclear power plant risk reduction can be achieved by risk-informed optimization of the deterministically-determined surveillance requirements. A living probabilistic safety assessment tool for time-dependent risk analysis on component, system and plant level is developed. The study herein focuses on the application of this living probabilistic safety assessment tool as a computer platform for multi-objective multi-dimensional optimization of the surveillance requirements of selected safety equipment seen from the aspect of the risk-informed reasoning. The living probabilistic safety assessment tool is based on a newly developed model for calculating time-dependent unavailability of ageing safety equipment within nuclear power plants. By coupling the time-dependent unavailability model with a commercial software used for probabilistic safety assessment modelling on plant level, the frames of the new platform i.e. the living probabilistic safety assessment tool are established. In such way, the time-dependent core damage frequency is obtained and is further on utilized as first objective function within a multi-objective multi-dimensional optimization case study presented within this paper. The test and maintenance costs are designated as the second and the incurred dose due to performing the test and maintenance activities as the third objective function. The obtained results underline, in general, the usefulness and importance of a living probabilistic safety assessment, seen as a dynamic probabilistic safety assessment tool opposing the conventional, time-averaged unavailability-based, probabilistic safety assessment. The results of the optimization, in particular, indicate that test intervals derived as optimal differ from the deterministically-determined ones defined within the existing technical specifications

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

    International Nuclear Information System (INIS)

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

    2001-12-01

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

  7. Construction products performances and basic requirements for fire safety of facades in energy rehabilitation of buildings

    Directory of Open Access Journals (Sweden)

    Laban Mirjana Đ.

    2015-01-01

    Full Text Available Construction product means any product or kit which is produced and placed on the market for incorporation in a permanent manner in construction works, or parts thereof, and the performance of which has an effect on the performance of the construction works with respect to the basic requirements for construction works. Safety in case of fire and Energy economy and heat retention represent two among seven basic requirements which building has to meet according to contemporary technical rules on planning and construction. Performances of external walls building materials (particularly reaction to fire could significantly affect to fire spread on the façade and other building parts. Therefore, façade shaping and materialization in building renewal process, has to meet the fire safety requirement, as well as the energy requirement. Brief survey of fire protection regulations development in Serbia is presented in the paper. Preventive measures for fire risk reduction in building façade energy renewal are proposed according to contemporary fire safety requirements.

  8. Data requirements for the Ferrocyanide Safety Issue developed through the data quality objectives process

    International Nuclear Information System (INIS)

    Meacham, J.E.; Cash, R.J.; Dukelow, G.T.; Babad, H.; Buck, J.W.; Anderson, C.M.; Pulsipher, B.A.; Toth, J.J.; Turner, P.J.

    1994-08-01

    This document records the data quality objectives (DQO) process applied to the Ferrocyanide Safety Issue at the Hanford Site. Specifically, the major recommendations and findings from this Ferrocyanide DQO process are presented. The decision logic diagrams and decision error tolerances also are provided. The document includes the DQO sample-size formulas for determining specific tank sampling requirements, and many of the justifications for decision thresholds and decision error tolerances are briefly described. More detailed descriptions are presented in other Ferrocyanide Safety Program companion documents referenced in this report. This is a living document, and the assumptions contained within will be refined as more data from sampling and characterization become available

  9. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  10. Traffic monitoring and modeling for Intersection Safety

    NARCIS (Netherlands)

    Pyykönen, P.; Molinier, M.; Klunder, G.A.

    2010-01-01

    The INTERSAFE-2 project aims to develop and demonstrate a Cooperative Intersection Safety System that is able to significantly reduce injury and fatal accidents at intersections. The cooperative sensor data fusion is based on state-of-the-art and advanced on-board sensors for object recognition and

  11. Model Transformation for a System of Systems Dependability Safety Case

    Science.gov (United States)

    Murphy, Judy; Driskell, Steve

    2011-01-01

    The presentation reviews the dependability and safety effort of NASA's Independent Verification and Validation Facility. Topics include: safety engineering process, applications to non-space environment, Phase I overview, process creation, sample SRM artifact, Phase I end result, Phase II model transformation, fault management, and applying Phase II to individual projects.

  12. Revised health and safety compliance model for the Ghanaian ...

    African Journals Online (AJOL)

    The construction industry in Ghana is faced with employees' negligence in obeying rules and regulations, and acts that conflict with health and safety. The purpose of the paper was to present the revised health and safety (H&S) compliance model for the construction industry based on a developed theoretical six factor ...

  13. Mathematical models for prediction of safety factors for a simply ...

    African Journals Online (AJOL)

    From the results obtained, mathematical prediction models were developed using a least square regression analysis for bending, shear and deflection modes of failure considered in the study. The results showed that the safety factors for material, dead and live load are not unique, but they are influenced by safety index ...

  14. Education and training requirements in the revised European Basic Safety Standards Directive

    International Nuclear Information System (INIS)

    Mundigl, S.

    2009-01-01

    The European Commission is currently developing a modified European Basic Safety Standards Directive covering two major objectives: the consolidation of existing European Radiation Protection legislation, and the revision of the European Basic Safety Standards. The consolidation will merge the following five Directives into one single Directive: the Basic Safety Standards Directive, the Medical Exposures Directive, the Public Information Directive, the Outside Workers Directive, and the Directive on the Control of high-activity sealed radioactive sources and orphan sources. The revision of the European Basic Safety Standards will take account of the latest recommendations by the International Commission on Radiological Protection (ICRP) and shall improve clarity of the requirements where appropriate. It is planned to introduce more binding requirements on natural radiation sources, on criteria for clearance, and on the cooperation between Member States for emergency planning and response, as well as a graded approach for regulatory control. One additional goal is to achieve greater harmonisation between the European BSS and the international BSS. Following a recommendation from the Article 31 Group of Experts, the current draft of the modified BSS will highlight the importance of education and training by dedicating a specific title to radiation protection education, training and information. This title will include a general requirement on the Member States to ensure the establishment of an adequate legislative and administrative framework for providing appropriate radiation protection education, training and information. In addition, there will be specific requirements on training in the medical field, on information and training of workers in general, of workers potentially exposed to orphan sources, and to emergency workers. The revised BSS directive will include requirements on the competence of a radiation protection expert (RPE) and of a radiation protection

  15. Road network safety evaluation using Bayesian hierarchical joint model.

    Science.gov (United States)

    Wang, Jie; Huang, Helai

    2016-05-01

    Safety and efficiency are commonly regarded as two significant performance indicators of transportation systems. In practice, road network planning has focused on road capacity and transport efficiency whereas the safety level of a road network has received little attention in the planning stage. This study develops a Bayesian hierarchical joint model for road network safety evaluation to help planners take traffic safety into account when planning a road network. The proposed model establishes relationships between road network risk and micro-level variables related to road entities and traffic volume, as well as socioeconomic, trip generation and network density variables at macro level which are generally used for long term transportation plans. In addition, network spatial correlation between intersections and their connected road segments is also considered in the model. A road network is elaborately selected in order to compare the proposed hierarchical joint model with a previous joint model and a negative binomial model. According to the results of the model comparison, the hierarchical joint model outperforms the joint model and negative binomial model in terms of the goodness-of-fit and predictive performance, which indicates the reasonableness of considering the hierarchical data structure in crash prediction and analysis. Moreover, both random effects at the TAZ level and the spatial correlation between intersections and their adjacent segments are found to be significant, supporting the employment of the hierarchical joint model as an alternative in road-network-level safety modeling as well. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. Bayesian Safety Risk Modeling of Human-Flightdeck Automation Interaction

    Science.gov (United States)

    Ancel, Ersin; Shih, Ann T.

    2015-01-01

    Usage of automatic systems in airliners has increased fuel efficiency, added extra capabilities, enhanced safety and reliability, as well as provide improved passenger comfort since its introduction in the late 80's. However, original automation benefits, including reduced flight crew workload, human errors or training requirements, were not achieved as originally expected. Instead, automation introduced new failure modes, redistributed, and sometimes increased workload, brought in new cognitive and attention demands, and increased training requirements. Modern airliners have numerous flight modes, providing more flexibility (and inherently more complexity) to the flight crew. However, the price to pay for the increased flexibility is the need for increased mode awareness, as well as the need to supervise, understand, and predict automated system behavior. Also, over-reliance on automation is linked to manual flight skill degradation and complacency in commercial pilots. As a result, recent accidents involving human errors are often caused by the interactions between humans and the automated systems (e.g., the breakdown in man-machine coordination), deteriorated manual flying skills, and/or loss of situational awareness due to heavy dependence on automated systems. This paper describes the development of the increased complexity and reliance on automation baseline model, named FLAP for FLightdeck Automation Problems. The model development process starts with a comprehensive literature review followed by the construction of a framework comprised of high-level causal factors leading to an automation-related flight anomaly. The framework was then converted into a Bayesian Belief Network (BBN) using the Hugin Software v7.8. The effects of automation on flight crew are incorporated into the model, including flight skill degradation, increased cognitive demand and training requirements along with their interactions. Besides flight crew deficiencies, automation system

  17. 78 FR 42889 - Pipeline Safety: Reminder of Requirements for Utility LP-Gas and LPG Pipeline Systems

    Science.gov (United States)

    2013-07-18

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Part 192 [Docket No. PHMSA-2013-0097] Pipeline Safety: Reminder of Requirements for Utility LP-Gas and LPG Pipeline Systems AGENCY: Pipeline and Hazardous Materials Safety Administration (PHMSA), DOT. ACTION...

  18. 78 FR 55230 - Safety and Environmental Management System Requirements for Vessels on the U.S. Outer Continental...

    Science.gov (United States)

    2013-09-10

    ...\\ including the regulation of workplace safety and health.\\2\\ The Coast Guard's regulatory authority extends... 147 [Docket No. USCG-2012-0779] RIN 1625-AC05 Safety and Environmental Management System Requirements... a vessel-specific Safety and Environmental Management System (SEMS) that incorporates the management...

  19. Descriptions and models of safety functions - a prestudy

    International Nuclear Information System (INIS)

    Harms-Ringdahl, L.

    1999-09-01

    A study has been made with the focus on different theories and applications concerning 'safety functions' and 'barriers'. In this report, a safety function is defined as a technical or organisational function with the aim to reduce probability and/or consequences associated with a hazard. The study contains a limited review of practice and theories related to safety, with a focus on applications from nuclear and industrial safety. The study is based on a literature review and interviews. A summary has been made of definitions and terminology, which shows a large variation. E.g. 'barrier' can have a precise physical and technical meaning, or it can include human, technical and organisational elements. Only a few theoretical models describing safety functions have been found. One section of the report summarises problems related to safety issues and procedures. They concern errors in procedure design and user compliance. A proposal for describing and structuring safety functions has been made. Dimensions in a description could be degree of abstraction, systems level, the different parts of the function, etc. A model for safety functions has been proposed, which includes the division of a safety function in a number connected 'safety function elements'. One conclusion is that there is a potential for improving theories and tools for safety work and procedures. Safety function could be a useful concept in such a development, and advantages and disadvantages with this is discussed. If further work should be done, it is recommended that this is made as a combination of theoretical analysis and case studies

  20. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    International Nuclear Information System (INIS)

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-01-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  1. FMCSA safety program effectiveness measurement: intervention model fiscal year 2009.

    Science.gov (United States)

    2013-04-01

    The Federal Motor Carrier Safety Administration (FMCSA), in cooperation with the researcher, has developed an analytic model to measure the effectiveness of roadside inspections and traffic enforcements in terms of crashes avoided, injuries avoided, ...

  2. Risk and safety requirements for diagnostic and therapeutic procedures in allergology: World Allergy Organization Statement

    Directory of Open Access Journals (Sweden)

    Marek L. Kowalski

    2016-10-01

    Full Text Available Abstract One of the major concerns in the practice of allergy is related to the safety of procedures for the diagnosis and treatment of allergic disease. Management (diagnosis and treatment of hypersensitivity disorders involves often intentional exposure to potentially allergenic substances (during skin testing, deliberate induction in the office of allergic symptoms to offending compounds (provocation tests or intentional application of potentially dangerous substances (allergy vaccine to sensitized patients. These situations may be associated with a significant risk of unwanted, excessive or even dangerous reactions, which in many instances cannot be completely avoided. However, adverse reactions can be minimized or even avoided if a physician is fully aware of potential risk and is prepared to appropriately handle the situation. Information on the risk of diagnostic and therapeutic procedures in allergic diseases has been accumulated in the medical literature for decades; however, except for allergen specific immunotherapy, it has never been presented in a systematic fashion. Up to now no single document addressed the risk of the most commonly used medical procedures in the allergy office nor attempted to present general requirements necessary to assure the safety of these procedures. Following review of available literature a group of allergy experts within the World Allergy Organization (WAO, representing various continents and areas of allergy expertise, presents this report on risk associated with diagnostic and therapeutic procedures in allergology and proposes a consensus on safety requirements for performing procedures in allergy offices. Optimal safety measures including appropriate location, type and required time of supervision, availability of safety equipment, access to specialized emergency services, etc. for various procedures have been recommended. This document should be useful for allergists with already established

  3. Edible safety requirements and assessment standards for agricultural genetically modified organisms.

    Science.gov (United States)

    Deng, Pingjian; Zhou, Xiangyang; Zhou, Peng; Du, Zhong; Hou, Hongli; Yang, Dongyan; Tan, Jianjun; Wu, Xiaojin; Zhang, Jinzhou; Yang, Yongcun; Liu, Jin; Liu, Guihua; Li, Yonghong; Liu, Jianjun; Yu, Lei; Fang, Shisong; Yang, Xiaoke

    2008-05-01

    This paper describes the background, principles, concepts and methods of framing the technical regulation for edible safety requirement and assessment of agricultural genetically modified organisms (agri-GMOs) for Shenzhen Special Economic Zone in the People's Republic of China. It provides a set of systematic criteria for edible safety requirements and the assessment process for agri-GMOs. First, focusing on the degree of risk and impact of different agri-GMOs, we developed hazard grades for toxicity, allergenicity, anti-nutrition effects, and unintended effects and standards for the impact type of genetic manipulation. Second, for assessing edible safety, we developed indexes and standards for different hazard grades of recipient organisms, for the influence of types of genetic manipulation and hazard grades of agri-GMOs. To evaluate the applicability of these criteria and their congruency with other safety assessment systems for GMOs applied by related organizations all over the world, we selected some agri-GMOs (soybean, maize, potato, capsicum and yeast) as cases to put through our new assessment system, and compared our results with the previous assessments. It turned out that the result of each of the cases was congruent with the original assessment.

  4. Requirements of radiation and safety protection for NORM in petroleum and gas facilities

    International Nuclear Information System (INIS)

    Machavane, Edna Felicina Lisboa

    2017-01-01

    The work establishes radiation protection and safety requirements for NORM in oil and gas installations, enabling the National Atomic Energy Agency to draw up regulations on NORM. A bibliographic review and measurement of oil sludge activity concentrations was carried out to reach the objective. Significant amounts of NORM originating from reservoir rock are encountered during production, maintenance and decommissioning. The oil and gas industry operates in all climates and environments including the most arduous conditions and is continually challenged to achieve high operating efficiency while maintaining a high standard of safety and control - this includes the need to maintain control over exposure as well as protecting the public and the environment through the proper management of tailings that may be radiologically and chemically hazardous. The main objective of this work was not only to present the main radiological protection and safety requirements for NORM in oil and gas installations, but also to guide the competent governmental authorities of the Republic of Mozambique, that the installation of a radiometry laboratory and elaboration of NORM regulations involve a great control of radiological safety. The regulatory authority is responsible for authorizing facilities for the storage of radioactive waste, including the storage of contaminated tailings. It is recommended that studies of this kind be made to analyze the concentration of naturally occurring radioisotope activity. (author)

  5. A SIL quantification approach based on an operating situation model for safety evaluation in complex guided transportation systems

    International Nuclear Information System (INIS)

    Beugin, J.; Renaux, D.; Cauffriez, L.

    2007-01-01

    Safety analysis in guided transportation systems is essential to avoid rare but potentially catastrophic accidents. This article presents a quantitative probabilistic model that integrates Safety Integrity Levels (SIL) for evaluating the safety of such systems. The standardized SIL indicator allows the safety requirements of each safety subsystem, function and/or piece of equipment to be specified, making SILs pivotal parameters in safety evaluation. However, different interpretations of SIL exist, and faced with the complexity of guided transportation systems, the current SIL allocation methods are inadequate for the task of safety assessment. To remedy these problems, the model developed in this paper seeks to verify, during the design phase of guided transportation system, whether or not the safety specifications established by the transport authorities allow the overall safety target to be attained (i.e., if the SIL allocated to the different safety functions are sufficient to ensure the required level of safety). To meet this objective, the model is based both on the operating situation concept and on Monte Carlo simulation. The former allows safety systems to be formalized and their dynamics to be analyzed in order to show the evolution of the system in time and space, and the latter make it possible to perform probabilistic calculations based on the scenario structure obtained

  6. Modelling of Condensation in Vertical Tubes for Passive Safety System

    International Nuclear Information System (INIS)

    Papini, D.; Ricotti, M.; Santini, L.; Grgic, D.

    2008-01-01

    Condensation in vertical tubes plays an important role in the performance of heat exchangers in passive safety systems, widely adopted in next generation reactors. Vertical pipe condensers are implemented in the GE-SBWR1000 Isolation Condenser as well as in the Emergency Heat Removal System (EHRS) of the IRIS reactor. The transient and safety analysis is usually carried out by means of best-estimate, thermalhydraulic codes, as RELAP. Suitable heat transfer correlations are required to duly model the two-phase processes. As far as the condensation process is concerned, RELAP5/MOD3.3 adopts the Nusselt correlation to calculate the heat transfer coefficient in laminar conditions and the Shah correlation for turbulent conditions; the maximum of the predictions from laminar and turbulent regimes is used to calculate the condensation heat transfer coefficient. Shah correlation is generally considered as the best empirical correlation for turbulent annular film condensation, but suitable in proper ranges of the various parameters. Nevertheless, recent investigations have pointed out that its validity is highly questionable for high pressure and large diameter tube applications with water, as should be for the utilization for vertical tube condensers in passive safety systems. Thus, a best-estimate model, based on the theory of film condensation on a plain wall, is proposed. Condensate velocity, expressed in terms of Reynolds number, governs the development of three different regime zones: laminar, laminar wavy and turbulent. The best correlation for each regime (Nusselt's for laminar, Kutateladze's for laminar wavy and Chen's for turbulent) is considered and then implemented in RELAP code. Comparison between the Nusselt-Shah and the proposed model shows substantial differences in heat transfer coefficient prediction. Especially, a trend of increasing value of the heat transfer coefficient with tube abscissa (and quality decreasing) is predicted, when turbulence

  7. Development of the switch requirements and architecture of a safety data communication system

    International Nuclear Information System (INIS)

    Jeong, K.I.; Lee, J.K.; Park, H.Y.; Koo, I.S.

    2004-12-01

    In accordance with digitalising the Instrumentation and Control(I and C) systems in the integral reactor, a communication network is required for effective information exchanges between the different equipment, an enhancement of the design flexibility, a simple installation and cost reduction. Generally, a communication network consists of a topology, the protocol, a communication medium, an interconnection device, etc. In this report, the development methods of switch and the architecture of a Safety Data Communication System(SDCS) are investigated and analyzed. In this report, the design requirements for switch are presented, which are the essential requirements to develop the switch in a SDCS of the SMART-P. To establish these requirements, the evaluation and analysis of the design and implementation method of the COTS switches, the architecture of SDCS and the design requirements of a SDCS were performed. At the detail design stage, these requirements will be used for the top-tier requirements, especially the design target and design basis. To develop the detail design requirements in the future, more quantitative and qualitative analyses are required. In the case of selecting the COTS switch and developing the switch, these requirements will also be used for the evaluation guide

  8. Development of the switch requirements and architecture of a safety data communication system

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, K.I.; Lee, J.K.; Park, H.Y.; Koo, I.S

    2004-12-01

    In accordance with digitalising the Instrumentation and Control(I and C) systems in the integral reactor, a communication network is required for effective information exchanges between the different equipment, an enhancement of the design flexibility, a simple installation and cost reduction. Generally, a communication network consists of a topology, the protocol, a communication medium, an interconnection device, etc. In this report, the development methods of switch and the architecture of a Safety Data Communication System(SDCS) are investigated and analyzed. In this report, the design requirements for switch are presented, which are the essential requirements to develop the switch in a SDCS of the SMART-P. To establish these requirements, the evaluation and analysis of the design and implementation method of the COTS switches, the architecture of SDCS and the design requirements of a SDCS were performed. At the detail design stage, these requirements will be used for the top-tier requirements, especially the design target and design basis. To develop the detail design requirements in the future, more quantitative and qualitative analyses are required. In the case of selecting the COTS switch and developing the switch, these requirements will also be used for the evaluation guide.

  9. Information requirements of the National Aeronautics and Space Administration's safety, environmental health, and occupational medicine programs

    Science.gov (United States)

    Whyte, A. A.

    1978-01-01

    A survey of the internal and external reporting and recordkeeping procedures of these programs was conducted and the major problems associated with them are outlined. The impact of probable future requirements on existing information systems is evaluated. This report also presents the benefits of combining the safety and health information systems into one computerized system and recommendations for the development and scope of that system.

  10. Statistical modelling of traffic safety development

    DEFF Research Database (Denmark)

    Christens, Peter

    2004-01-01

    there were 6861 injury trafficc accidents reported by the police, resulting in 4519 minor injuries, 3946 serious injuries, and 431 fatalities. The general purpose of the research was to improve the insight into aggregated road safety methodology in Denmark. The aim was to analyse advanced statistical methods......, that were designed to study developments over time, including effects of interventions. This aim has been achieved by investigating variations in aggregated Danish traffic accident series and by applying state of the art methodologies to specific case studies. The thesis comprises an introduction...

  11. Proposal for basic safety requirements regarding the disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    1980-04-01

    A working group commissioned to prepare proposals for basic safety requirements for the storage and transport of radioactive waste prepared its report to the Danish Agency of Environmental Protection. The proposals include: radiation protection requirements, requirements concerning the properties of high-level waste units, the geological conditions of the waste disposal location, the supervision of waste disposal areas. The proposed primary requirements for safety evaluation of the disposal of high-level waste in deep geological formations are of a general nature, not being tied to specific assumptions regarding the waste itself, the geological and other conditions at the place of disposal, and the technical methods of disposal. It was impossible to test the proposals for requirements on a working repository. As no country has, to the knowledge of the working group, actually disposed of hifg-level radioactive waste or approved of plans for such disposal. Methods for evaluating the suitability of geological formations for waste disposal, and background material concerning the preparation of these proposals for basic safety requirements relating to radiation, waste handling and geological conditions are reviewed. Appended to the report is a description of the phases of the fuel cycle that are related to the storage of spent fuel and the disposal of high-level reprocessing waste in a salt formation. It should be noted that the proposals of the working group are not limited to the disposal of reprocessed fuel, but also include the direct disposal of spent fuel as well as disposal in geological formations other than salt. (EG)

  12. Key natural analogue input required to build a safety case for direct disposal of spent nuclear fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    McKinley, I.G.; Hardie, S.M.L.; Klein, E. [MCM Consulting, Baden-Dättwil (Switzerland); Kawamura, H. [Obayashi Corporation, Nuclear Facilities Division, Tokyo (Japan); Beattie, T.M. [MCM Consulting, Bristol (United Kingdom)

    2015-06-15

    Natural analogues have been previously used to support the safety case for direct disposal of spent nuclear fuel, but the focus of such work was very dependent on the key barriers of specific national disposal concepts. Investigations of the feasibility of such disposal in Japan are at an early stage but, nevertheless, it is clear that building a robust safety case will be very challenging and would benefit from focused support from natural analogue studies—both in terms of developing/testing required models and, as importantly, presenting safety arguments to a wide range of stakeholders. This paper identifies key analogues that support both longevity and spread of failure times of massive steel overpacks, the effectiveness of buffering of radiolytic oxidants and the chemical and physical mechanisms retarding release of radionuclides from the engineered barriers. It is concluded that, for countries like Japan where performance needs to be assessed as realistically as possible, natural analogues can complement the existing laboratory and theoretical knowledge base and contribute towards development of a robust safety case. (authors)

  13. Modelling requirements for future assessments based on FEP analysis

    International Nuclear Information System (INIS)

    Locke, J.; Bailey, L.

    1998-01-01

    This report forms part of a suite of documents describing the Nirex model development programme. The programme is designed to provide a clear audit trail from the identification of significant features, events and processes (FEPs) to the models and modelling processes employed within a detailed safety assessment. A scenario approach to performance assessment has been adopted. It is proposed that potential evolutions of a deep geological radioactive waste repository can be represented by a base scenario and a number of variant scenarios. The base scenario is chosen to be broad-ranging and to represent the natural evolution of the repository system and its surrounding environment. The base scenario is defined to include all those FEPs that are certain to occur and those which are judged likely to occur for a significant period of the assessment timescale. The structuring of FEPs on a Master Directed Diagram (MDD) provides a systematic framework for identifying those FEPs that form part of the natural evolution of the system and those, which may define alternative potential evolutions of the repository system. In order to construct a description of the base scenario, FEPs have been grouped into a series of conceptual models. Conceptual models are groups of FEPs, identified from the MDD, representing a specific component or process within the disposal system. It has been found appropriate to define conceptual models in terms of the three main components of the disposal system: the repository engineered system, the surrounding geosphere and the biosphere. For each of these components, conceptual models provide a description of the relevant subsystem in terms of its initial characteristics, subsequent evolution and the processes affecting radionuclide transport for the groundwater and gas pathways. The aim of this document is to present the methodology that has been developed for deriving modelling requirements and to illustrate the application of the methodology by

  14. Metamodel comparison and model comparison for safety assurance

    NARCIS (Netherlands)

    Luo, Y.; Engelen, L.J.P.; Brand, van den M.G.J.; Bondavelli, A.; Ceccarelli, A.; Ortmeier, F.

    2014-01-01

    In safety-critical domains, conceptual models are created in the form of metamodels using different concepts from possibly overlapping domains. Comparison between those conceptual models can facilitate the reuse of models from one domain to another. This paper describes the mappings detected when

  15. Applying a realistic evaluation model to occupational safety interventions

    DEFF Research Database (Denmark)

    Pedersen, Louise Møller

    2018-01-01

    Background: Recent literature characterizes occupational safety interventions as complex social activities, applied in complex and dynamic social systems. Hence, the actual outcomes of an intervention will vary, depending on the intervention, the implementation process, context, personal characte......Background: Recent literature characterizes occupational safety interventions as complex social activities, applied in complex and dynamic social systems. Hence, the actual outcomes of an intervention will vary, depending on the intervention, the implementation process, context, personal...... and qualitative methods. This revised model has, however, not been applied in a real life context. Method: The model is applied in a controlled, four-component, integrated behaviour-based and safety culture-based safety intervention study (2008-2010) in a medium-sized wood manufacturing company. The interventions...... involve the company’s safety committee, safety manager, safety groups and 130 workers. Results: The model provides a framework for more valid evidence of what works within injury prevention. Affective commitment and role behaviour among key actors are identified as crucial for the implementation...

  16. Innovative nuclear reactor - Indian approach to meet user requirements for safety

    International Nuclear Information System (INIS)

    Saha, D.; Sinha, R.K.

    2002-01-01

    Full text: For sustainable development of nuclear energy, a number of key issues are to be addressed. It should be economically competitive; it must address the issues related to nuclear safety, proliferation resistance, environmental impact, waste disposal and cross cutting issues like social and infra-structural aspects. To compete successfully in the long term, in the highly competitive energy market and to overcome other challenges, it is necessary to introduce innovative reactor and fuel cycle concepts. Indian Advanced Heavy Water Reactor (AHWR) is one such innovative reactor. To guide the research and development activities related to innovative concepts, user requirements are to be formulated. User requirements covering various aspects of sustainable development are being formulated at both national and international levels. One such international project involved in the formulation of user requirements is the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper deals with INPRO user requirements for safety and Indian approach to meet these requirements through AHWR

  17. Gamified Requirements Engineering: Model and Experimentation

    NARCIS (Netherlands)

    Lombriser, Philipp; Dalpiaz, Fabiano; Lucassen, Garm; Brinkkemper, Sjaak

    2016-01-01

    [Context & Motivation] Engaging stakeholders in requirements engineering (RE) influences the quality of the requirements and ultimately of the system to-be. Unfortunately, stakeholder engagement is often insufficient, leading to too few, low-quality requirements. [Question/problem] We aim to

  18. Modular reliability modeling of the TJNAF personnel safety system

    International Nuclear Information System (INIS)

    Cinnamon, J.; Mahoney, K.

    1997-01-01

    A reliability model for the Thomas Jefferson National Accelerator Facility (formerly CEBAF) personnel safety system has been developed. The model, which was implemented using an Excel spreadsheet, allows simulation of all or parts of the system. Modularity os the model's implementation allows rapid open-quotes what if open-quotes case studies to simulate change in safety system parameters such as redundancy, diversity, and failure rates. Particular emphasis is given to the prediction of failure modes which would result in the failure of both of the redundant safety interlock systems. In addition to the calculation of the predicted reliability of the safety system, the model also calculates availability of the same system. Such calculations allow the user to make tradeoff studies between reliability and availability, and to target resources to improving those parts of the system which would most benefit from redesign or upgrade. The model includes calculated, manufacturer's data, and Jefferson Lab field data. This paper describes the model, methods used, and comparison of calculated to actual data for the Jefferson Lab personnel safety system. Examples are given to illustrate the model's utility and ease of use

  19. Safety of High Speed Magnetic Levitation Transportation Systems - Comparison of U.S. and Foreign Safety Requirements for Application to U.S. Maglev Systems

    Science.gov (United States)

    1993-09-01

    This report presents the results of a systematic review of the safety requirements selected for the German Transrapid : electromagnetic (EMS) type maglev system to determine their applicability and completeness with respect to the : construction and ...

  20. Probabilistic approaches to LCO's and surveillance requirements for standby safety systems

    International Nuclear Information System (INIS)

    Lofgren, E.V.; Varcolik, F.

    1982-11-01

    Results are presented for a comprehensive analysis of risk-based methods for establishing Limiting Conditions for Operation (LCO) and surveillance requirements for on-line test and repair of nuclear power plant safety system components. Limiting Conditions for Operation refers to the legal constraint on safety system component outage times that are imposed by the NRC as part of the reactor operating license. Generally, when a safety system component is removed for repair or test for a period of time there is a period of increased vulnerability concerning the probability that the affected safety system will be available to mitigate an accident. This period of increased vulnerability exists until the component is restored to service. The constraint on the duration of this period, the allowed outage time (AOT), is the aspect of LCOs that is of interest here. In particular, methods are reviewed and developed that relate measures of risk to the AOT. Only by explicitly relating risk to AOT can outage times be constrained by placing limits on risk. Methods developed for relating risk measures to outage times are presented. The review and analysis of risk related methods for establishing LCOs are described

  1. Requirement analysis of the safety-critical software implementation for the nuclear power plant

    International Nuclear Information System (INIS)

    Chang, Hoon Seon; Jung, Jae Cheon; Kim, Jae Hack; Nam, Sang Ku; Kim, Hang Bae

    2005-01-01

    The safety critical software shall be implemented under the strict regulation and standards along with hardware qualification. In general, the safety critical software has been implemented using functional block language (FBL) and structured language like C in the real project. Software design shall comply with such characteristics as; modularity, simplicity, minimizing the use of sub-routine, and excluding the interrupt logic. To meet these prerequisites, we used the computer-aided software engineering (CASE) tool to substantiate the requirements traceability matrix that were manually developed using Word processors or Spreadsheets. And the coding standard and manual have been developed to confirm the quality of software development process, such as; readability, consistency, and maintainability in compliance with NUREG/CR-6463. System level preliminary hazard analysis (PHA) is performed by analyzing preliminary safety analysis report (PSAR) and FMEA document. The modularity concept is effectively implemented for the overall module configurations and functions using RTP software development tool. The response time imposed on the basis of the deterministic structure of the safety-critical software was measured

  2. Safety requirements and options for a large size fast neutron reactor

    International Nuclear Information System (INIS)

    Cogne, F.; Megy, J.; Robert, E.; Benmergui, A.; Villeneuve, J.

    1977-01-01

    Starting from the experience gained in the safety evaluation of the PHENIX reactor, and from results already obtained in the safety studies on fast neutron reactors, the French regulatory bodies have defined since 1973 what could be the requirements and the recommendations in the matter of safety for the first large size ''prototype'' fast neutron power plant of 1200 MWe. Those requirements and recommendations, while not being compulsory due to the evolution of this type of reactors, will be used as a basis for the technical regulation that will be established in France in this field. They define particularly the care to be taken in the following areas which are essential for safety: the protection systems, the primary coolant system, the prevention of accidents at the core level, the measures to be taken with regard to the whole core accident and to the containment, the protection against sodium fires, and the design as a function of external aggressions. In applying these recommendations, the CREYS-MALVILLE plant designers have tried to achieve redundancy in the safety related systems and have justified the safety of the design with regard to the various involved phenomena. In particular, the extensive research made at the levels of the fuel and of the core instrumentation makes it possible to achieve the best defence to avoid the development of core accidents. The overall examination of the measures taken, from the standpoint of prevention and surveyance as well as from the standpoint of means of action led the French regulatory bodies to propose the construction permit of the CREYS MALVILLE plant, provided that additional examinations by the regulatory bodies be made during the construction of the plant on some technological aspects not fully clarified at the authorization time. The conservatism of the corresponding requirements should be demonstrated prior to the commissioning of the power plant. To pursue a programme on reactors of this type, or even more

  3. Integrated model of port oil piping transportation system safety including operating environment threats

    OpenAIRE

    Kołowrocki, Krzysztof; Kuligowska, Ewa; Soszyńska-Budny, Joanna

    2017-01-01

    The paper presents an integrated general model of complex technical system, linking its multistate safety model and the model of its operation process including operating environment threats and considering variable at different operation states its safety structures and its components safety parameters. Under the assumption that the system has exponential safety function, the safety characteristics of the port oil piping transportation system are determined.

  4. A study on the primary requirement for the safety of the Wolsong tritium removal facility

    International Nuclear Information System (INIS)

    Hwang, K. H.; Lee, K. J.; Jeong, C. W.

    2001-01-01

    Owing to the using a heavy water as a moderator and a coolant in Heavy water reactor, A large mount of tritium is produced due to a reaction of deuterium with neutron in the reactor and some of tritium is released to the environment. In Wolsong, 4 units (CANDU-600 type) Heavy water reactor is in operation. And the generated amount of tritium is increased with the increase of operational year of the Wolsong nuclear reactor. Decommissioning of the Wolsong unit 1 is expected to start at 2013. Before 2013, to reduce the workers internal radiation doses and environmental release of tritium, Tritium Removal Facility (TRF) is required and should be operated. Wolsong TRF (WTRF) is under developing stage by Korea Electric Power Corporation(KEPCO)and scheduled to start operation about 2006. Once the facility begins operation it can be contributed to the greatly reduction of tritium release to the environment and worker's expose. In this situation, study about the safety assessment method and regulatory requirement is essential for safety insurance of WTRF. And this helps the safety acquirement, successful operation and reliance of WTRF

  5. Evaluation of safety requirements of erbium laser equipment used in dentistry

    International Nuclear Information System (INIS)

    Braga, Flavio Hamilton

    2002-01-01

    The erbium laser (Er:YAG) has been used in several therapeutic processes. Erbium lasers, however, operate with energies capable to produce lesions in biological tissues. Aiming the safe use, the commercialization of therapeutic laser equipment is controlled in Brazil, where the equipment should comply with quality and safety requirement prescribed in technical regulations. The objective of this work is to evaluate the quality and safety requirements of a commercial therapeutic erbium laser according to Brazilian regulations, and to discuss a risk control program intended to minimize the accidental exposition at dangerous laser radiation levels. It was verified that the analyzed laser can produce lesions in the skin and eyes, when exposed to laser radiation at distances smaller than 80 cm by 10 s or more. In these conditions, the use of protection glasses is recommended to the personnel that have access to the laser operation ambient. It was verified that the user's training and the presence of a target indicator are fundamental to avoid damages in the skin and buccal cavity. It was also verified that the knowledge and the correct use of the equipment safety devices, and the application of technical and administrative measures is efficient to minimize the risk of dangerous expositions to the laser radiation. (author)

  6. Identifying environmental safety and health requirements for the Fernald Environmental Restoration Management Corporation

    International Nuclear Information System (INIS)

    Beckman, W.H.; Cossel, S.C.; Alhadeff, N.; Lindamood, S.B.; Beers, J.A.

    1994-01-01

    This presentation will describe the Fernald Environmental Restoration Management Corporation's (FERMCO) Standards/Requirements Identification Documents (S/RlDs) Program, the unique process used to implement it, and the status of the program. We will also discuss the lessons learned as the program was implemented. The Department of Energy (DOE) established the Fernald site to produce uranium metals for the nation's defense programs in 1953. In 1989, DOE suspended production and, in 1991, the mission of the site was formally changed to one of environmental cleanup and restoration. The site was renamed the Fernald Environmental Management Project (FEMP). FERMCO's mission is to provide safe, early, and least-cost final clean-up of the site in compliance with all regulations and commitments. DOE has managed nuclear facilities primarily through its oversight of Management and Operating contractors. Comprehensive nuclear industry standards were absent when most DOE sites were first established, Management and Operating contractors had to apply existing non-nuclear industry standards and, in many cases, formulate new technical standards. Because it was satisfied with the operation of its facilities, DOE did not incorporate modern practices and standards as they became available. In March 1990, the Defense Nuclear Facilities Safety Board issued Recommendation 90-2, which called for DOE to identify relevant standards and requirements, conduct adequacy assessments of requirements in protecting environmental, public, and worker health and safety, and determine the extent to which the requirements are being implemented. The Environmental Restoration and Waste Management Office of DOE embraced the recommendation for facilities under its control. Strict accountability requirements made it essential that FERMCO and DOE clearly identify applicable requirements necessary, determine the requirements' adequacy, and assess FERMCO's level of compliance

  7. Standard model for safety analysis report of hexafluoride power plants from natural uranium

    International Nuclear Information System (INIS)

    1983-01-01

    The standard model for safety analysis report for hexafluoride production power plants from natural uranium is presented, showing the presentation form, the nature and the degree of detail, of the minimal information required by the Brazilian Nuclear Energy Commission - CNEN. (E.G.) [pt

  8. From requirements to Java in a snap model-driven requirements engineering in practice

    CERN Document Server

    Smialek, Michal

    2015-01-01

    This book provides a coherent methodology for Model-Driven Requirements Engineering which stresses the systematic treatment of requirements within the realm of modelling and model transformations. The underlying basic assumption is that detailed requirements models are used as first-class artefacts playing a direct role in constructing software. To this end, the book presents the Requirements Specification Language (RSL) that allows precision and formality, which eventually permits automation of the process of turning requirements into a working system by applying model transformations and co

  9. Psychometric model for safety culture assessment in nuclear research facilities

    International Nuclear Information System (INIS)

    Nascimento, C.S. do; Andrade, D.A.; Mesquita, R.N. de

    2017-01-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  10. Psychometric model for safety culture assessment in nuclear research facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, C.S. do, E-mail: claudio.souza@ctmsp.mar.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), Av. Professor Lineu Prestes 2468, 05508-000 São Paulo, SP (Brazil); Andrade, D.A., E-mail: delvonei@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil); Mesquita, R.N. de, E-mail: rnavarro@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil)

    2017-04-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  11. Modeling uncertainty in requirements engineering decision support

    Science.gov (United States)

    Feather, Martin S.; Maynard-Zhang, Pedrito; Kiper, James D.

    2005-01-01

    One inherent characteristic of requrements engineering is a lack of certainty during this early phase of a project. Nevertheless, decisions about requirements must be made in spite of this uncertainty. Here we describe the context in which we are exploring this, and some initial work to support elicitation of uncertain requirements, and to deal with the combination of such information from multiple stakeholders.

  12. Firefighter safety for PV systems: Overview of future requirements and protection systems

    DEFF Research Database (Denmark)

    Spataru, Sergiu; Sera, Dezso; Blaabjerg, Frede

    2013-01-01

    for operators during maintenance or fire-fighting. One of the solutions is individual module shutdown by short-circuiting or disconnecting each PV module from the PV string. However, currently no standards have been adopted either for implementing or testing these methods, or doing an evaluation of the module...... shutdown procedures. This paper gives an overview on the most recent fire - and firefighter safety requirements for PV systems, with focus on system and module shutdown systems. Several solutions are presented, analyzed and compared by considering a number of essential characteristics, including......An important and highly discussed safety issue for photovoltaic systems is that, as long as they are illuminated, a high voltage is present at the PV string terminals and cables between the string and inverters, independent of the state of the inverter's dc disconnection switch, which poses a risk...

  13. Use of a Graded Approach in the Application of the Safety Requirements for Research Reactors. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    The IAEA's Statute authorizes the Agency to 'establish or adopt? standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

  14. Evaluation of atmospheric dispersion/consequence models supporting safety analysis

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Lazaro, M.A.; Woodard, K.

    1996-01-01

    Two DOE Working Groups have completed evaluation of accident phenomenology and consequence methodologies used to support DOE facility safety documentation. The independent evaluations each concluded that no one computer model adequately addresses all accident and atmospheric release conditions. MACCS2, MATHEW/ADPIC, TRAC RA/HA, and COSYMA are adequate for most radiological dispersion and consequence needs. ALOHA, DEGADIS, HGSYSTEM, TSCREEN, and SLAB are recommended for chemical dispersion and consequence applications. Additional work is suggested, principally in evaluation of new models, targeting certain models for continued development, training, and establishing a Web page for guidance to safety analysts

  15. Training courses on integrated safety assessment modelling for waste repositories

    International Nuclear Information System (INIS)

    Mallants, D.

    2007-01-01

    Near-surface or deep repositories of radioactive waste are being developed and evaluated all over the world. Also, existing repositories for low- and intermediate-level waste often need to be re-evaluated to extend their license or to obtain permission for final closure. The evaluation encompasses both a technical feasibility as well as a safety analysis. The long term safety is usually demonstrated by means of performance or safety assessment. For this purpose computer models are used that calculate the migration of radionuclides from the conditioned radioactive waste, through engineered barriers to the environment (groundwater, surface water, and biosphere). Integrated safety assessment modelling addresses all relevant radionuclide pathways from source to receptor (man), using in combination various computer codes in which the most relevant physical, chemical, mechanical, or even microbiological processes are mathematically described. SCK-CEN organizes training courses in Integrated safety assessment modelling that are intended for individuals who have either a controlling or supervising role within the national radwaste agencies or regulating authorities, or for technical experts that carry out the actual post-closure safety assessment for an existing or new repository. Courses are organised by the Department of Waste and Disposal

  16. Technical Support Section Instrument Support Program for nuclear and nonnuclear facilities with safety requirements

    International Nuclear Information System (INIS)

    Adkisson, B.P.; Allison, K.L.

    1995-01-01

    This document describes requirements, procedures, and supervisory responsibilities of the Oak Ridge National Laboratory (ORNL) Instrumentation and Controls (I ampersand C) Division's Technical Support Section (TSS) for instrument surveillance and maintenance in nonreactor nuclear facilities having identified Operational Safety Requirements (OSRs) or Limiting Conditions Document (LCDs). Implementation of requirements comply with the requirements of U.S. Department of Energy (DOE) Orders 5480.5, 5480.22, and 5481.1B; Martin Marietta Energy Systems, Inc. (Energy Systems), Policy Procedure ESS-FS-201; and ORNL SPP X-ESH-15. OSRs and LCDs constitute an agreement or contract between DOE and the facility operating management regarding the safe operation of the facility. One basic difference between OSRs and LCDs is that violation of an OSR is considered a Category II occurrence, whereas violation of an LCD requirement is considered a Category III occurrence (see Energy Systems Standard ESS-OP-301 and ORNL SPP X-GP-13). OSRs are required for high- and moderate-hazard nuclear facilities, whereas the less-rigorous LCDs are required for low-hazard nuclear facilities and selected open-quotes generally acceptedclose quotes operations. Hazard classifications are determined through a hazard screening process, which each division conducts for its facilities

  17. Efficacy and safety of intravenous secukinumab in noninfectious uveitis requiring steroid-sparing immunosuppressive therapy.

    Science.gov (United States)

    Letko, Erik; Yeh, Steven; Foster, C Stephen; Pleyer, Uwe; Brigell, Mitchell; Grosskreutz, Cynthia L

    2015-05-01

    Secukinumab, a fully human anti-interleukin-17A monoclonal antibody, exhibited promising activity in a proof-of-concept study when administered in intravenous (IV) doses to patients with active, chronic, noninfectious uveitis. This study compared the efficacy and safety of different IV and subcutaneous (SC) doses of secukinumab in patients with noninfectious uveitis. Multicenter, randomized, double-masked, dose-ranging, phase 2 clinical trial. Thirty-seven patients with active noninfectious intermediate uveitis, posterior uveitis, or panuveitis who required corticosteroid-sparing immunosuppressive therapy. Patients were randomized to secukinumab 300 mg SC every 2 weeks for 4 doses, secukinumab 10 mg/kg IV every 2 weeks for 4 doses, or secukinumab 30 mg/kg IV every 4 weeks for 2 doses. Intravenous or SC saline was administered to maintain masking. Efficacy was assessed on day 57 (2-4 weeks after last dose). Percentage of patients with treatment response, defined as (1) at least a 2-grade reduction in vitreous haze score or trace or absent vitreous haze in the study eye without an increase in corticosteroid dose and without uveitis worsening or (2) reduction in corticosteroid dosages to prespecified levels without uveitis worsening. Percentage of patients with remission, defined as anterior chamber cell and vitreous haze scores of 0 or 0.5+ in both eyes without corticosteroid therapy or uveitis worsening. Secukinumab 30 mg/kg IV and 10 mg/kg IV, compared with the 300 mg SC dose, produced higher responder rates (72.7% and 61.5% vs. 33.3%, respectively) and remission rates (27.3% and 38.5% vs. 16.7%, respectively). Statistical and clinical superiority for the 30 mg/kg IV dose compared with the 300 mg SC dose was established in a Bayesian probability model. Other measures, including time to response onset, change in visual acuity, and change in vitreous haze score, showed numeric trends favoring IV dosing. Secukinumab, administered in IV or SC formulations, appeared

  18. Models for Aircrew Safety Assessment: Uses, Limitations and Requirements

    Science.gov (United States)

    1999-08-01

    altitude, the preoxygenation time, and the calorimetric or other thermometric methods. Accurate SAR level of exercise undertaken. Taking these...air- structed with an approximately 13 mm thick aluminium craft that meet the objectives of the Federal Aviation plate supported by a steel triangular

  19. Plant design and layout of the different buildings with respect to safety, operational and maintenance requirements

    International Nuclear Information System (INIS)

    Linder, C.

    1981-01-01

    Design and layout of the buildings of a nuclear power plant are governed by the safety requirements regarding nearby population as called for by government regulations as well as by operational and maintenance requirements called for by the power utilities in order to assure smooth operation and easy service conditions. The lecture will focus on the different functional circumstances to be considered, their relative importance, criteria to be applied, pertinent regulations etc. and also give examples on the solutions to the above requirements. Main topics to be covered will be those circumstances that impose the highest demands on the civil engineering layout and design: airplane impact, earthquake, loss of coolant accident, pipe whipping, fuel cask transfer, annual overhaul, leak detection etc. (orig./RW)

  20. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

  1. Evaluating Performance of Safety Management and Occupational Health Using Total Quality Safety Management Model (TQSM

    Directory of Open Access Journals (Sweden)

    E Mohammadfam

    2015-11-01

    Full Text Available Introduction: All organizations, whether public or private, necessitate performance evaluation systems in regard with growth, stability, and development in the competitive fields. One of the existing models for performance evaluation of occupational health and safety management is Total Quality Safety Management model (TQSM. Therefore, the present study aimed to evaluate performance of safety management and occupational health utilizing TQSM model. Methods: In this descriptive-analytic study, the population consisted of 16 individuals, including managers, supervisors, and members of technical protection and work health committee. Then the participants were asked to respond to TQSM questionnaire before and after the implementation of Occupational Health & Safety Advisory Services 18001 (OHSAS18001. Ultimately, the level of each program as well as the TQSM status were determined before and after the implementation of OHSAS18001. Results: The study results showed that the scores obtained by the company before OHSAS 18001’s implementation, was 43.7 out of 312. After implementing OHSAS 18001 in the company and receiving the related certificate, the total score of safety program that company could obtain was 127.12 out of 312 demonstrating a rise of 83.42 scores (26.8%. The paired t-test revealed that mean difference of TQSM scores before and after OHSAS 18001 implementation was proved to be significant (p> 0.05. Conclusion: The study findings demonstrated that TQSM can be regarded as an appropriate model in order to monitor the performance of safety management system and occupational health, since it possesses the ability to quantitatively evaluate the system performance.

  2. A goal-oriented requirements modelling language for enterprise architecture

    NARCIS (Netherlands)

    Quartel, Dick; Engelsman, W.; Jonkers, Henk; van Sinderen, Marten J.

    2009-01-01

    Methods for enterprise architecture, such as TOGAF, acknowledge the importance of requirements engineering in the development of enterprise architectures. Modelling support is needed to specify, document, communicate and reason about goals and requirements. Current modelling techniques for

  3. ON THE IMPACT OF FLIGHT SAFETY CERTIFICATION REQUIREMENTS ON THE AERODYNAMIC EFFICIENCY OF COMMERCIAL AIRPLANES

    Directory of Open Access Journals (Sweden)

    Vladimir I. Shevyakov

    2018-01-01

    Full Text Available The article considers the issue of aerodynamics efficiency implementation taking into account certification requirements for flight safety. Aerodynamics efficiency means high aerodynamic performance (depending on the airplane size, aerodynamic performance in cruise flight, high aerodynamic performance at takeoff, as well as lift performance at landing.The author estimated the impact on aerodynamics efficiency of both the requirements for aerodynamics performance and requirements for aircraft systems, noncompliance with which may result in significant change of expected operating conditions. It was shown that the use of supercritical wing profiles may result in flight mode limitations due to failure of the required buffeting capacities. It does not allow engaging all the advantages of aerodynamics layout and requires special design solutions to prevent such cases.There were reviewed certification requirements for flight level pressure altitude accuracy and icing conditions warning sysytem. The research presented the methods of aerodynamic efficiency increase by meeting the requirements for reduced vertical separation minima flights and in icing conditions, including requirements for air data probes. Reduced vertical separation minima flight requirements are met by means of efficient air data probes location. Theoretical methods of flow calculation determine areas on the airplane skin surface where static probes minimize errors depending on angle-of-attack and sideslip. It was shown that if certification requirements are not met and in case of flight out of reduced vertical separation minima area, aerodynamics efficiency is significantly reduced and fuel consumption can be increased by 10% and higher. Suggested approaches implementation allows increasing commercial airplanes competitiveness.

  4. Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

    CERN Document Server

    Marshall, T; Topilski, L; Merrill, B

    2002-01-01

    For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

  5. 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

  6. Modeling of passengers' safety perception for buses on mountainous roads.

    Science.gov (United States)

    Khoo, Hooi Ling; Ahmed, Muaid

    2018-04-01

    This study had developed a passenger safety perception model specifically for buses taking into consideration the various factors, namely driver characteristics, environmental conditions, and bus characteristics using Bayesian Network. The behaviour of bus driver is observed through the bus motion profile, measured in longitudinal, lateral, and vertical accelerations. The road geometry is recorded using GPS and is computed with the aid of the Google map while the perceived bus safety is rated by the passengers in the bus in real time. A total of 13 variables were derived and used in the model development. The developed Bayesian Network model shows that the type of bus and the experience of the driver on the investigated route could have an influence on passenger's perception of their safety on buses. Road geometry is an indirect influencing factor through the driver's behavior. The findings of this model are useful for the authorities to structure an effective strategy to improve the level of perceived bus safety. A high level of bus safety will definitely boost passenger usage confidence which will subsequently increase ridership. Copyright © 2018 Elsevier Ltd. All rights reserved.

  7. Biphasic poly-dispersions modelling with applications in nuclear safety

    International Nuclear Information System (INIS)

    Gimenez, Marcelo

    2005-01-01

    Many technological, scientific and environmental areas require of the understanding of multiple-phase, polydisperse systems.Particularly, in nuclear safety area, under hypothetical severe accident conditions, the fission products are released as gases and airborne particles.In the present thesis a contribution to the modeling of confined polydisperse systems (aerosols and liquid-gas) is presented. A family of one and two dimensional models was developed, describing the impact of the particles size on the aerosol evolution, taking into account the transport mechanisms by the carrying media, settling, lift, thermophoresis, diffusion, coagulation and condensation, considering the spatial dependence. The aerosol general dynamic balance equation -no-linear integral-differential equation-, is treated by means of the Moments Method, imposing a prescribed volume size distribution, and obtaining a model based in the first few moments.This yields to a relatively simple convection-diffusion set of coupled equations, that describes the evolution of the parameters that characterize the size distribution, with a low computational cost that allows to deal the heterogeneities of the spatial distribution of the aerosol.A complementary model, based also on the Moments Method, is developed to perform sensitivity analysis due to uncertainties in the constitutive equations and in the input parameters.The Perturbative Method, Differential-formalism, is used to develop a set of sensitivity equations, which are useful for designing experiments and theoretical studies.The model results are compared with exact solutions and numerical and experimental data extracted from the open literature.Various physical phenomena involved in different simulated case are analysed in detail.Particularly, the study is oriented to the analysis of the concentration gradients and the validation of the aerosol well-mixed hypothesis, typically used in present numerical codes.Deviations were observed from the

  8. Preparation of safety regulatory requirements for new technology like digital system

    International Nuclear Information System (INIS)

    Ito, Juichiro; Takita, Masami

    2011-01-01

    The current regulatory requirements on digital instrumentation and control system have been reviewed by JNES, considering international trend discussed in DICWG (Digital Instrumentation and Control Working Group) of MDEP (Multinational Design Evaluation Program). MDEP DICWG held in OECD/NEA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency) gives the opportunity to identify the convergence of applicable standards. The working group's activities include: identifying and prioritising the member countries' challenges, practices, and needs regarding standards and regulatory guidance regarding digital instrumentation and control; identifying areas of importance and needs for convergence of existing standards and guidance or development of new standards; sharing of information; and identifying common positions among the member countries for areas of particular importance and need. The DICWG drafted common positions on specific issues which are based on the existing standards, national regulatory guidance, best practices, and group inputs using an agreed upon process and framework. Five general common positions are under discussion in this fiscal year. Simplicity in Design, Software Common Cause Failures, Software Tools, Data communication, Verification and Validation throughout the life cycle of safety systems using digital computers. In addition, the technical evaluation of standards of the Japan Electric Association about digital system for safety was made to support NISA (Nuclear and Industrial Safety Agency). (author)

  9. Biosphere models for safety assesment of radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Proehl, G; Olyslaegers, G; Zeevaert, T [SCK/CEN, Mol (Belgium); Kanyar, B [University of Veszprem (Hungary). Dept. of Radiochemistry; Pinedo, P; Simon, I [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain); Bergstroem, U; Hallberg, B [Studsvik Ecosafe, Nykoeping (Sweden); Mobbs, S; Chen, Q; Kowe, R [NRPB, Chilton, Didcot (United Kingdom)

    2004-07-01

    The aim of the BioMoSA project has been to contribute in the confidence building of biosphere models, for application in performance assessments of radioactive waste disposal. The detailed objectives of this project are: development and test of practical biosphere models for application in long-term safety studies of radioactive waste disposal to different European locations, identification of features, events and processes that need to be modelled on a site-specific rather than on a generic base, comparison of the results and quantification of the variability of site-specific models developed according to the reference biosphere methodology, development of a generic biosphere tool for application in long term safety studies, comparison of results from site-specific models to those from generic one, Identification of possibilities and limitations for the application of the generic biosphere model. (orig.)

  10. Biosphere models for safety assessment of radioactive waste disposal

    International Nuclear Information System (INIS)

    Proehl, G.; Olyslaegers, G.; Zeevaert, T.; Kanyar, B.; Bergstroem, U.; Hallberg, B.; Mobbs, S.; Chen, Q.; Kowe, R.

    2004-01-01

    The aim of the BioMoSA project has been to contribute in the confidence building of biosphere models, for application in performance assessments of radioactive waste disposal. The detailed objectives of this project are: development and test of practical biosphere models for application in long-term safety studies of radioactive waste disposal to different European locations, identification of features, events and processes that need to be modelled on a site-specific rather than on a generic base, comparison of the results and quantification of the variability of site-specific models developed according to the reference biosphere methodology, development of a generic biosphere tool for application in long term safety studies, comparison of results from site-specific models to those from generic one, Identification of possibilities and limitations for the application of the generic biosphere model. (orig.)

  11. Relationship between general safety requirements and safety culture in the improvement of safe operation of I.N.R. TRIGA reactor facilities

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Preda, M.; Chiritescu, M.; Dumitru, M.

    1996-01-01

    Acquiring of the basic principles of ''safety culture'' by a large number of profesionals in the nuclear field drew the attention of the decision factors in the INR managerial structure, who decided to promote certain practical actions at each level in order to improve nuclear safety. Starting from the ''Republican Standards for Nuclear Safety'' issued by CSEN in 1975, where general safety criteria are defined for nuclear reactors and NPPs, the specialists at the TRIGA reactor originated and implemented a coherent and secure system to ensure nuclear safety over all steps of nuclear activities: research, conception, execution, commissioning and operation. This system has been continuosly corrected so that now it is completely integrated in a modern safety system. The paper presents the way in which a modern system for nuclear safety at the TRIGA reactor has been implemented and developed, in accordance to specific criteria and requirements imposed by related National Regulations and with the principles of safety culture. Starting from the definition of specific responsabilities, there are presented the internal stipulations and practical actions at all levels in order to enhance nuclear safety. (orig.)

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

    International Nuclear Information System (INIS)

    Liu Sujuan

    2002-01-01

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

  13. Considerations in the development of safety requirements for innovative reactors: Application to modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    2003-08-01

    Member States of the IAEA have frequently requested this organization to assess, at the conceptual stage, the safety of the design of nuclear reactors that rely on a variety of technologies and are of a high degree of innovation. However, to date, for advanced and innovative reactors and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist. This TECDOC is an outcome of the efforts deployed by the IAEA to develop a general approach for assessing the safety of the design of advanced and innovative reactors, and of all reactors in general including research reactors, with characteristics that differ from those of light water reactors. This publication puts forward a method for safety assessment that is based on the well established and accepted principle of defence in depth. The need to develop a general approach for assessing the safety of the design of reactors that applies to all kinds of advanced reactors was emphasized by the request to the IAEA by South Africa to review the safety of the South African pebble bed modular reactor. This reactor, as other modular high temperature gas cooled reactors (MHTGRs), adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors. In this TECDOC, the MHTGR has been selected as a case study to demonstrate the viability of the method proposed. The approach presented is based on an extended interpretation of the concept of defence in depth and its link with the general safety objectives and fundamental safety functions as set out in 'Safety of Nuclear Power Plants: Design', IAEA Safety Standards No. NS-R.1, issued by the IAEA in 2000. The objective

  14. Safety requirements of the BMU to be met in final storage of heat-producing waste: An evaluation

    International Nuclear Information System (INIS)

    Thomauske, Bruno

    2009-01-01

    On August 12, 2008, The German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) published a draft of July 29, 2008 of the ''Safety Requirements to Be Met in Final Storage of Heat-producing Radioactive Waste.'' As announced by the BMU, these safety requirements are to bring up to the state of the art the safety criteria of 1983. Over a couple of years, efforts had been made to adapt the criteria to the internationally accepted standard as demanded by the Advisory Committees on Reactor Safeguards (RSK) and Radiation Protection (SSK). There is no waste management concept underlying the safety requirements. As a consequence, the draft should be withdrawn by the Federal Ministry for the Environment and replaced by a version revised from scratch and offering assured quality. (orig./GL)

  15. Requirements model generation to support requirements elicitation: The Secure Tropos experience

    NARCIS (Netherlands)

    Kiyavitskaya, N.; Zannone, N.

    2008-01-01

    In recent years several efforts have been devoted by researchers in the Requirements Engineering community to the development of methodologies for supporting designers during requirements elicitation, modeling, and analysis. However, these methodologies often lack tool support to facilitate their

  16. A strategy to establish Food Safety Model Repositories.

    Science.gov (United States)

    Plaza-Rodríguez, C; Thoens, C; Falenski, A; Weiser, A A; Appel, B; Kaesbohrer, A; Filter, M

    2015-07-02

    Transferring the knowledge of predictive microbiology into real world food manufacturing applications is still a major challenge for the whole food safety modelling community. To facilitate this process, a strategy for creating open, community driven and web-based predictive microbial model repositories is proposed. These collaborative model resources could significantly improve the transfer of knowledge from research into commercial and governmental applications and also increase efficiency, transparency and usability of predictive models. To demonstrate the feasibility, predictive models of Salmonella in beef previously published in the scientific literature were re-implemented using an open source software tool called PMM-Lab. The models were made publicly available in a Food Safety Model Repository within the OpenML for Predictive Modelling in Food community project. Three different approaches were used to create new models in the model repositories: (1) all information relevant for model re-implementation is available in a scientific publication, (2) model parameters can be imported from tabular parameter collections and (3) models have to be generated from experimental data or primary model parameters. All three approaches were demonstrated in the paper. The sample Food Safety Model Repository is available via: http://sourceforge.net/projects/microbialmodelingexchange/files/models and the PMM-Lab software can be downloaded from http://sourceforge.net/projects/pmmlab/. This work also illustrates that a standardized information exchange format for predictive microbial models, as the key component of this strategy, could be established by adoption of resources from the Systems Biology domain. Copyright © 2015. Published by Elsevier B.V.

  17. Modeling crash injury severity by road feature to improve safety.

    Science.gov (United States)

    Penmetsa, Praveena; Pulugurtha, Srinivas S

    2018-01-02

    The objective of this research is 2-fold: to (a) model and identify critical road features (or locations) based on crash injury severity and compare it with crash frequency and (b) model and identify drivers who are more likely to contribute to crashes by road feature. Crash data from 2011 to 2013 were obtained from the Highway Safety Information System (HSIS) for the state of North Carolina. Twenty-three different road features were considered, analyzed, and compared with each other as well as no road feature. A multinomial logit (MNL) model was developed and odds ratios were estimated to investigate the effect of road features on crash injury severity. Among the many road features, underpass, end or beginning of a divided highway, and on-ramp terminal on crossroad are the top 3 critical road features. Intersection crashes are frequent but are not highly likely to result in severe injuries compared to critical road features. Roundabouts are least likely to result in both severe and moderate injuries. Female drivers are more likely to be involved in crashes at intersections (4-way and T) compared to male drivers. Adult drivers are more likely to be involved in crashes at underpasses. Older drivers are 1.6 times more likely to be involved in a crash at the end or beginning of a divided highway. The findings from this research help to identify critical road features that need to be given priority. As an example, additional advanced warning signs and providing enlarged or highly retroreflective signs that grab the attention of older drivers may help in making locations such as end or beginning of a divided highway much safer. Educating drivers about the necessary skill sets required at critical road features in addition to engineering solutions may further help them adopt safe driving behaviors on the road.

  18. Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, N.; Kawatta, N.; Niwa, H.; Kondo, S.; Maeda, K

    1996-12-31

    This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given.

  19. New safety and security requirements for the transport of nuclear and other radioactive materials in Hungary

    International Nuclear Information System (INIS)

    Katona, T.; Horvath, K.; Safar, J.

    2016-01-01

    In addition to the promulgation of mode-specific regulations of international transport of dangerous goods, some Hungarian governmental and ministerial decrees impose further conditions upon the transport of nuclear and other radioactive materials. One of these ministerial decrees on the transport, carriage and packaging of radioactive materials is under revision and it will require • approval of emergency response plan (including security and safety contingency plan); • report on transport incidents and accidents for classifying them in accordance with the INES scale; • the competent authority to request experts’ support for the approval of package designs, radioactive material designs and shipments. Regarding the security of the transport of nuclear and other radioactive materials a new Hungarian governmental decree and a related guidance are about to be published which will supply additional requirements in the field of the transport security especially concerning radioactive materials, implementing - among others - IAEA recommendations of the NSS No9 and No14. The main and relevant features of the Hungarian nuclear regulatory system and the details of both new decrees regarding the safety and security issues of transport of nuclear and other radioactive materials will be discussed. (author)

  20. Operational characteristics of nuclear power plants - modelling of operational safety

    International Nuclear Information System (INIS)

    Studovic, M.

    1984-01-01

    By operational experience of nuclear power plants and realize dlevel of availability of plant, systems and componenst reliabiliuty, operational safety and public protection, as a source on nature of distrurbances in power plant systems and lessons drawn by the TMI-2, in th epaper are discussed: examination of design safety for ultimate ensuring of safe operational conditions of the nuclear power plant; significance of the adequate action for keeping proess parameters in prescribed limits and reactor cooling rquirements; developed systems for measurements detection and monitoring all critical parameters in the nuclear steam supply system; contents of theoretical investigation and mathematical modeling of the physical phenomena and process in nuclear power plant system and components as software, supporting for ensuring of operational safety and new access in staff education process; program and progress of the investigation of some physical phenomena and mathematical modeling of nuclear plant transients, prepared at faculty of mechanical Engineering in Belgrade. (author)

  1. DECISION MAKING MODELING OF CONCRETE REQUIREMENTS

    Directory of Open Access Journals (Sweden)

    Suhartono Irawan

    2001-01-01

    Full Text Available This paper presents the results of an experimental evaluation between predicted and practice concrete strength. The scope of the evaluation is the optimisation of the cement content for different concrete grades as a result of bringing the target mean value of tests cubes closer to the required characteristic strength value by reducing the standard deviation. Abstract in Bahasa Indonesia : concrete+mix+design%2C+acceptance+control%2C+optimisation%2C+cement+content.

  2. 46 CFR 53.05-1 - Safety valve requirements for steam boilers (modifies HG-400 and HG-401).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Safety valve requirements for steam boilers (modifies HG-400 and HG-401). 53.05-1 Section 53.05-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY... requirements for steam boilers (modifies HG-400 and HG-401). (a) The pressure relief valve requirements and the...

  3. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (French Edition); Cadre gouvernemental, legislatif et reglementaire de la surete. Prescriptions generales de surete. Partie 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-11-15

    The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

  4. Governmental, Legal and Regulatory Framework for Safety. General Safety Requirements. Part 1 (Spanish Edition); Marco gubernamental, juridico y regulador para la seguridad. Requisitos de Seguridad Generales. Parte 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-11-15

    The objective of this publication is to establish requirements in respect of the governmental, legal and regulatory framework for safety. It covers the essential aspects of the framework for establishing a regulatory body and taking other actions necessary to ensure the effective regulatory control of facilities and activities utilized for peaceful purposes. Other responsibilities and functions, such as liaison within the global safety regime and on support services for safety (including radiation protection), emergency preparedness and response, nuclear security, and the State system of accounting for and control of nuclear material, are also covered.

  5. Safety analysis report on Model UC-609 shipping package

    International Nuclear Information System (INIS)

    Sandberg, R.R.

    1977-08-01

    This Safety Analysis Report for Packaging demonstrates that model UC-609 shipping package can safely transport tritium in any of its forms. The package and its contents are described. The package when subjected to the transport conditions specified in the Code of Federal Regulations, Title 10, Part 71 is evaluated. Finally, compliance with these regulations is discussed

  6. Review of SFR Design Safety using Preliminary Regulatory PSA Model

    International Nuclear Information System (INIS)

    Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

    2013-01-01

    The major objective of this research is to develop a risk model for regulatory verification of the SFR design, and thereby, make sure that the SFR design is adequate from a risk perspective. In this paper, the development result of preliminary regulatory PSA model of SFR is discussed. In this paper, development and quantification result of preliminary regulatory PSA model of SFR is discussed. It was confirmed that the importance PDRC and ADRC dampers is significant as stated in the result of KAERI PSA model. However, the importance can be changed significantly depending on assumption of CCCG and CCF factor of PDRC and ADRC dampers. SFR (sodium-cooled fast reactor) which is Gen-IV nuclear energy system, is designed to accord with the concept of stability, sustainability and proliferation resistance. KALIMER-600, which is under development in Korea, includes passive safety systems (e. g. passive reactor shutdown, passive residual heat removal, and etc.) as well as active safety systems. Risk analysis from a regulatory perspective is needed to support the regulatory body in its safety and licensing review for SFR (KALIMER-600). Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and delay of the SFR licensing schedule that may be caused otherwise

  7. The Swiss cheese model of safety incidents: are there holes in the metaphor?

    Directory of Open Access Journals (Sweden)

    Perneger Thomas V

    2005-11-01

    Full Text Available Abstract Background Reason's Swiss cheese model has become the dominant paradigm for analysing medical errors and patient safety incidents. The aim of this study was to determine if the components of the model are understood in the same way by quality and safety professionals. Methods Survey of a volunteer sample of persons who claimed familiarity with the model, recruited at a conference on quality in health care, and on the internet through quality-related websites. The questionnaire proposed several interpretations of components of the Swiss cheese model: a slice of cheese, b hole, c arrow, d active error, e how to make the system safer. Eleven interpretations were compatible with this author's interpretation of the model, 12 were not. Results Eighty five respondents stated that they were very or quite familiar with the model. They gave on average 15.3 (SD 2.3, range 10 to 21 "correct" answers out of 23 (66.5% – significantly more than 11.5 "correct" answers that would expected by chance (p Conclusion The interpretations of specific features of the Swiss cheese model varied considerably among quality and safety professionals. Reaching consensus about concepts of patient safety requires further work.

  8. Experimental and theoretical requirements for fuel modelling

    International Nuclear Information System (INIS)

    Gatesoupe, J.P.

    1979-01-01

    From a scientific point of view it may be considered that any event in the life of a fuel pin under irradiation should be perfectly well understood and foreseen from that deterministic point of view, the whole behaviour of the pin maybe analysed and dismantled with a specific function for every component part and each component part related to one basic phenomenon which can be independently studied on pure physical grounds. When extracted from the code structure the subroutine is studied for itself by specialists who try to keep as close as possible to the physics involved in the phenomenon; that often leads to an impressive luxury in details and a subsequent need for many unavailable input data. It might seem more secure to follow that approach since it tries to be firmly based on theoretical grounds. One should think so if the phenomenological situation in the pin were less complex than it is. The codes would not be adequate for off-normal operating conditions since for the accidental transient conditions the key-phenomena would not be the same as for steady-state or slow transient conditions. The orientation given to fuel modelling is based on our two main technological constraints which are: no fuel melting; no cladding failure; no excessive cladding deformation. In this context, the only relevant models are those which have a significant influence on the maximum temperatures in the fuel or on the cladding damage hence the selection between key models and irrelevant models which will next be done. A rather pragmatic view is kept on codification with a special focus on a few determinant aspects of fuel behaviour and no attention to models which are nothing but decorative. Fuel modeling is merely considered as a link between experimental knowledge; it serves as a guide for further improvements in fuel design and as so happens to be quite useful. On this basis the main lacks in of fuel behaviour is described. These are mainly concerning: thermal transfer through

  9. Modelling and data prerequisites for specific applications of PSA in the management of nuclear plant safety

    International Nuclear Information System (INIS)

    1994-04-01

    The IAEA has a programme which supports the performance and use of probabilistic safety assessments (PSAS) to improve nuclear safety internationally. The assistance offered in this areas by the IAEA to Member States has traditionally focused on planning, performance and peer review of PSAs. PSA activities within the IAEA's programme in the area of applications are presently being expanded. The various applications of PSAs require that PSAs being developed have certain characteristics in terms of their scope, the degree of details in the modelling, the flexibility in performing desired calculations, the quality and type of the data used, and the assumptions made in treating safety significant aspects. In many cases, existing PSAs or PSAs being completed can be extended to fulfill the requirements for uses in many applications to enhance the safety of nuclear power plants. This report provides information on how to carry such extensions by matching PSA characteristics to various applications that are being considered. This report was prepared by consultants together with the IAEA following the recommendations of a Technical Committee Meeting on PSA Requirements for Use in Safety Management, held by the IAEA in co-operation with the Swedish Nuclear Power Inspectorate in Stockholm, Sweden, 16-20 September 1991. 42 refs, 1 tab

  10. Use of FPGA and CPLD in nuclear reactor safety systems and its regulatory review requirements for reactor safety

    International Nuclear Information System (INIS)

    Roy, Suvadip; Biswas, Animesh; Pradhan, S.K.

    2015-01-01

    Field Programmable Gate Arrays (FPGA) and Complex Programmable Logic Devices (CPLD) is being used widely in safety critical and safety related systems in nuclear power plans like in trip logic units, Engineered Safety Feature (ESF) actuation decision logic and neutronic signal processing for their reprogrammability feature and compact design. These HDL Programmable devices (HPD) are complex devices consisting of both hardware and software which is used to implement the logic on the FPGA. It is observed that these Programmable devices suffer from various modes of failure and the major failures in these devices are due to Single Event Upset (SEU), where a highly energetic ionizing radiation may lead to device failure which can even occur in radiologically benign environment. Other failures can occur during steps of developing the hardware using software tools like during Synthesis and placement and routing of the desired hardware. Here a study on use of such devices in Nuclear Reactors, study on mode of failures of these devices, way to tackle such failure and development of review guidelines for review of such devices used in safety critical and safety related systems with special emphasis on choice of software tools, way to mitigate effects of SEU and simulation and hardware testing results to be reviewed by regulatory body during design safety review is done. (author)

  11. Technical Safety Requirements for the B695 Segment of the Decontamination and Waste Treatment Facility

    International Nuclear Information System (INIS)

    Larson, H L

    2007-01-01

    This document contains Technical Safety Requirements (TSRs) for the Radioactive and Hazardous Waste Management (RHWM) Division's B695 Segment of the Decontamination and Waste Treatment Facility (DWTF) at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the B695 Segment of the DWTF. The TSRs are derived from the Documented Safety Analysis (DSA) for the B695 Segment of the DWTF (LLNL 2004). The analysis presented there determined that the B695 Segment of the DWTF is a low-chemical hazard, Hazard Category 3, nonreactor nuclear facility. The TSRs consist primarily of inventory limits as well as controls to preserve the underlying assumptions in the hazard analyses. Furthermore, appropriate commitments to safety programs are presented in the administrative controls section of the TSRs. The B695 Segment of the DWTF (B695 and the west portion of B696) is a waste treatment and storage facility located in the northeast quadrant of the LLNL main site. The approximate area and boundary of the B695 Segment of the DWTF are shown in the B695 Segment of the DWTF DSA. Activities typically conducted in the B695 Segment of the DWTF include container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. B695 is used to store and treat radioactive, mixed, and hazardous waste, and it also contains equipment used in conjunction with waste processing operations to treat various liquid and solid wastes. The portion of the building called Building 696 Solid Waste Processing Area (SWPA), also referred to as B696S in this report, is used primarily to manage solid radioactive waste. Operations specific to the SWPA include sorting and segregating low-level waste (LLW) and transuranic (TRU) waste, lab-packing, sampling, and crushing empty drums that previously contained LLW. A permit modification for B696S was submitted to DTSC in January 2004 to store and treat hazardous and mixed

  12. Technical Safety Requirements for the B695 Segment of the Decontamination and Waste Treatment Facility

    Energy Technology Data Exchange (ETDEWEB)

    Larson, H L

    2007-09-07

    This document contains Technical Safety Requirements (TSRs) for the Radioactive and Hazardous Waste Management (RHWM) Division's B695 Segment of the Decontamination and Waste Treatment Facility (DWTF) at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the B695 Segment of the DWTF. The TSRs are derived from the Documented Safety Analysis (DSA) for the B695 Segment of the DWTF (LLNL 2004). The analysis presented there determined that the B695 Segment of the DWTF is a low-chemical hazard, Hazard Category 3, nonreactor nuclear facility. The TSRs consist primarily of inventory limits as well as controls to preserve the underlying assumptions in the hazard analyses. Furthermore, appropriate commitments to safety programs are presented in the administrative controls section of the TSRs. The B695 Segment of the DWTF (B695 and the west portion of B696) is a waste treatment and storage facility located in the northeast quadrant of the LLNL main site. The approximate area and boundary of the B695 Segment of the DWTF are shown in the B695 Segment of the DWTF DSA. Activities typically conducted in the B695 Segment of the DWTF include container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. B695 is used to store and treat radioactive, mixed, and hazardous waste, and it also contains equipment used in conjunction with waste processing operations to treat various liquid and solid wastes. The portion of the building called Building 696 Solid Waste Processing Area (SWPA), also referred to as B696S in this report, is used primarily to manage solid radioactive waste. Operations specific to the SWPA include sorting and segregating low-level waste (LLW) and transuranic (TRU) waste, lab-packing, sampling, and crushing empty drums that previously contained LLW. A permit modification for B696S was submitted to DTSC in January 2004 to store and treat hazardous and

  13. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    D. Kokkinos

    2005-01-01

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  14. 30 CFR 250.806 - Safety and pollution prevention equipment quality assurance requirements.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 2 2010-07-01 2010-07-01 false Safety and pollution prevention equipment... Gas Production Safety Systems § 250.806 Safety and pollution prevention equipment quality assurance... install only certified safety and pollution prevention equipment (SPPE) in wells located on the OCS. SPPE...

  15. FMCSA safety program effectiveness measurement : carrier intervention effectiveness model, version 1.0 : [analysis brief].

    Science.gov (United States)

    2015-01-01

    The Carrier Intervention Effectiveness Model (CIEM) : provides the Federal Motor Carrier Safety : Administration (FMCSA) with a tool for measuring : the safety benefits of carrier interventions conducted : under the Compliance, Safety, Accountability...

  16. Implications of safety requirements for the treatment of THMC processes in geological disposal systems for radioactive waste

    Directory of Open Access Journals (Sweden)

    Frédéric Bernier

    2017-06-01

    Full Text Available The mission of nuclear safety authorities in national radioactive waste disposal programmes is to ensure that people and the environment are protected against the hazards of ionising radiations emitted by the waste. It implies the establishment of safety requirements and the oversight of the activities of the waste management organisation in charge of implementing the programme. In Belgium, the safety requirements for geological disposal rest on the following principles: defence-in-depth, demonstrability and the radiation protection principles elaborated by the International Commission on Radiological Protection (ICRP. Applying these principles requires notably an appropriate identification and characterisation of the processes upon which the safety functions fulfilled by the disposal system rely and of the processes that may affect the system performance. Therefore, research and development (R&D on safety-relevant thermo-hydro-mechanical-chemical (THMC issues is important to build confidence in the safety assessment. This paper points out the key THMC processes that might influence radionuclide transport in a disposal system and its surrounding environment, considering the dynamic nature of these processes. Their nature and significance are expected to change according to prevailing internal and external conditions, which evolve from the repository construction phase to the whole heating–cooling cycle of decaying waste after closure. As these processes have a potential impact on safety, it is essential to identify and to understand them properly when developing a disposal concept to ensure compliance with relevant safety requirements. In particular, the investigation of THMC processes is needed to manage uncertainties. This includes the identification and characterisation of uncertainties as well as for the understanding of their safety-relevance. R&D may also be necessary to reduce uncertainties of which the magnitude does not allow

  17. Spreadsheet Decision Support Model for Training Exercise Material Requirements Planning

    National Research Council Canada - National Science Library

    Tringali, Arthur

    1997-01-01

    ... associated with military training exercises. The model combines the business practice of Material Requirements Planning and the commercial spreadsheet software capabilities of Lotus 1-2-3 to calculate the requirements for food, consumable...

  18. Extending enterprise architecture modelling with business goals and requirements

    NARCIS (Netherlands)

    Engelsman, W.; Quartel, Dick; Jonkers, Henk; van Sinderen, Marten J.

    The methods for enterprise architecture (EA), such as The Open Group Architecture Framework, acknowledge the importance of requirements modelling in the development of EAs. Modelling support is needed to specify, document, communicate and reason about goals and requirements. The current modelling

  19. Evaluating Models of Human Performance: Safety-Critical Systems Applications

    Science.gov (United States)

    Feary, Michael S.

    2012-01-01

    This presentation is part of panel discussion on Evaluating Models of Human Performance. The purpose of this panel is to discuss the increasing use of models in the world today and specifically focus on how to describe and evaluate models of human performance. My presentation will focus on discussions of generating distributions of performance, and the evaluation of different strategies for humans performing tasks with mixed initiative (Human-Automation) systems. I will also discuss issues with how to provide Human Performance modeling data to support decisions on acceptability and tradeoffs in the design of safety critical systems. I will conclude with challenges for the future.

  20. Radiation safety requirements for radioactive waste management in the framework of a quality management system

    International Nuclear Information System (INIS)

    Salgado, M.M.; Benitez, J.C.; Pernas, R.; Gonzalez, N.

    2007-01-01

    The Center for Radiation Protection and Hygiene (CPHR) is the institution responsible for the management of radioactive wastes generated from nuclear applications in medicine, industry and research in Cuba. Radioactive Waste Management Service is provided at a national level and it includes the collection and transportation of radioactive wastes to the Centralized Waste Management Facilities, where they are characterized, segregated, treated, conditioned and stored. A Quality Management System, according to the ISO 9001 Standard has been implemented for the RWM Service at CPHR. The Management System includes the radiation safety requirements established for RWM in national regulations and in the Licence's conditions. The role of the Regulatory Body and the Radiation Protection Officer in the Quality Management System, the authorization of practices, training and personal qualification, record keeping, inspections of the Regulatory Body and internal inspection of the Radiation Protection Officer, among other aspects, are described in this paper. The Quality Management System has shown to be an efficient tool to demonstrate that adequate measures are in place to ensure the safety in radioactive waste management activities and their continual improvement. (authors)

  1. Preparation of safety regulatory requirements for new technology like digital system

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The current regulatory requirements on digital instrumentation and control system have been reviewed by JNES, considering international trend discussed in DICWG of MDEP. MDEP DICWG held in OECD/NEA gives the opportunity to identify the convergence of applicable standards. The working group's activities include: identifying and prioritising the member countries' challenges, practices, and needs regarding standards and regulatory guidance on digital instrumentation and control; identifying areas of importance and needs for convergence of existing standards and guidance or development of new standards; sharing of information; and identifying common positions among the member countries for areas of particular importance and need. The DICWG drafted common positions on specific issues which are based on the existing standards, national regulatory guidance, best practices, and group inputs using an agreed process and framework. The following four general common positions have been discussed in this fiscal year. The Treatment of Common Cause Failure Resulting from Software within Digital Safety Systems, The Treatment of Hardware Description Language(HDL) Programmed Devices for Use in Nuclear Safety System, Factory Acceptance Test and Site Acceptance Test, The Use of Automatic Tests to Perform Surveilance for Digital Systems. (author)

  2. Framework conditions and requirements to ensure the technical functional safety of reprocessed medical devices.

    Science.gov (United States)

    Kraft, Marc

    2008-09-03

    Testing and restoring technical-functional safety is an essential part of medical device reprocessing. Technical functional tests have to be carried out on the medical device in the course of the validation of reprocessing procedures. These ensure (in addition to the hygiene tests) that the reprocessing procedure is suitable for the medical device. Functional tests are, however, also a part of reprocessing procedures. As a stage in the reprocessing, they ensure for the individual medical device that no damage or other changes limit the performance. When determining which technical-functional tests are to be carried out, the current technological standard has to be taken into account in the form of product-specific and process-oriented norms. Product-specific norms primarily define safety-relevant requirements. The risk management method described in DIN EN ISO 14971 is the basis for recognising hazards; the likelihood of such hazards arising can be minimised through additional technical-functional tests, which may not yet have been standardised. Risk management is part of a quality management system, which must be bindingly certified for manufacturers and processors of critical medical devices with particularly high processing demands by a body accredited by the competent authority.

  3. Preparation of safety regulatory requirements for new technology like digital system

    International Nuclear Information System (INIS)

    2012-01-01

    The current regulatory requirements on digital instrumentation and control system have been reviewed by JNES, considering international trend discussed in DICWG of MDEP. MDEP DICWG held in OECD/NEA gives the opportunity to identify the convergence of applicable standards. The working group's activities include: identifying and prioritising the member countries' challenges, practices, and needs regarding standards and regulatory guidance on digital instrumentation and control; identifying areas of importance and needs for convergence of existing standards and guidance or development of new standards; sharing of information; and identifying common positions among the member countries for areas of particular importance and need. The DICWG drafted common positions on specific issues which are based on the existing standards, national regulatory guidance, best practices, and group inputs using an agreed process and framework. The following two general common positions are discussed and to be issued in this fiscal year. Verification and Validation throughout the life cycle of safety systems using digital computers. The Impact of Cyber Security Features on Digital I and C Safety Systems. (author)

  4. A modeling approach to support safety assurance in the automotive domain

    NARCIS (Netherlands)

    Luo, Y.; Brand, van den M.G.J.; Engelen, L.J.P.; Klabbers, M.D.; Selvaraj, H.; Zydek, D.; Chmaj, G.

    2015-01-01

    As safety standards are widely used in safety-critical domains, such as ISO 26262 in the automotive domain, the use of safety cases to demonstrate product safety is stimulated. It is crucial to ensure that a safety case is both correct and clear. To support this, we proposed to make use of modeling

  5. Insights on in vitro models for safety and toxicity assessment of cosmetic ingredients.

    Science.gov (United States)

    Almeida, Andreia; Sarmento, Bruno; Rodrigues, Francisca

    2017-03-15

    According to the current European legislation, the safety assessment of each individual cosmetic ingredient of any formulation is the basis for the safety evaluation of a cosmetic product. Also, animal testing in the European Union is prohibited for cosmetic ingredients and products since 2004 and 2009, respectively. Additionally, the commercialization of any cosmetic products containing ingredients tested on animal models was forbidden in 2009. In consequence of these boundaries, the European Centre for the Validation of Alternative Methods (ECVAM) proposes a list of validated cell-based in vitro models for predicting the safety and toxicity of cosmetic ingredients. These models have been demonstrated as valuable and effective tools to overcome the limitations of animal in vivo studies. Although the use of in vitro cell-based models for the evaluation of absorption and permeability of cosmetic ingredients is widespread, a detailed study on the properties of these platforms and the in vitro-in vivo correlation compared with human data are required. Moreover, additional efforts must be taken to develop in vitro models to predict carcinogenicity, repeat dose toxicity and reproductive toxicity, for which no alternative in vitro methods are currently available. This review paper summarizes and characterizes the most relevant in vitro models validated by ECVAM employed to predict the safety and toxicology of cosmetic ingredients. Copyright © 2017 Elsevier B.V. All rights reserved.

  6. Developments of radiation safety requirements for the management of radiation devices

    International Nuclear Information System (INIS)

    Lee, Hee Seock; Choi, Jin Ho; Cheong, Yuon Young

    2002-03-01

    The approach of the risk-informed regulatory options was studied to develop the radiation safety requirements for the managements for radiation devices. The task analysis, exposure, accident scenario development, risk analysis, and systematic approach for regulatory options was considered in full, based on the NRC report, 'NUREG/CR-6642', and the translation of its core part was conducted for ongoing research. In this methodology, the diamond tree that includes human factors, etc, additionally with normal event tree, was used. According to the analysis results of this approach, the risk analysis and the development of regulatory options were applied for the electron linear accelerators and the qualitative results were obtained. Because the field user groups were participated in this study could contribute to the basis establishment of the risk-informed regulation policy through securing consensus and inducing particle interests. It will make an important role of establishing the detail plan of ongoing research

  7. Utilization of the safety functional analysis techniques to optimize the separation requirements in case of fire

    International Nuclear Information System (INIS)

    Alvarez, L.M.

    1983-01-01

    The present philosophy for the fire protection of the safe shutdown capability in nuclear power plants is based on the separation of the safety-related systems in different fire areas in such a way that the redundant systems are not subject to damage from a single fire risk. The purpose ofthis paper is to show the experience gained in the application of a symmetric method of analysis to minimize the number of fire barriers being compatible with the regulatory requirements and with capability of achieving and maintaining the safe plant shutdown in the event of a fire. As a conclusion of the analysis, the separation criteria for the divisions involved in the safe plant shutdown are obtained

  8. Developments of radiation safety requirements for the management of radiation devices

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee Seock [Pohang Accelerator Lab, Pohang (Korea, Republic of); Choi, Jin Ho [Gachun University of Medicine and science, Incheon (Korea, Republic of); Cheong, Yuon Young [Asan Medical Center, Seoul (Korea, Republic of)] (and others)

    2002-03-15

    The approach of the risk-informed regulatory options was studied to develop the radiation safety requirements for the managements for radiation devices. The task analysis, exposure, accident scenario development, risk analysis, and systematic approach for regulatory options was considered in full, based on the NRC report, 'NUREG/CR-6642', and the translation of its core part was conducted for ongoing research. In this methodology, the diamond tree that includes human factors, etc, additionally with normal event tree, was used. According to the analysis results of this approach, the risk analysis and the development of regulatory options were applied for the electron linear accelerators and the qualitative results were obtained. Because the field user groups were participated in this study could contribute to the basis establishment of the risk-informed regulation policy through securing consensus and inducing particle interests. It will make an important role of establishing the detail plan of ongoing research.

  9. Safety requirement of the nuclear power plants, after TMI-2 accident and their possible implementation on Bushehr NPP

    International Nuclear Information System (INIS)

    Mirhabibi, N.; Tochai, M.T.M.; Ashrafi, A.; Farnoudi, E.

    1985-01-01

    Based on the lessons learned from the TMI-2 accident and other research and developments, many improvements have been required for the design, manufacturing and operation of nuclear power plants in recent years. These requirements have already been implemented to the plants in operation and considered as new safety requirements for new plants. In the present paper these requirements and their possible implementation on Bushehr NPP are discussed. (Author)

  10. Evaluation of the Cable Types for Safety Requirements during Fire Conditions in Nuclear Facility

    International Nuclear Information System (INIS)

    Al-kattan, W.A.

    2015-01-01

    In Nuclear Power Plants (NPPs), the fire in cables causes many dangerous events in electrical or mechanical operations causing a nuclear reactor melt down. Main Control Room (MCR) in nuclear power plants have therefore, special concern in the fire protection systems. The Nuclear International Atomic Energy Agency (IAEA) has promoted the use of risk-informed and performance based methods for fire protection. These methods affirm the relevant needs to develop realistic methods to quantify the risk of fire to NPPs safety. The recent electrical cable testing has been carried out to provide empirical data on the failure modes and likelihood of fire-induced damage. In this thesis, will use fire modeling to develop fire probabilistic safety assessment to estimate the likelihood of fire induced cable damage given a specified fire profile. The objective of this work is to provide a comprehensive evaluation of the most recent fire-induced circuit failure due to different cables type that used inside the NPPs by simulation using fire modeling. One of this work scope is to suggest a suitable cable insulation material especially in case of the thermal failure thresholds, for developing the electrical cable thermal fragility distributions and evaluate parameters that influence fire-induced circuit failure modes. The main control room is implementing using the CFAST (fire simulation package). The simulation results represent the full development fire temperature and heat flux as well as the output gases. The results can be used as the basic parameters of the cables comparison and then evaluation.The importance of these results are not only for evaluating the cables but one can efficiently use them to improve the whole NPPs safety levels. The gases results of the fire simulation inside the room are oxygen, carbon monoxide, carbon dioxide, and hydrogen chloride. These gases are being used lot achieving the healthy protection of NPPs. Finally, one can measure the healthy environment

  11. Practice specific model regulations: Radiation safety of non-medical irradiation facilities. Interim report for comment

    International Nuclear Information System (INIS)

    2003-08-01

    the infrastructure aimed at achieving its maximum efficiency, and extensively covers performance regulations. The BSS cover the application of ionizing radiation for all practices and interventions and are, therefore, basic and general in nature. Users must apply these basic requirements to their own particular practices. In this context, the preamble of the BSS states that: 'The Regulatory Authority may need to provide guidance on how certain regulatory requirements are to be fulfilled for various practices, for example in regulatory guideline documents.' There are certain requirements that, when applied to specific practices, can be fulfilled through virtually only one practical solution. In these cases, the regulatory authority would use a 'shall' statement for this solution. To meet other requirements, there may be more than one option. In these cases the regulatory authority would usually indicate the recommended option with a 'should' statement, which implies that licensees may choose another alternative provided that the level of safety is equivalent. This distinction has been maintained in this 'model regulations' for irradiation facilities in order to facilitate the decision of regulatory authorities on the degree of obligation

  12. Extending enterprise architecture modelling with business goals and requirements

    Science.gov (United States)

    Engelsman, Wilco; Quartel, Dick; Jonkers, Henk; van Sinderen, Marten

    2011-02-01

    The methods for enterprise architecture (EA), such as The Open Group Architecture Framework, acknowledge the importance of requirements modelling in the development of EAs. Modelling support is needed to specify, document, communicate and reason about goals and requirements. The current modelling techniques for EA focus on the products, services, processes and applications of an enterprise. In addition, techniques may be provided to describe structured requirements lists and use cases. Little support is available however for modelling the underlying motivation of EAs in terms of stakeholder concerns and the high-level goals that address these concerns. This article describes a language that supports the modelling of this motivation. The definition of the language is based on existing work on high-level goal and requirements modelling and is aligned with an existing standard for enterprise modelling: the ArchiMate language. Furthermore, the article illustrates how EA can benefit from analysis techniques from the requirements engineering domain.

  13. Animal models for microbicide safety and efficacy testing.

    Science.gov (United States)

    Veazey, Ronald S

    2013-07-01

    Early studies have cast doubt on the utility of animal models for predicting success or failure of HIV-prevention strategies, but results of multiple human phase 3 microbicide trials, and interrogations into the discrepancies between human and animal model trials, indicate that animal models were, and are, predictive of safety and efficacy of microbicide candidates. Recent studies have shown that topically applied vaginal gels, and oral prophylaxis using single or combination antiretrovirals are indeed effective in preventing sexual HIV transmission in humans, and all of these successes were predicted in animal models. Further, prior discrepancies between animal and human results are finally being deciphered as inadequacies in study design in the model, or quite often, noncompliance in human trials, the latter being increasingly recognized as a major problem in human microbicide trials. Successful microbicide studies in humans have validated results in animal models, and several ongoing studies are further investigating questions of tissue distribution, duration of efficacy, and continued safety with repeated application of these, and other promising microbicide candidates in both murine and nonhuman primate models. Now that we finally have positive correlations with prevention strategies and protection from HIV transmission, we can retrospectively validate animal models for their ability to predict these results, and more importantly, prospectively use these models to select and advance even safer, more effective, and importantly, more durable microbicide candidates into human trials.

  14. Systems and context modeling approach to requirements analysis

    Science.gov (United States)

    Ahuja, Amrit; Muralikrishna, G.; Patwari, Puneet; Subhrojyoti, C.; Swaminathan, N.; Vin, Harrick

    2014-08-01

    Ensuring completeness and correctness of the requirements for a complex system such as the SKA is challenging. Current system engineering practice includes developing a stakeholder needs definition, a concept of operations, and defining system requirements in terms of use cases and requirements statements. We present a method that enhances this current practice into a collection of system models with mutual consistency relationships. These include stakeholder goals, needs definition and system-of-interest models, together with a context model that participates in the consistency relationships among these models. We illustrate this approach by using it to analyze the SKA system requirements.

  15. An Assessment of the VHTR Safety Distance Using the Reliability Physics Model

    International Nuclear Information System (INIS)

    Lee, Joeun; Kim, Jintae; Jae, Moosung

    2015-01-01

    In Korea planning the production of hydrogen using high temperature from nuclear power is in progress. To produce hydrogen from nuclear plants, supplying temperature above 800 .deg. C is required. Therefore, Very High Temperature Reactor (VHTR) which is able to provide about 950 .deg. C is suitable. In situation of high temperature and corrosion where hydrogen might be released easily, hydrogen production facility using VHTR has a danger of explosion. Moreover explosion not only has a bad influence upon facility itself but also on VHTR. Those explosions result in unsafe situation that cause serious damage. However, In terms of thermal-hydraulics view, long distance makes low efficiency Thus, in this study, a methodology for the safety assessment of safety distance between the hydrogen production facilities and the VHTR is developed with reliability physics model. Based on the standard safety criteria which is a value of 1 x 10 -6 , the safety distance between the hydrogen production facilities and the VHTR using reliability physics model are calculated to be a value of 60m - 100m. In the future, assessment for characteristic of VHTR, the capacity to resist pressure from outside hydrogen explosion and the overpressure for the large amount of detonation volume in detail is expected to identify more precise safety distance using this reliability physics model

  16. Customer requirement modeling and mapping of numerical control machine

    Directory of Open Access Journals (Sweden)

    Zhongqi Sheng

    2015-10-01

    Full Text Available In order to better obtain information about customer requirement and develop products meeting customer requirement, it is necessary to systematically analyze and handle the customer requirement. This article uses the product service system of numerical control machine as research objective and studies the customer requirement modeling and mapping oriented toward configuration design. It introduces the conception of requirement unit, expounds the customer requirement decomposition rules, and establishes customer requirement model; it builds the house of quality using quality function deployment and confirms the weight of technical feature of product and service; it explores the relevance rules between data using rough set theory, establishes rule database, and solves the target value of technical feature of product. Using economical turning center series numerical control machine as an example, it verifies the rationality of proposed customer requirement model.

  17. Historical development of the seismic requirements for construction of nuclear power plants in the U.S. and worldwide and their current impact on cost and safety

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    2003-01-01

    The following topics are described and discussed: Historical development of NPP seismic design requirements: Peak ground acceleration; Response spectra and damping; Floor or amplified response spectra; Effective high frequency response spectra; Seismic modeling procedures; Impact on cost (site preparation and foundations; site seismic response and generation of site dependent spectra). Potential use of indirect earthquake experience data in design and construction of NPP. Seismic contribution to safety. The following facts are summarized in two Appendices: Seismic intensity scales, and GRS safety codes and guides. (P.A.)

  18. Using cognitive modeling for requirements engineering in anesthesiology

    NARCIS (Netherlands)

    Pott, C; le Feber, J

    2005-01-01

    Cognitive modeling is a complexity reducing method to describe significant cognitive processes under a specified research focus. Here, a cognitive process model for decision making in anesthesiology is presented and applied in requirements engineering. Three decision making situations of

  19. Validation study of safety assessment model for radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Munakata, Masahiro; Takeda, Seiji; Kimura, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    The JAERI-AECL collaboration research program has been conducted to validate a groundwater flow and radionuclide transport models for safety assessment. JAERI have developed a geostatistical model for radionuclide transport through a heterogeneous geological media and verify using experimental results of field tracer tests. The simulated tracer plumes explain favorably the experimental tracer plumes. A regional groundwater flow and transport model using site-scale parameter obtained from tracer tests have been verified by comparing simulation results with observation ones of natural environmental tracer. (author)

  20. 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...

  1. Describing and analyzing effects of international differences in food safety requirements -the case of the EU versus US-

    NARCIS (Netherlands)

    Bremmers, H.J.; Meulen, van der B.M.J.; Poppe, K.J.; Wijnands, J.H.M.

    2010-01-01

    Abstract This paper compares SPS-requirements of the USA and of the EU from the perspective of the processing establishment, and analyzes the consequences of differences for national as well as firm policies. Differences in safety requirements may impede the competitiveness of the food industry.

  2. Design safety improvements of Kozloduy NPP to meet the modern safety requirements towards the old generation PWR

    International Nuclear Information System (INIS)

    Hinovski, M.P.; Sabinov, S.

    2001-01-01

    Activities related to safety improvement of Kozloduy NPP units, started at the end of 1970s included seismic resistance upgrading, fire safety improvement, reliable heat final absorber etc. During the last 10 years the approach was systematized and improved. Units 1 to 4 are of great interest; therefore here we will discuss these units only. As a result of studies and analyses performed at the end of the 1980s and the beginning of the 1990s, problems related to the safety were identified and complex of technical measures was developed and planned. A considerable part of these measures has already been implemented, and the rest will be performed during the next years. Activities were performed by stages, and at the moment the last stage is under way. It shall be finished by the year 2003. The number of the measures is quite large to describe them here in full scope -- during the first stage of the safety program (1991-1993) were developed and analyzed more than 4200 documents and more than 160 measures were executed. During the second and third stages more than 300 important improvements were realized. In the frame of the program, financed by EBRD, 10 new systems with great importance were implemented and 8 systems were significantly modified. The main measures are described below. (author)

  3. 78 FR 20454 - Safety Zones; Annual Events Requiring Safety Zones in the Captain of the Port Lake Michigan Zone

    Science.gov (United States)

    2013-04-05

    ... Zone. The last three entries within this rule have been added for races in the Chicago, IL area and on... written--Celebrate Americafest/Fire over the Fox. This event has historically involved both a fireworks... day of the event. To ensure the safety of the Celebrate Americafest/Fire over the Fox event in its...

  4. Meeting Human Reliability Requirements through Human Factors Design, Testing, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Boring

    2007-06-01

    In the design of novel systems, it is important for the human factors engineer to work in parallel with the human reliability analyst to arrive at the safest achievable design that meets design team safety goals and certification or regulatory requirements. This paper introduces the System Development Safety Triptych, a checklist of considerations for the interplay of human factors and human reliability through design, testing, and modeling in product development. This paper also explores three phases of safe system development, corresponding to the conception, design, and implementation of a system.

  5. Verification of voltage/ frequency requirement for emergency diesel generator in nuclear power plant using dynamic modeling

    International Nuclear Information System (INIS)

    Hur, J.S.; Roh, M.S.

    2013-01-01

    Full-text: One major cause of the plant shutdown is the loss of electrical power. The study is to comprehend the coping action against station blackout including emergency diesel generator, sequential loading of safety system and to ensure that the emergency diesel generator should meet requirements, especially voltage and frequency criteria using modeling tool. This paper also considered the change of the sequencing time and load capacity only for finding electrical design margin. However, the revision of load list must be verified with safety analysis. From this study, it is discovered that new load calculation is a key factor in EDG localization and in-house capability increase. (author)

  6. Integrated model of port oil piping transportation system safety including operating environment threats

    Directory of Open Access Journals (Sweden)

    Kołowrocki Krzysztof

    2017-06-01

    Full Text Available The paper presents an integrated general model of complex technical system, linking its multistate safety model and the model of its operation process including operating environment threats and considering variable at different operation states its safety structures and its components safety parameters. Under the assumption that the system has exponential safety function, the safety characteristics of the port oil piping transportation system are determined.

  7. Modeling for safety in a synthesis-centric systems engineering framework

    NARCIS (Netherlands)

    Markovski, J.; Mortel - Fronczak, van de J.M.; Ortmeier, F.; Daniel, P.

    2012-01-01

    The ever-increasing complexity of safety-critical systems puts high demands on safety assurance and certification. We focus on the development of control software, where safety) requirements engineering plays a crucial and delicate role. Nowadays, most of the safety features are ensured by the

  8. From Safety Critical Java Programs to Timed Process Models

    DEFF Research Database (Denmark)

    Thomsen, Bent; Luckow, Kasper Søe; Thomsen, Lone Leth

    2015-01-01

    frameworks, we have in recent years pursued an agenda of translating hard-real-time embedded safety critical programs written in the Safety Critical Java Profile [33] into networks of timed automata [4] and subjecting those to automated analysis using the UPPAAL model checker [10]. Several tools have been...... built and the tools have been used to analyse a number of systems for properties such as worst case execution time, schedulability and energy optimization [12–14,19,34,36,38]. In this paper we will elaborate on the theoretical underpinning of the translation from Java programs to timed automata models...... and briefly summarize some of the results based on this translation. Furthermore, we discuss future work, especially relations to the work in [16,24] as Java recently has adopted first class higher order functions in the form of lambda abstractions....

  9. Overview of the reactor safety study consequence model

    International Nuclear Information System (INIS)

    Wall, I.B.; Yaniv, S.S.; Blond, R.M.; McGrath, P.E.; Church, H.W.; Wayland, J.R.

    1977-01-01

    The Reactor Safety Study (WASH-1400) is a comprehensive assessment of the potential risk to the public from accidents in light water power reactors. The engineering analysis of the plants is described in detail in the Reactor Safety Study: it provides an estimate of the probability versus magnitude of the release of radioactive material. The consequence model, which is the subject of this paper, describes the progression of the postulated accident after the release of the radioactive material from the containment. A brief discussion of the manner in which the consequence calculations are performed is presented. The emphasis in the description is on the models and data that differ significantly from those previously used for these types of assessments. The results of the risk calculations for 100 light water power reactors are summarized

  10. Modeling interaction in the safety-critical embedded system using hybrid modeling language

    International Nuclear Information System (INIS)

    Lee, Na Young; Choi, Jin Young; Kim, Jin Hyun; Bang, Ki Seok; Lee, Jang Soo

    2004-01-01

    To adapt the advanced digital technologies in the Instrumentation and Control (I and C) system of Nuclear Power Plants (NPPs), the more rigorous certification process including a formal verification is required to apply the advanced digital technologies in the NPPs. In this work, we concentrated on development procedure of Real Time Operating System (RTOS) software for use in one of the safety critical systems, Plant Protection System (PPS). Statecharts is used during development process to specify and simulate the model RTOS model. Model certifier is used to verify properties, such as Schedulability, priority inversion. Since the RTOS cannot operate by itself, we assume set of tasks to check properties. Based on the assumption, two sets of tasks are implemented in this work. We executed simulation to check whether it shows correct behavior as we designed. Important properties are verified using Model certifier. For the RTOS, however, timing properties should be checked, and Statecharts has limitation since it does not support time in it, therefore, time is considered as discrete tick. So we chose timed automata based tool, UPPAAL to verify timing properties. Model was simplified and modified. But timing constraints can be more realistic. When properties are not satisfied we can modify scheduler based on timing records during simulation. (author)

  11. Modeling of preventive maintenance changes influence upon flight safety indexes

    Directory of Open Access Journals (Sweden)

    А.В. Гончаренко

    2004-03-01

    Full Text Available  It is considered a simplified model of connection between the catastrophic events flow frequency and both preventive maintenance changes periodicity and diagnosis depth of aviation equipment. It is deduced specific formulas for computing the changes and diagnostics parameters influence upon the values of both the catastrophic events flow frequency and technical-economical control factor criterion of flight safety levels.

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

    International Nuclear Information System (INIS)

    Ostenso, A.; May, R.

    1996-12-01

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

  13. 33 CFR 96.240 - What functional requirements must a safety management system meet?

    Science.gov (United States)

    2010-07-01

    ... a safety management system meet? 96.240 Section 96.240 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.240 What functional...

  14. Needs, requirements and challenges for technical support to nuclear safety authority

    International Nuclear Information System (INIS)

    Madonna, A.; Orsini, G.

    2010-01-01

    To face the very broad range of technical matters on which the regulatory and licensing activity are based, and related research and development activity, the Nuclear Safety Authorities (NSA) may need to rely upon external technical and scientific support. In providing technical support to NSA, the experience shows, from one side, the importance to have technical support organizations (TSO) with recognized competence, independence and appropriate regulatory view, and from the other side, the importance to have within the NSAs well developed management and technical capability to address, coordinate and use the results of the external technical support. Retaining the NSA the full responsibility for the final decision. Under which conditions and modus operandi the external support shall be provided in order to comply with requirements of being independent, competent and timely provided, fulfilling the administrative procedures, is the subject of attention and consideration of TSO function today. The Italian regulatory body is currently going to be institutionally re-established according to new law approved in 2009 /1/ and it needs to be resourced and fully organized with necessary capacities in the nearest future. The perspective of a new nuclear program, recently launched by the government, with significant incoming tasks for regulation and licensing, against the existing limited resources, let foresee a substantial potential need for technical support and advice. ITER-Consult (Ltd), created in 2003 in Italy, has well developed capabilities to provide independent technical evaluation and support to NSAs, to maintain safety culture and updated knowledge, to transfer know how and to establish international cooperation and networking. This mission is guided assuming as values the independence, the professional competence, the transparency, the credibility and the establishment of respectful relationship with the partners. Challenges exist for funding and operational

  15. Capabilities and requirements for modelling radionuclide transport in the geosphere

    International Nuclear Information System (INIS)

    Paige, R.W.; Piper, D.

    1989-02-01

    This report gives an overview of geosphere flow and transport models suitable for use by the Department of the Environment in the performance assessment of radioactive waste disposal sites. An outline methodology for geosphere modelling is proposed, consisting of a number of different types of model. A brief description of each of the component models is given, indicating the purpose of the model, the processes being modelled and the methodologies adopted. Areas requiring development are noted. (author)

  16. Modelizing home safety as experienced by people with mental illness.

    Science.gov (United States)

    Désormeaux-Moreau, Marjorie; Larivière, Nadine; Aubin, Ginette

    2018-05-01

    As more individuals with mental disorders now live in the community and as the custodial care housing model has shifted to supported housing, home safety has become a rising issue, however, not well documented. To describe the phenomenon of home safety for people with a mental disorder as well as its contributing factors. A descriptive qualitative design was used. Individual interviews were conducted with persons with a mental disorder (n = 8), while focus groups were conducted with relatives, health and social service providers and community stakeholders (n = 21). The data were analyzed with the grounded theory analysis as described by Paillé (1994). Findings suggest that home safety implies risk and protective factors, which are associated with (1) the person's characteristics; (2) the quality of the home environment; (3) the nature of the activities in which the individual engages. These dimensions are interrelated so that home incidents arise from a dynamic interaction between risk and protective factors. Home incidents therefore occur when the interaction between these dimensions is altered. Considering this situation, Occupational Therapists are well positioned to play a leading role and act as key contributors in the area of home safety in people with mental disorders.

  17. An effective technique for the software requirements analysis of NPP safety-critical systems, based on software inspection, requirements traceability, and formal specification

    International Nuclear Information System (INIS)

    Koo, Seo Ryong; Seong, Poong Hyun; Yoo, Junbeom; Cha, Sung Deok; Yoo, Yeong Jae

    2005-01-01

    A thorough requirements analysis is indispensable for developing and implementing safety-critical software systems such as nuclear power plant (NPP) software systems because a single error in the requirements can generate serious software faults. However, it is very difficult to completely analyze system requirements. In this paper, an effective technique for the software requirements analysis is suggested. For requirements verification and validation (V and V) tasks, our technique uses software inspection, requirement traceability, and formal specification with structural decomposition. Software inspection and requirements traceability analysis are widely considered the most effective software V and V methods. Although formal methods are also considered an effective V and V activity, they are difficult to use properly in the nuclear fields as well as in other fields because of their mathematical nature. In this work, we propose an integrated environment (IE) approach for requirements, which is an integrated approach that enables easy inspection by combining requirement traceability and effective use of a formal method. The paper also introduces computer-aided tools for supporting IE approach for requirements. Called the nuclear software inspection support and requirements traceability (NuSISRT), the tool incorporates software inspection, requirement traceability, and formal specification capabilities. We designed the NuSISRT to partially automate software inspection and analysis of requirement traceability. In addition, for the formal specification and analysis, we used the formal requirements specification and analysis tool for nuclear engineering (NuSRS)

  18. 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.)

  19. Requirements Validation: Execution of UML Models with CPN Tools

    DEFF Research Database (Denmark)

    Machado, Ricardo J.; Lassen, Kristian Bisgaard; Oliveira, Sérgio

    2007-01-01

    Requirements validation is a critical task in any engineering project. The confrontation of stakeholders with static requirements models is not enough, since stakeholders with non-computer science education are not able to discover all the inter-dependencies between the elicited requirements. Eve...... requirements, where the system to be built must explicitly support the interaction between people within a pervasive cooperative workflow execution. A case study from a real project is used to illustrate the proposed approach....

  20. Consequence model of the German reactor safety study

    International Nuclear Information System (INIS)

    Bayer, A.; Aldrich, D.; Burkart, K.; Horsch, F.; Hubschmann, W.; Schueckler, M.; Vogt, S.

    1979-01-01

    The consequency model developed for phase A of the German Reactor Safety Study (RSS) is similar in many respects to its counterpart in WASH-1400. As in that previous study, the model describes the atmosphere dispersion and transport of radioactive material released from the containment during a postulated reactor accident, and predicts its interaction with and influence on man. Differences do exist between the two models however, for the following reasons: (1) to more adequately reflect central European conditions, (2) to include improved submodels, and (3) to apply additional data and knowledge that have become available since publication of WASH-1400. The consequence model as used in phase A of the German RSS is described, highlighting differences between it and the U.S. model