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Sample records for safety requirements doe

  1. 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.

  2. 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

  3. 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

  4. DOE standard: Firearms safety

    International Nuclear Information System (INIS)

    1996-02-01

    Information in this document is applicable to all DOE facilities, elements, and contractors engaged in work that requires the use of firearms as provided by law or contract. The standard in this document provides principles and practices for implementing a safe and effective firearms safety program for protective forces and for non-security use of firearms. This document describes acceptable interpretations and methods for meeting Order requirements

  5. DOE standard: Firearms safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Information in this document is applicable to all DOE facilities, elements, and contractors engaged in work that requires the use of firearms as provided by law or contract. The standard in this document provides principles and practices for implementing a safe and effective firearms safety program for protective forces and for non-security use of firearms. This document describes acceptable interpretations and methods for meeting Order requirements.

  6. DOE handbook electrical safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    Electrical Safety Handbook presents the Department of Energy (DOE) safety standards for DOE field offices or facilities involved in the use of electrical energy. It has been prepared to provide a uniform set of electrical safety guidance and information for DOE installations to effect a reduction or elimination of risks associated with the use of electrical energy. The objectives of this handbook are to enhance electrical safety awareness and mitigate electrical hazards to employees, the public, and the environment.

  7. ORNL implementation of new health and safety requirements (DOE Order 5480.11)

    International Nuclear Information System (INIS)

    Abercrombie, J.S.

    1988-01-01

    New mandates in radiological protection outlined in DOE Order 5480. 11, include changes in the methodology for determining total radiation dose, ALARA program accountability, monitoring requirements, and standards for public entrance into controlled areas. The new order places distinct requirements concerning training at all DOE facilities. Radiation protection training requirements are addressed, including the effective communication of operations changes to all employees. This paper details the endeavors underway at ORNL in designing, developing, and delivering the training required by the new mandates. Strategies taken to reach the intended goals are explained. Efforts involve the design and implementation of the above mentioned radiation protection programs, a job-specific ALARA instructional package, and a Risk-Based Philosophy program matched to operational changes. 4 refs., 5 tabs

  8. Does competition influence safety?

    International Nuclear Information System (INIS)

    Pamme, H.

    2000-01-01

    Competition in the deregulated electricity market does not leave nuclear power plants unaffected. Operators seek to run their plants at maximum availability and with optimized cost structures so that specific generating costs are minimized. The 'costs of safety', with their fixed-cost character, are elements of this cost structure. Hence the question whether safety is going to suffer under the cost pressure on the market. The study shows that the process of economic optimization does not permit cost minimization for its own sake in the area of operating costs which can be influenced by management or are 'avoidable'. The basis of assessment rather must be potential risks which could entail losses of availability. Prophylactic investments made in order to avoid losses of availability to a large extent also imply unchanged or even higher levels of safety. Economic viability and safety thus are closely correlated. Competition in a deregulated marekt so far has not done any direct harm to plant safety. An even more efficient use of scarce funds and, hopefully, a tolerable political environment should allow the safety level of nuclear power plants to be upheld, and safety culture to be maintained, also in the future. (orig.) [de

  9. Removing unreasonable conservatisms in DOE safety analysis

    International Nuclear Information System (INIS)

    BISHOP, G.E.

    1999-01-01

    While nuclear safety analyses must always be conservative, invoking excessive conservatisms does not provide additional margins of safety. Rather, beyond a fairly narrow point, conservatisms skew a facility's true safety envelope by exaggerating risks and creating unreasonable bounds on what is required for safety. The conservatism has itself become unreasonable. A thorough review of the assumptions and methodologies contained in a facility's safety analysis can provide substantial reward, reducing both construction and operational costs without compromising actual safety

  10. 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.

  11. 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

  12. 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...

  13. Why does Safety Culture Matter?

    International Nuclear Information System (INIS)

    Dahlgren-Persson, Kerstin

    2008-01-01

    Dr. Kerstin Dahlgren-Persson, from the IAEA presented a plenary paper on 'Why does safety culture matter?'. The paper discussed the main conclusions of a 1998 IAEA conference on shortcomings in safety management. The conference included case studies of TVA, Cooper, Peach Bottom, Millstone, Ontario Hydro, Barsebaeck and Oskarshamn. Common symptoms included insularity; disproportionate focus on technical issues, high initial performance, lack of corporate oversight, changing management direction and cost cutting, repeat problems, and regulatory dissatisfaction. Behind these symptoms was lack of senior utility leadership with the insight, knowledge and ability to manage the unique interaction between the technology, economics, human factors and safety in a changing nuclear environment. Shortcomings relating to the regulator included lack of criteria for when regulatory actions should be taken in response to degradations in safety management, and the inability of some regulators to influence at the senior utility management level. The paper also made the following key points: - Human error is not always symptomatic of a poor safety culture. Effective root cause analysis (such as that carried out for the Columbia accident investigation) is essential to correctly differentiate between situational issues at a point in time and those rooted in organizational culture. - Leaders change culture by holding different assumptions and by making them visible through their words and action. - Regulators should consider how their regulatory strategy influences licensees. For example, a prescriptive strategy can foster a compliance based approach

  14. 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.)

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

    International Nuclear Information System (INIS)

    2000-01-01

    other reactor types, including innovative developments in future systems, some of the requirements may not be applicable, or may need some judgment in their interpretation. Various Safety Guides will provide guidance in the interpretation and implementation of these requirements. This publication is intended for use by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies. This publication establishes design requirements for structures, systems and components important to safety that must be met for safe operation of a nuclear power plant, and for preventing or mitigating the consequences of events that could jeopardize safety. It also establishes requirements for a comprehensive safety assessment, which is carried out in order to identify the potential hazards that may arise from the operation of the plant, under the various plant states (operational states and accident conditions). The safety assessment process includes the complementary techniques of deterministic safety analysis and probabilistic safety analysis. These analyses necessitate consideration of postulated initiating events (PlEs), which include many factors that, singly or in combination, may affect safety and which may: originate in the operation of the nuclear power plant itself; be caused by human action; be directly related to the nuclear power plant and its environment. This publication also addresses events that are very unlikely to occur, such as severe accidents that may result in major radioactive releases, and for which it may be appropriate and practicable to provide preventive or mitigatory features in the design. This publication does not address: external natural or human induced events that are extremely unlikely (such as the impact of a meteorite or an artificial satellite); conventional industrial accidents that under no circumstances could affect the safety of the nuclear power plant; or non-radiological effects arising

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

    International Nuclear Information System (INIS)

    2004-01-01

    other reactor types, including innovative developments in future systems, some of the requirements may not be applicable, or may need some judgment in their interpretation. Various Safety Guides will provide guidance in the interpretation and implementation of these requirements. This publication is intended for use by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies. This publication establishes design requirements for structures, systems and components important to safety that must be met for safe operation of a nuclear power plant, and for preventing or mitigating the consequences of events that could jeopardize safety. It also establishes requirements for a comprehensive safety assessment, which is carried out in order to identify the potential hazards that may arise from the operation of the plant, under the various plant states (operational states and accident conditions). The safety assessment process includes the complementary techniques of deterministic safety analysis and probabilistic safety analysis. These analyses necessitate consideration of postulated initiating events (PlEs), which include many factors that, singly or in combination, may affect safety and which may: originate in the operation of the nuclear power plant itself. Be caused by human action. Be directly related to the nuclear power plant and its environment. This publication also addresses events that are very unlikely to occur, such as severe accidents that may result in major radioactive releases, and for which it may be appropriate and practicable to provide preventive or mitigatory features in the design. This publication does not address: external natural or human induced events that are extremely unlikely (such as the impact of a meteorite or an artificial satellite). Conventional industrial accidents that under no circumstances could affect the safety of the nuclear power plant. Or non-radiological effects arising

  17. DOE explosives safety manual. Revision 7

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    This manual prescribes the Department of Energy (DOE) safety rules used to implement the DOE safety policy for operations involving explosives. This manual is applicable to all DOE facilities engaged in operations of development, manufacturing, handling, storage, transportation, processing, or testing of explosives, pyrotechnics and propellants, or assemblies containing these materials. The standards of this manual deal with the operations involving explosives, pyrotechnics and propellants, and the safe management of such operations. The design of all new explosives facilities shall conform to the requirements established in this manual and implemented in DOE 6430.1A, ``General Design Criteria Manual.`` It is not intended that existing physical facilities be changed arbitrarily to comply with these provisions, except as required by law. Existing facilities that do not comply with these standards may continue to be used for the balance of their functional life, as long as the current operation presents no significantly greater risk than that assumed when the facility was originally designed and it can be demonstrated clearly that a modification to bring the facility into compliance is not feasible. However, in the case of a major renovation, the facility must be brought into compliance with current standards. The standards are presented as either mandatory or advisory. Mandatory standards, denoted by the words ``shall,`` ``must,`` or ``will,`` are requirements that must be followed unless written authority for deviation is granted as an exemption by the DOE. Advisory standards denoted by ``should`` or ``may`` are standards that may be deviated from with a waiver granted by facility management.

  18. 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

  19. 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?

  20. Application of the DOE Nuclear Safety Policy goal

    International Nuclear Information System (INIS)

    Coles, G.A.; Hey, B.E.; Leach, D.S.; Muhlestein, L.D.

    1992-08-01

    The US Department of Energy (DOE) issued their Nuclear Safety Policy for implementation on September 9, 1991. The statement noted that it was the DOE's policy that the general public should be protected such that no individual would bear significant additional risk to health and safety from operation of their nuclear facilities above the risks to which members of the general population were normally exposed. The intent is that from the nuclear safety policy will follow specific safety rules, orders, standards and other requirements. The DOE Nuclear Safety Policy provides general statements in the areas of management involvement and accountability, providing technically competent personnel, oversight and self-assessment, promoting a safety culture, and quantitative safety goals as aiming points for performance. In general, most DOE Management and Operating Contractors should have programs in place which address the general statements noted above. Thus, compliance with the general statements of the DOE Nuclear Safety Policy should present no significant difficulty. Consequently, the focus of this paper will be the two quantitative safety goals reproduced below from the DOE Nuclear Safety Policy. ''The risk to an average individual in the vicinity of a DOE facility for prompt fatalities that might result from accidents should not exceed one tenth of one percent (0.1 %) of the sum of prompt fatalities resulting from other accidents to which members of the population are generally exposed. For evaluation purposes, individuals are assumed to be located within one mile of the site boundary.'' ''The risk to the population in the area of a DOE nuclear facility for cancer fatalities that might result from operations should not exceed one tenth of one percent (0.1 %) of the sum of all cancer fatality risks resulting from all other causes. For evaluation purposes, individuals are assumed to be located within 10 miles of the site boundary.''

  1. 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

  2. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

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

  3. 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.

  4. DOE Defense Program (DP) safety programs. Final report, Task 003

    International Nuclear Information System (INIS)

    1998-01-01

    The overall objective of the work on Task 003 of Subcontract 9-X52-W7423-1 was to provide LANL with support to the DOE Defense Program (DP) Safety Programs. The effort included the identification of appropriate safety requirements, the refinement of a DP-specific Safety Analysis Report (SAR) Format and Content Guide (FCG) and Comprehensive Review Plan (CRP), incorporation of graded approach instructions into the guidance, and the development of a safety analysis methodologies document. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective is provided here

  5. US DOE Pipeline Unplugging Requirements Development

    International Nuclear Information System (INIS)

    Rivera, J.; McDaniel, D.

    2009-01-01

    Department of Energy (DOE) sites around the country have an ongoing effort to transport and process several tons of radioactive waste in the form of slurry (liquids and solids) from storage tanks to processing facilities. The system of pipes used for the transportation of this waste needs technology for maintenance and for the prevention (and correction) of pipeline plugging. The unplugging technologies that have been tested and evaluated at Florida International University include ones from NuVision Engineering, AIMM and AquaMiser. NuVision's technology acts as an ocean wave does on beach erosion. It can operate on a long pipeline that has drained down below a blockage. AIMM Technology's Hydrokinetic TM process uses a sonic resonance with a cleaning water stream. This sonic resonance travels through the water stream and transfers vibration to both the pipe and the blockage. The AquaMiser line of water blasting equipment combines 15,000- to 40,000-psi water injection technology to unplug pipelines. Some sites cannot allow this level of pressure in their pipes. After reviewing the results of every test, including the benefits, advantages and disadvantages of each technology, requirements were developed for pressure, personnel training, environmental concerns, safety, and compatibility with current systems, operability, reliability, maintainability and cost. (authors)

  6. Comparison of selected DOE and non-DOE requirements, standards, and practices for Low-Level Radioactive Waste Disposal

    International Nuclear Information System (INIS)

    Cole, L.; Kudera, D.; Newberry, W.

    1995-12-01

    This document results from the Secretary of Energy's response to Defense Nuclear Facilities Safety Board Recommendation 94--2. The Secretary stated that the US Department of Energy (DOE) would ''address such issues as...the need for additional requirements, standards, and guidance on low-level radioactive waste management. '' The authors gathered information and compared DOE requirements and standards for the safety aspects Of low-level disposal with similar requirements and standards of non-DOE entities

  7. 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

  8. 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

  9. 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.

  10. Nuclear criticality safety training: guidelines for DOE contractors

    International Nuclear Information System (INIS)

    Crowell, M.R.

    1983-09-01

    The DOE Order 5480.1A, Chapter V, Safety of Nuclear Facilities, establishes safety procedures and requirements for DOE nuclear facilities. This guide has been developed as an aid to implementing the Chapter V requirements pertaining to nuclear criticality safety training. The guide outlines relevant conceptual knowledge and demonstrated good practices in job performance. It addresses training program operations requirements in the areas of employee evaluations, employee training records, training program evaluations, and training program records. It also suggests appropriate feedback mechanisms for criticality safety training program improvement. The emphasis is on academic rather than hands-on training. This allows a decoupling of these guidelines from specific facilities. It would be unrealistic to dictate a universal program of training because of the wide variation of operations, levels of experience, and work environments among DOE contractors and facilities. Hence, these guidelines do not address the actual implementation of a nuclear criticality safety training program, but rather they outline the general characteristics that should be included

  11. Orientation to occupational safety compliance in DOE: Student manual, Book 1

    International Nuclear Information System (INIS)

    1993-01-01

    The purpose of this document is to present Federal safety and health program requirements, provide an overview of DOE Safety and Health Programs, and to present information on the changes presently underway in the Department of Energy regarding safety and health. Public Law 91--596, Executive Order 12196,29 CFR 1960 and various related DOE Orders will be presented

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. DOE-RL Integrated Safety Management System Description

    International Nuclear Information System (INIS)

    SHOOP, D.S.

    2000-01-01

    The purpose of this Integrated Safety Management System Description (ISMSD) is to describe the U.S. Department of Energy (DOE), Richland Operations Office (RL) ISMS as implemented through the RL Integrated Management System (RIMS). This ISMSD does not impose additional requirements but rather provides an overview describing how various parts of the ISMS fit together. Specific requirements for each of the core functions and guiding principles are established in other implementing processes, procedures, and program descriptions that comprise RIMS. RL is organized to conduct work through operating contracts; therefore, it is extremely difficult to provide an adequate ISMS description that only addresses RL functions. Of necessity, this ISMSD contains some information on contractor processes and procedures which then require RL approval or oversight. This ISMSD does not purport to contain a full description of the contractors' ISM System Descriptions

  4. DOE-RL Integrated Safety Management System Description

    CERN Document Server

    Shoop, D S

    2000-01-01

    The purpose of this Integrated Safety Management System Description (ISMSD) is to describe the U.S. Department of Energy (DOE), Richland Operations Office (RL) ISMS as implemented through the RL Integrated Management System (RIMS). This ISMSD does not impose additional requirements but rather provides an overview describing how various parts of the ISMS fit together. Specific requirements for each of the core functions and guiding principles are established in other implementing processes, procedures, and program descriptions that comprise RIMS. RL is organized to conduct work through operating contracts; therefore, it is extremely difficult to provide an adequate ISMS description that only addresses RL functions. Of necessity, this ISMSD contains some information on contractor processes and procedures which then require RL approval or oversight. This ISMSD does not purport to contain a full description of the contractors' ISM System Descriptions.

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

    International Nuclear Information System (INIS)

    Jones, G.

    2013-01-01

    assessment of these aspects will, in general, have many commonalities with the safety assessment that is carried out to address associated radiation risks. However, it should be noted that the Safety Requirements publication, does not establish requirements for combining these non-radiological assessments with the radiological safety assessment, or make recommendations for how to assess non-radiological hazards; consequently these aspects are not addressed in the paper. The requirements of the safety assessment Safety Requirements publication are discussed below. (authors)

  6. 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

  7. 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.

  8. 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

  9. Nuclear health and safety. Status of GAO's environmental, safety, and health recommendations to DOE

    International Nuclear Information System (INIS)

    England-Joseph, Judy; Bannerman, Carl J.; Fenzel, William F.; Brack, David L.; Harter, Frederick A.

    1990-04-01

    DOE's operations are carried out at many contractor-operated sites around the country, including major sites within the nuclear weapons complex that are involved in the production of nuclear material for weapons and naval fuel. At these sites DOE contractors routinely use and generate large quantities of a wide range of hazardous and radioactive materials. Because these materials require special handling by workers. Also pursuant to Mar. 15, 1989, request, we provided you a report entitled Nuclear Health and Safety: Need for Improved Responsiveness to Problems at DOE Sites (GAO/RCED-90-101, Mar. 28, 1990). to prevent exposure to themselves or releases into the environment, DOE's weapons complex, considered in its entirety, is among the potentially more dangerous industrial operations in the world. Over the last decade, at the request of the Congress, we have carried out a series of assessments and evaluations of various aspects of the complex. In over 60 reports and testimonies published since 1990, we have called attention to the mounting problems facing DOE's nuclear weapons complex. This body of work includes (1) identifying serious, costly, and widespread environmental, safety, and health problems at DOE facilities, (2) calling for outside independent oversight of DOE's nuclear operations, and (3) making recommendations to DOE to strengthen its oversight, providing more detailed information and plans to the Congress, and improving its management and accounting practices. In total, our reports and testimonies have included 54 recommendations to DOE, in addition to recommendations to the Congress, concerning environmental, safety, and health matters at the complex. We consider 23 of the 54 recommendations to be still open. The open recommendations call for improvements such as tighter program controls and clearer standards and policies related to environmental, safety, and health matters

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

    International Nuclear Information System (INIS)

    Webb, A.B.

    1985-01-01

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

  11. 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)

  12. Overview of DOE/ONS criticality safety projects

    International Nuclear Information System (INIS)

    Barber, R.W.; Brown, B.P.; Hopper, C.M.

    1985-01-01

    The evolution of Federal involvement with nuclear criticality safety has traversed through the 1940's and early 1950's with the Manhattan Engineering District, the 1950's and 1960's with the Atomic Energy Commission, the early 1970's with the Energy Research and Development Administration, and the late 1970's to date with the US Department of Energy. The importance of nuclear criticality safety has been maintained throughout these periods; however, criticality safety has received shifting emphases in research/applications, promulgations of regulations/standards, origins of fiscal support and organization. In June 1981 the Office of Nuclear Safety was established in response to a Department of Energy study of the impact of the March 1979 Three Mile Island accident. The organizational structure of the ONS, its program for establishing and maintaining a progressive nuclear criticality safety program, and associated projects, and current history of ONS's fiscal support of program projects is presented. With the establishment of the ONS came concomitant missions to develop and maintain nuclear safety policy and requirements, to provide independent assurance that nuclear operations are performed safely, to provide resources and management for DOE responses to nuclear accidents, and to provide technical support. In the past four years, ONS has developed and initiated a continuing Department Nuclear Criticality Safety Program in such areas as communications and information, physics of criticality, knowledge of factors affecting criticality, and computational capability

  13. 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

  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. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. DOE/DOE Tight Oil Flammability & Transportation Spill Safety

    Energy Technology Data Exchange (ETDEWEB)

    Lord, David L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-12-01

    This presentation describes crude oils, their phase behavior, the SPR vapor pressure program, and presents data comparisons from various analytical techniques. The overall objective is to describe physical properties of crude oil relevant to flammability and transport safety

  5. 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.

  6. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1993-11-01

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

  7. Nursing home safety: does financial performance matter?

    Science.gov (United States)

    Oetjen, Reid M; Zhao, Mei; Liu, Darren; Carretta, Henry J

    2011-01-01

    This study examines the relationship between financial performance and selected safety measures of nursing homes in the State of Florida. We used descriptive analysis on a total sample of 1,197. Safety information was from the Online Survey, Certification and Reporting (OSCAR) data of 2003 to 2005, while the financial performance measures were from the Medicare cost reports of 2002 to 2004. Finally, we examined the most frequently cited deficiencies as well as the relationship between financial performance and quality indicators. Nursing homes in the bottom quartile of financial performance perform poorly on most resident-safety measures of care; however, nursing homes in the top two financial categories also experienced a higher number of deficiencies. Nursing homes in the next to lowest quartile of financial performance category best perform on most of these safety measures. The results reinforce the need to monitor nursing home quality and resident safety in US nursing homes, especially among facilities with poor overall financial performance.

  8. Safety Oversight of Decommissioning Activities at DOE Nuclear Sites

    International Nuclear Information System (INIS)

    Zull, Lawrence M.; Yeniscavich, William

    2008-01-01

    The Defense Nuclear Facilities Safety Board (Board) is an independent federal agency established by Congress in 1988 to provide nuclear safety oversight of activities at U.S. Department of Energy (DOE) defense nuclear facilities. The activities under the Board's jurisdiction include the design, construction, startup, operation, and decommissioning of defense nuclear facilities at DOE sites. This paper reviews the Board's safety oversight of decommissioning activities at DOE sites, identifies the safety problems observed, and discusses Board initiatives to improve the safety of decommissioning activities at DOE sites. The decommissioning of former defense nuclear facilities has reduced the risk of radioactive material contamination and exposure to the public and site workers. In general, efforts to perform decommissioning work at DOE defense nuclear sites have been successful, and contractors performing decommissioning work have a good safety record. Decommissioning activities have recently been completed at sites identified for closure, including the Rocky Flats Environmental Technology Site, the Fernald Closure Project, and the Miamisburg Closure Project (the Mound site). The Rocky Flats and Fernald sites, which produced plutonium parts and uranium materials for defense needs (respectively), have been turned into wildlife refuges. The Mound site, which performed R and D activities on nuclear materials, has been converted into an industrial and technology park called the Mound Advanced Technology Center. The DOE Office of Legacy Management is responsible for the long term stewardship of these former EM sites. The Board has reviewed many decommissioning activities, and noted that there are valuable lessons learned that can benefit both DOE and the contractor. As part of its ongoing safety oversight responsibilities, the Board and its staff will continue to review the safety of DOE and contractor decommissioning activities at DOE defense nuclear sites

  9. 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.)

  10. 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

  11. Construction safety management guide for use with DOE Order 440.1

    International Nuclear Information System (INIS)

    1997-01-01

    DOE Order (DOE O) 440.1, Worker Protection Management for DOE Federal and Contractor Employees, establishes the framework for an effective worker protection program to reduce or eliminate accidental losses, injuries, and illnesses by providing workers with places of employment free of recognized hazards. In addition to prescribing program requirements applicable to all activities performed by DOE and its contractors, DOE O 440.1 provides specific requirements applicable only to construction activities. The intent of these construction-specific requirements is to compel the proactive management of construction safety on a project-by-project basis and, to the greatest extent possible, integrate the management of safety and health, both in terms of project personnel and management methodologies, with the management of the other primary elements of construction project performance: quality, cost and schedule

  12. 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.)

  13. Construction safety in DOE. Part 1, Students guide

    Energy Technology Data Exchange (ETDEWEB)

    Handwerk, E C

    1993-08-01

    This report is the first part of a compilation of safety standards for construction activities on DOE facilities. This report covers the following areas: general safety and health provisions; occupational health and environmental control/haz mat; personal protective equipment; fire protection and prevention; signs, signals, and barricades; materials handling, storage, use, and disposal; hand and power tools; welding and cutting; electrical; and scaffolding.

  14. 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

  15. 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

  16. Does underground storage still require sophisticated studies?

    International Nuclear Information System (INIS)

    Marsily, G. de

    1997-01-01

    Most countries agree to the necessity of burying high or medium-level wastes in geological layers situated at a few hundred meters below the ground level. The advantages and disadvantages of different types of rock such as salt, clay, granite and volcanic material are examined. Sophisticated studies are lead to determine the best geological confinement but questions arise about the time for which safety must be ensured. France has chosen 3 possible sites. These sites are geologically described in the article. The final place will be proposed after a testing phase of about 5 years in an underground facility. (A.C.)

  17. Nuclear Data Activities in Support of the DOE Nuclear Criticality Safety Program

    International Nuclear Information System (INIS)

    Westfall, R.M.; McKnight, R.D.

    2005-01-01

    The DOE Nuclear Criticality Safety Program (NCSP) provides the technical infrastructure maintenance for those technologies applied in the evaluation and performance of safe fissionable-material operations in the DOE complex. These technologies include an Analytical Methods element for neutron transport as well as the development of sensitivity/uncertainty methods, the performance of Critical Experiments, evaluation and qualification of experiments as Benchmarks, and a comprehensive Nuclear Data program coordinated by the NCSP Nuclear Data Advisory Group (NDAG).The NDAG gathers and evaluates differential and integral nuclear data, identifies deficiencies, and recommends priorities on meeting DOE criticality safety needs to the NCSP Criticality Safety Support Group (CSSG). Then the NDAG identifies the required resources and unique capabilities for meeting these needs, not only for performing measurements but also for data evaluation with nuclear model codes as well as for data processing for criticality safety applications. The NDAG coordinates effort with the leadership of the National Nuclear Data Center, the Cross Section Evaluation Working Group (CSEWG), and the Working Party on International Evaluation Cooperation (WPEC) of the OECD/NEA Nuclear Science Committee. The overall objective is to expedite the issuance of new data and methods to the DOE criticality safety user. This paper describes these activities in detail, with examples based upon special studies being performed in support of criticality safety for a variety of DOE operations

  18. 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...

  19. CONACS: the DOE safety analysis system

    International Nuclear Information System (INIS)

    Martin, F.J.; Armstrong, G.R.; Niccoli, L.G.

    1985-03-01

    The CONtainment Analysis Code System (CONACS) is a large, comprehensive scientific simulation system for predicting conditions in an LMR facility following the occurrence of a postulated accident. It has now been developed to a stage of completion that can be referred to as a limited operational version. This version forms a permanent portion of the ultimate system. Because CONACS was developed with change in mind, it is now possible to draw on this strength to respond to changing requirements arising from advanced design concepts. The generalized design applications in the nuclear and non-nuclear fields and the quality assurance applied to the project make those adaptations reliable. In this paper the results of prototype tests and the implications of limited version tests are presented along with a brief description of CONACS and its relationship to LMR design optimization and cost reduction

  20. CONACS, the DOE safety analysis system

    International Nuclear Information System (INIS)

    Martin, F.J.; Armstrong, G.R.; Niccoli, L.G.

    1985-01-01

    The CONtainment Analysis Code System (CONACS) is a large, comprehensive scientific simulation system for predicting conditions in an LMR facility following the occurrence of a postulated accident. It has now been developed to a stage of completion that can be referred to as a limited operational version. This version forms a permanent portion of the ultimate system. Because CONACS was developed with change in mind it is now possible to draw on this strength to respond to changing requirements arising from advanced design concepts. The generalized design applications in the nuclear and non-nuclear fields and the quality assurance applied to the project make those adaptations reliable. In this paper the results of prototype tests and the implications of limited version tests are presented along with a brief description of CONACS and its relationship to LMR design optimization and cost reduction

  1. 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.

  2. 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.

  3. 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.

  4. When does food refusal require professional intervention?

    OpenAIRE

    Dovey, Terence M.; Farrow, Claire V.; Martin, Clarissa I.; Isherwood, Elaine; Halford, Jason C.G.

    2009-01-01

    Food refusal can have the potential to lead to nutritional deficiencies, which increases the risk of a variety of communicable and non-communicable diseases. Deciding when food refusal requires professional intervention is complicated by the fact that there is a natural and appropriate stage in a child's development that is characterised by increased levels of rejection of both previously accepted and novel food items. Therefore, choosing to intervene is difficult, which if handled badly can ...

  5. 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.

  6. 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.

  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

    In June 1995, an OECD Support Group was set up to perform a broad study of the safety research needs of Russian-designed reactors. The emphasis of the study is on the VVER-type reactors in part because of the larger base of knowledge within the NEA Member countries related to LWRs. For the RBMKs, the study does not make the judgement that such reactors can be brought to acceptable levels of safety but focuses on near term efforts that can contribute to reducing the risk to the public. The need for the safety research must be evaluated in context of the lifetime of the reactors. The principal outcome of the work of the Support Group is the identification of a number of research topics which the members believe should receive priority attention over the next several years if risk levels are to be reduced and public safety enhanced. These appear in the Conclusions and Recommendations section of the report, and are the following: - The most important near-term need for VVER and RBMK safety research is to establish a sound technical basis for the emergency operating procedures used by the plant staff to prevent or halt the progression of accidents (i.e., Accident Management) and for plant safety improvements. - Co-operation of Western and Eastern experts should help to avoid East-West know-how gaps in the future, as safety technology continues to improve. - Safety research in Eastern countries will make an important contribution to public safety as it has in OECD countries. - RBMK safety research, including verification of codes, starts from a smaller base of experience than VVER, and is at an earlier stage of development. Technical Conclusions: - Research to improve human performance and operational safety of VVER and RBMK plants is extremely important. - VVER thermal-hydraulic and reactor physics research should focus on full validation of codes to VVER-specific features, and on extension of experimental data base. - Methods of assessing VVER pressure boundary

  8. DOE Handbook: Supplementary guidance and design experience for the fusion safety standards DOE-STD-6002-96 and DOE-STD-6003-96

    Energy Technology Data Exchange (ETDEWEB)

    None

    1999-01-01

    Two standards have been developed that pertain to the safety of fusion facilities. These are DOE- STD-6002-96, Safety of Magnetic Fusion Facilities: Requirements, and DOE-STD-6003-96, Safety of Magnetic Fusion Facilities: Guidance. The first of these standards identifies requirements that subscribers to that standard must meet to achieve safety in fusion facilities. The second standard contains guidance to assist in meeting the requirements identified in the first This handbook provides additional documentation on good operations and design practices as well as lessons learned from the experiences of designers and operators of previous fusion facilities and related systems. It is intended to capture the experience gained in the various fields and pass it on to designers of future fusion facilities as a means of enhancing success and safety. The sections of this document are presented according to the physical location of the major systems of a fusion facility, beginning with the vacuum vessel and proceeding to those systems and components outside the vacuum vessel (the "Ex-vessel Systems"). The last section describes administrative procedures that cannot be localized to specific components. It has been tacitly assumed that the general structure of the fusion facilities addressed is that of a tokamak though the same principles would apply to other magnetic confinement options.

  9. Background and planning requirements for spent fuel shipments to DOE

    Energy Technology Data Exchange (ETDEWEB)

    Ravenscroft, Norman [Edlow International Company, 1666 Connecticut Avenue, NW, Suite 201, Washington, DC 20009 (United States)

    1996-10-01

    Information is provided on the planning required and the factors that must be included in the planning process for spent fuel shipments to DOE. A summary is also provided on the background concerning renewal of the DOE spent fuel acceptance policy in May 1996. (author)

  10. Does perceptual learning require consciousness or attention?

    Science.gov (United States)

    Meuwese, Julia D I; Post, Ruben A G; Scholte, H Steven; Lamme, Victor A F

    2013-10-01

    , suggesting that the memory trace that is formed during inattention is latent until accessed. The results suggest that learning requires consciousness, and not attention, and further strengthen the idea that consciousness is separate from attention.

  11. 40 CFR 33.301 - What does this subpart require?

    Science.gov (United States)

    2010-07-01

    ... AGENCY PROGRAMS Good Faith Efforts § 33.301 What does this subpart require? A recipient, including one... good faith efforts whenever procuring construction, equipment, services and supplies under an EPA...

  12. 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. ...

  13. 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)

  14. 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).

  15. Construction safety in DOE. Part 2, Students guide

    Energy Technology Data Exchange (ETDEWEB)

    Handwerk, E.C.

    1993-08-01

    This report is the second part of a compilation of safety standards for construction activities on DOE facilities. This report covers the following areas: floor and wall openings; cranes, derricks, hoists, elevators, and conveyors; motor vehicles, mechanized equipment, and marine operations; excavations; concrete and masonry construction; steel erection; underground construction, caisson, cofferdams, and compressed air; demolition; blasting and the use of explosives; power transmission and distribution; rollover protective structures, overhead protection; and ladders.

  16. DOE Handbook: Supplementary guidance and design experience for the fusion safety standards DOE-STD-6002-96 and DOE-STD-6003-96

    International Nuclear Information System (INIS)

    1999-01-01

    Two standards have been developed that pertain to the safety of fusion facilities. These are DOE- STD-6002-96, Safety of Magnetic Fusion Facilities: Requirements, and DOE-STD-6003-96, Safety of Magnetic Fusion Facilities: Guidance. The first of these standards identifies requirements that subscribers to that standard must meet to achieve safety in fusion facilities. The second standard contains guidance to assist in meeting the requirements identified inthefirst This handbook provides additional documentation on good operations and design practices as well as lessons learned from the experiences of designers and operators of previous fusion facilities and related systems. It is intended to capture the experience gained in the various fields and pass it on to designers of future fusion facilities as a means of enhancing success and safeiy. The sections of this document are presented according to the physical location of the major systems of a t%sion facility, beginning with the vacuum vessel and proceeding to those systems and components outside the vacuum vessel (the ''Ex-vessel Systems''). The last section describes administrative procedures that cannot be localized to specific components. It has been tacitly assumed that the general structure of the fusion facilities addressed is that of a tokamak though the same principles would apply to other magnetic confinement options

  17. Cognitive screening of older drivers does not produce safety benefits

    DEFF Research Database (Denmark)

    Siren, Anu Kristiina; Meng, Annette

    2012-01-01

    Although screening policies for older drivers based on chronological age are widely used in many countries, previous research has shown that increasing age does not cause higher crash rates and that consequently, chronological age per se is at best only a weak predictor of safe driving performanc...... is an example of a political measure that intuitively makes sense, but fails to produce the desired benefits. On the contrary, on a system level, it decreases the overall safety and is connected to various direct and indirect costs........ Previous research on age-based mandatory screening of older drivers has not been able to demonstrate any safety benefits from screening measures.The present study is a population-based evaluation of the safety effects that the introduction of the cognitive test as an age-based screening tool has had...

  18. 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.)

  19. 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)

  20. 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.

  1. 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

  2. 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)

  3. 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

  4. 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

  5. A review of the safety features of 6M packagings for DOE programs

    International Nuclear Information System (INIS)

    1988-12-01

    This report, prepared by a US Department of Energy (DOE) Task Force and organized for clarity into two-page modules, argues that the US Department of Transportation (DOT) Specification-6M packagings (hereafter referred to as 6M packaging, or simply 6M) merit continued DOE use and, if necessary, DOE certification. This report is designed to address the specific requirements of a Safety Analysis Report for Packaging (SARP). While not a SARP, this report constitutes a compilation of all available documentation on 6M packagings. The authors individually, and the Task Force collectively, believe their investigation provides justification for the continued use of 6M packagings because they meet criteria for quality assurance and for safety under normal and accident conditions as defined by the US Nuclear Regulatory Commission (NRC) regulations. This report may be used by DOE managers to assist in deliberations on future requirements for 6M packagings as they are required to support DOE programs. For the purpose of ready evaluation, this report includes categorical topics found in Nuclear Regulatory Guide 7.9, the topical guideline for SARPs. The format, however, will (it is hoped) pleasantly surprise customary reader expectations. For, while maintaining categorical headings and subheadings found in SARPs as a skeleton, the Task Force chose to adopt the document design principles developed by Hughes Aircraft in the 1960s, ''Sequential Thematic Organization of Publications'' (STOP). 37 figs

  6. 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

  7. 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...

  8. 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.

  9. Review of the DOE Packaging and Transportation Safety Program

    International Nuclear Information System (INIS)

    Snyder, B.J.; Cece, J.M.

    1992-12-01

    This report documents the results of a year-long self-assessment of DOE-EH transportation and packaging safety activities. The self-assessment was initiated in September 1991 and concluded in August 1992. The self-assessment identified several significant issues, some of which have been resolved by EH. Also, improvements in the EH program were made during the course of the self-assessment. The report reflects the status of the EH transportation and packaging safety activities at the conclusion of the self-assessment. This report consists of several sections which discuss background, objectives and description of the review. Another section includes summary discussion and key conclusions. Appendix A, Issues, Observations and Recommendations, lists fifteen issues, including appropriate observations and recommendations. A Corrective Action Plan, which documents EH managements resolve to implement the agreed-upon recommendations, is included. The Corrective Action Plan reflects the status of completed and planned actions as of the date of the report

  10. 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)

  11. [Does simulator-based team training improve patient safety?].

    Science.gov (United States)

    Trentzsch, H; Urban, B; Sandmeyer, B; Hammer, T; Strohm, P C; Lazarovici, M

    2013-10-01

    Patient safety became paramount in medicine as well as in emergency medicine after it was recognized that preventable, adverse events significantly contributed to morbidity and mortality during hospital stay. The underlying errors cannot usually be explained by medical technical inadequacies only but are more due to difficulties in the transition of theoretical knowledge into tasks under the conditions of clinical reality. Crew Resource Management and Human Factors which determine safety and efficiency of humans in complex situations are suitable to control such sources of error. Simulation significantly improved safety in high reliability organizations, such as the aerospace industry.Thus, simulator-based team training has also been proposed for medical areas. As such training is consuming in cost, time and human resources, the question of the cost-benefit ratio obviously arises. This review outlines the effects of simulator-based team training on patient safety. Such course formats are not only capable of creating awareness and improvements in safety culture but also improve technical team performance and emphasize team performance as a clinical competence. A few studies even indicated improvement of patient-centered outcome, such as a reduced rate of adverse events but further studies are required in this respect. In summary, simulator-based team training should be accepted as a suitable strategy to improve patient safety.

  12. Technical Safety Requirements for the B695 Segment

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-09-11

    Ci/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, {sup 3}H. Chapter 5 of the DSA documents the derivation of TSRs and develops the operational limits that protect the safety envelope defined for this facility. The DSA is applicable to the handling of radioactive waste stored and treated in the B695 Segment. Section 5 of the TSR, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the B695 Segment. A basis explanation for each of the requirements described in Section 5.5, Specific Administrative Controls is provided in Appendix B. The basis explanation does not constitute an additional requirement, but is intended as an expansion of the logic and reasoning behind development of the requirement. Programmatic Administrative Controls are addressed in Section 5.6. This introduction to the B695 Segment TSRs is not part of the TSR limits or conditions and contains no requirements related to B695 Segment operations or to the safety analyses in the DSA.

  13. Technical Safety Requirements for the B695 Segment

    International Nuclear Information System (INIS)

    Laycak, D.

    2008-01-01

    Ci/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., 90 Sr, 137 Cs, 3 H. Chapter 5 of the DSA documents the derivation of TSRs and develops the operational limits that protect the safety envelope defined for this facility. The DSA is applicable to the handling of radioactive waste stored and treated in the B695 Segment. Section 5 of the TSR, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the B695 Segment. A basis explanation for each of the requirements described in Section 5.5, Specific Administrative Controls is provided in Appendix B. The basis explanation does not constitute an additional requirement, but is intended as an expansion of the logic and reasoning behind development of the requirement. Programmatic Administrative Controls are addressed in Section 5.6. This introduction to the B695 Segment TSRs is not part of the TSR limits or conditions and contains no requirements related to B695 Segment operations or to the safety analyses in the DSA

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. 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.

  20. Improving the regulation of safety at DOE nuclear facilities. Final report: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

    The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

  1. Improving the regulation of safety at DOE nuclear facilities. Final report

    International Nuclear Information System (INIS)

    1995-12-01

    The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by DOE itself. The three major recommendations are: under any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE's nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation

  2. 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)

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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

  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. 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)

  16. 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)

  17. 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

  18. 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

  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. 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...

  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. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); MacKinnon, Robert J. [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Leigh, Christi D. [Defense Waste Management Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Hansen, Frank D. [Geoscience Research and Applications Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

    2013-07-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  6. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    International Nuclear Information System (INIS)

    Sevougian, S. David; MacKinnon, Robert J.; Leigh, Christi D.; Hansen, Frank D.

    2013-01-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. Safety Analysis Report for Packaging (SARP) for USA/5790/BLF (DOE-AL) and USA/5791/BLF (DOE-AL)

    International Nuclear Information System (INIS)

    Roome, L.G.; Watkins, R.A.; Bertram, R.E.; Kreider, H.B.

    1980-01-01

    This revised Safety Analysis Report for Packaging (SARP) includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control of shipping containers. Much of the information was previously submitted to AEC/OSD/ALO and the Department of Transportation (DOT) and provided the basis for obtaining special permits DOT-SP-5790 and DOT-SP-5791 as well as the Interim Certificates of Compliance until the original SARP could be prepared and Certificates of Compliance issued by ERDA. This SARP revision incorporates information on certain design changes, the most significant of which relate to the inner container for the type 5790 package. Complete physical and technical descriptions of the packages are presented. Each package consists of a cylindrical steel inner container centered within an insulating steel drum assembly. The contents may be any radioactive materials which satisfy the requirements established in this SARP. A shipment of plutonium-238 in the form of a solid oxide is evaluated in this SARP as an example. The results of the nuclear criticality safety analysis show how much of the fissile isotopes may be shipped as Fissile Class I, II, or III for each container. Design and development considerations, the tests and evaluations required to prove the ability of the containers to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Mound Facility experience in using the containers, and copies of the DOE Certificates of Compliance are included. Internal reviews of the original and revised SARP's have been performed in compliance with the requirement of DOEM 5201-Part V

  12. Improving the regulation of safety at DOE nuclear facilities. Final report: Appendices

    International Nuclear Information System (INIS)

    1995-12-01

    The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE's nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation

  13. 60-day aging requirement does not ensure safety of surface-mold-ripened soft cheeses manufactured from raw or pasteurized milk when Listeria monocytogenes is introduced as a postprocessing contaminant.

    Science.gov (United States)

    D'Amico, Dennis J; Druart, Marc J; Donnelly, Catherine W

    2008-08-01

    Because of renewed interest in specialty cheeses, artisan and farmstead producers are manufacturing surface-mold-ripened soft cheeses from raw milk, using the 60-day holding standard (21 CFR 133.182) to achieve safety. This study compared the growth potential of Listeria monocytogenes on cheeses manufactured from raw or pasteurized milk and held for > 60 days at 4 degrees C. Final cheeses were within federal standards of identity for soft ripened cheese, with low moisture targets to facilitate the holding period. Wheels were surface inoculated with a five-strain cocktail of L. monocytogenes at approximately 0.2 CFU/ cm2 (low level) or 2 CFU/cm2 (high level), ripened, wrapped, and held at 4 degrees C. Listeria populations began to increase by day 28 for all treatments after initial population declines. From the low initial inoculation level, populations in raw and pasteurized milk cheese reached maximums of 2.96 +/- 2.79 and 2.33 +/- 2.10 log CFU/g, respectively, after 60 days of holding. Similar growth was observed in cheese inoculated at high levels, where populations reached 4.55 +/- 4.33 and 5.29 +/- 5.11 log CFU/g for raw and pasteurized milk cheeses, respectively. No significant differences (P milk types. Independent of the milk type, cheeses held for 60 days supported growth from very low initial levels of L. monocytogenes introduced as a postprocess contaminant. The safety of cheeses of this type must be achieved through control strategies other than aging, and thus revision of current federal regulations is warranted.

  14. 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)

  15. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  16. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    International Nuclear Information System (INIS)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

  17. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 1 of 2: Technical standard

    International Nuclear Information System (INIS)

    1998-05-01

    This Department of Energy (DOE) technical standard (referred to as the Standard) provides guidance for integrating and enhancing worker, public, and environmental protection during facility disposition activities. It provides environment, safety, and health (ES and H) guidance to supplement the project management requirements and associated guidelines contained within DOE O 430.1A, Life-Cycle Asset Management (LCAM), and amplified within the corresponding implementation guides. In addition, the Standard is designed to support an Integrated Safety Management System (ISMS), consistent with the guiding principles and core functions contained in DOE P 450.4, Safety Management System Policy, and discussed in DOE G 450.4-1, Integrated Safety Management System Guide. The ISMS guiding principles represent the fundamental policies that guide the safe accomplishment of work and include: (1) line management responsibility for safety; (2) clear roles and responsibilities; (3) competence commensurate with responsibilities; (4) balanced priorities; (5) identification of safety standards and requirements; (6) hazard controls tailored to work being performed; and (7) operations authorization. This Standard specifically addresses the implementation of the above ISMS principles four through seven, as applied to facility disposition activities

  18. 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)

  19. Preparation of Phased and Merged Safety Analysis Reports for New DOE Nuclear Facilities

    International Nuclear Information System (INIS)

    BISHOP, G.E.

    2000-01-01

    The Spent Nuclear Fuels Project (SNFP) is charged with moving to storage 2,100 metric tons of spent nuclear fuel elements left over from plutonium production at DOE'S Hanford site in Washington state. Two new facilities, the Cold Vacuum Drying Facility (CVDF) and the Canister Storage Building (CSB) are in final construction. In order to meet aggressive schedule commitments, the SNFP chose to prepare the safety analysis reports (SAR's) in phases that covered only specific portions of each facility's design as it was built. Each SAR also merged the preliminary and final safety analysis reports into a single SAR, thereby covering all aspects of design, construction, and operation for that portion (phase) of the facility. A policy of ''NRC equivalency'' was also implemented in parallel with this effort, with the goal of achieving a rigor of safety analysis equivalent to that of NRC-licensed fuel processing facilities. DOE Order 5480.23. ''Nuclear Safety Analysis Reports'' allows preparation of both a phased and a merged SAR to accelerate construction schedules. However, project managers must be aware that such acceleration is not guaranteed. Managers considering this approach for their project should be cognizant of numerous obstacles that will be encountered. Merging and phasing SAR's will create new, unique, and unanticipated difficulties which may actually slow construction unless expeditiously and correctly managed. Pitfalls to be avoided and good practices to be implemented in preparing phased and merged SAR's are presented. The value of applying NRC requirements to the DOE safety analysis process is also discussed. As of December, 1999, the SNFP has completed and approved a SAR for the CVDF. Approval of the SAR for the CSB is pending

  20. Perspectives on nuclear material safety management methods at DOE sites

    International Nuclear Information System (INIS)

    Hyder, M.L.

    1997-01-01

    The management of nuclear materials, and fissile materials in particular, at the USDOE facilities is undergoing significant changes. These result in large part from decreasing requirements for these materials in the US weapons program. Not only is new production no longer required, but returns must be handled and safely stored. Eventually surplus fissile material will be used for power production, or else put into a form suitable for long term disposition. In the meanwhile concentrates must be stored with protection against releases of radioactive material to the environment, and also against theft or deliberate dispersion. In addition, cleaning up large volumes of materials contaminated with fissile isotopes will be a major activity, and there will also be some quantity of spent fuel containing enriched uranium that cannot readily be processed. All these activities pose safety problems, some of which are addressed here

  1. 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

  2. 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

  3. 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 ...

  4. 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...

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  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. 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.

  15. Student manual, Book 2: Orientation to occupational safety compliance in DOE

    Energy Technology Data Exchange (ETDEWEB)

    Colley, D.L.

    1993-10-01

    This is a student hand-book an Occupational Safety Compliance in DOE. Topics include the following: Electrical; materials handling & storage; inspection responsibilities & procedures; general environmental controls; confined space entry; lockout/tagout; office safety, ergonomics & human factors; medical & first aid, access to records; construction safety; injury/illness reporting system; and accident investigation procedures.

  16. Student manual, Book 2: Orientation to occupational safety compliance in DOE

    International Nuclear Information System (INIS)

    Colley, D.L.

    1993-01-01

    This is a student hand-book an Occupational Safety Compliance in DOE. Topics include the following: Electrical; materials handling ampersand storage; inspection responsibilities ampersand procedures; general environmental controls; confined space entry; lockout/tagout; office safety, ergonomics ampersand human factors; medical ampersand first aid, access to records; construction safety; injury/illness reporting system; and accident investigation procedures

  17. International inspection activity impacts upon DOE safeguards requirements

    International Nuclear Information System (INIS)

    Zack, N.R.

    1995-01-01

    The US has placed certain special nuclear materials declared excess to their strategic needs under international safeguards through the International Atomic Energy Agency (IAEA). This Presidential initiative has obligated materials at several Department of Energy (DOE) facilities for these safeguards activities to demonstrate the willingness of the US to ban production or use of nuclear materials outside of international safeguards. However, IAEA inspection activities generally tend to be intrusive in nature and are not consistent with several domestic safeguards procedures implemented to reduce worker radiation exposures and increase the cost-effectiveness and efficiency of accounting for and storing of special nuclear materials. To help identify and provide workable solutions to these concerns, the Office of Safeguards and Security has conducted a program to determine possible changes to the DOE safeguards and security requirements designed to help facilities under international safeguards inspections more easily comply with domestic safeguards goals during international inspection activities. This paper will discuss the impact of international inspection activities on facility safeguards operations and departmental safeguards procedures and policies

  18. 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)

  19. 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

  20. 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

  1. 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)

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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

  7. 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.

  8. American National Standards and the DOE - A cooperative effort to promote nuclear criticality safety

    International Nuclear Information System (INIS)

    Rothleder, B.M.

    1996-01-01

    The U.S. Department of Energy's (DOE's) new criticality safety order, DOE Order 420.1 (open-quotes Facility Safety,close quotes October 13, 1995), Sec. 4.3 (open-quotes Nuclear Criticality Safetyclose quotes), invokes, as an integral part, 12 appropriate American National Standards Institute/American Nuclear Society (ANSI/ANS) Series-8 standards for nuclear criticality safety, but with modifications. (The order that 420.1/4.3 replaced also invoked some ANSI/ANS Series-8 standards.) These modifications include DOE operation-specific exceptions to the standards and elaborations on some of the wording in the standards

  9. 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

  10. 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

  11. Radiological safety training for accelerator facilities: DOE handbook

    International Nuclear Information System (INIS)

    1997-03-01

    This program management guide describes the proper implementation standard for core training as outline in the DOE Radiological Control (RadCon) Manual. Its purpose is to assist DOE employees and Managing and Operating (M ampersand O) contractors having responsibility for implementing the core training recommended by the RadCon Manual

  12. Radiological safety training for accelerator facilities: DOE handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This program management guide describes the proper implementation standard for core training as outline in the DOE Radiological Control (RadCon) Manual. Its purpose is to assist DOE employees and Managing and Operating (M&O) contractors having responsibility for implementing the core training recommended by the RadCon Manual.

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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.

  18. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-01

    This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist.

  19. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1998-05-01

    This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist

  20. 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

  1. 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

  2. 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

  3. 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

  4. does road safety projects relate to community capacity building?

    African Journals Online (AJOL)

    USER

    The AARSI organization is currently based in Lagos State and its mission is to ... organizational structures, resources and commitments to health promotion, .... donated breathalyzer equipment worth of $90,000 to Road Safety Personnel as at.

  5. 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

  6. 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...

  7. 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

  8. 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

  9. Does the concept of safety culture help or hinder systems thinking in safety?

    Science.gov (United States)

    Reiman, Teemu; Rollenhagen, Carl

    2014-07-01

    The concept of safety culture has become established in safety management applications in all major safety-critical domains. The idea that safety culture somehow represents a "systemic view" on safety is seldom explicitly spoken out, but nevertheless seem to linger behind many safety culture discourses. However, in this paper we argue that the "new" contribution to safety management from safety culture never really became integrated with classical engineering principles and concepts. This integration would have been necessary for the development of a more genuine systems-oriented view on safety; e.g. a conception of safety in which human, technological, organisational and cultural factors are understood as mutually interacting elements. Without of this integration, researchers and the users of the various tools and methods associated with safety culture have sometimes fostered a belief that "safety culture" in fact represents such a systemic view about safety. This belief is, however, not backed up by theoretical or empirical evidence. It is true that safety culture, at least in some sense, represents a holistic term-a totality of factors that include human, organisational and technological aspects. However, the departure for such safety culture models is still human and organisational factors rather than technology (or safety) itself. The aim of this paper is to critically review the various uses of the concept of safety culture as representing a systemic view on safety. The article will take a look at the concepts of culture and safety culture based on previous studies, and outlines in more detail the theoretical challenges in safety culture as a systems concept. The paper also presents recommendations on how to make safety culture more systemic. Copyright © 2013 Elsevier Ltd. All rights reserved.

  10. Educational status and organizational safety climate: does educational attainment influence workers' perceptions of workplace safety?

    OpenAIRE

    Gyekye, Seth; Salminen, Simo

    2009-01-01

    From a practical perspective, understanding the impact of education on perceptions of workplace safety would benefit management’s decisions regarding workers’ adaptability, general work effectiveness, accident frequency, implementation of safety management policies, and handling of education-related accident characteristics. The current study thus examined the relationship between educational attainment and (i) safety perception, (ii) job satisfaction, (iii) compliance with safety management ...

  11. 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)

  12. 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)

  13. 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

  14. [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.

  15. 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.

  16. 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

  17. Does Employee Safety Matter for Patients Too? Employee Safety Climate and Patient Safety Culture in Health Care.

    Science.gov (United States)

    Mohr, David C; Eaton, Jennifer Lipkowitz; McPhaul, Kathleen M; Hodgson, Michael J

    2015-04-22

    We examined relationships between employee safety climate and patient safety culture. Because employee safety may be a precondition for the development of patient safety, we hypothesized that employee safety culture would be strongly and positively related to patient safety culture. An employee safety climate survey was administered in 2010 and assessed employees' views and experiences of safety for employees. The patient safety survey administered in 2011 assessed the safety culture for patients. We performed Pearson correlations and multiple regression analysis to examine the relationships between a composite measure of employee safety with subdimensions of patient safety culture. The regression models controlled for size, geographic characteristics, and teaching affiliation. Analyses were conducted at the group level using data from 132 medical centers. Higher employee safety climate composite scores were positively associated with all 9 patient safety culture measures examined. Standardized multivariate regression coefficients ranged from 0.44 to 0.64. Medical facilities where staff have more positive perceptions of health care workplace safety climate tended to have more positive assessments of patient safety culture. This suggests that patient safety culture and employee safety climate could be mutually reinforcing, such that investments and improvements in one domain positively impacts the other. Further research is needed to better understand the nexus between health care employee and patient safety to generalize and act upon findings.

  18. 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

  19. 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

  20. 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

  1. 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.

  2. 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)

  3. 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)

  4. 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...

  5. 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

  6. Does external evaluation of laboratories improve patient safety?

    Science.gov (United States)

    Noble, Michael A

    2007-01-01

    Laboratory accreditation and External Quality Assessment (also called proficiency testing) are mainstays of laboratory quality assessment and performance. Both practices are associated with examples of improved laboratory performance. The relationship between laboratory performance and improved patient safety is more difficult to assess because of the many variables that are involved with patient outcome. Despite this difficulty, the argument to continue external evaluation of laboratories is too compelling to consider the alternative.

  7. 76 FR 16758 - DOE Response to Recommendation 2010-1 of the Defense Nuclear Facilities Safety Board, Safety...

    Science.gov (United States)

    2011-03-25

    ... Recommendation; a detailed explanation is provided below. We have clarified aspects of sub-recommendation 1, 2.... Sub-recommendation 1--Immediately affirm the requirement that unmitigated, bounding-type accident... details DOE's expectations for accident analyses to identify hazard controls for most DOE nuclear...

  8. 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.

  9. 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.

  10. 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

  11. DOE Safety Metrics Indicator Program (SMIP) Fiscal Year 2000 Annual Report of Packaging- and Transportation-related Occurrences

    International Nuclear Information System (INIS)

    Dickerson, L.S.

    2001-01-01

    The U.S. Department of Energy (DOE) Occurrence Reporting and Processing System (ORPS) is an interactive computer system designed to support DOE-owned or -operated facilities in reporting and processing information concerning occurrences related to facility operations. The Oak Ridge National Laboratory has been charged by the DOE National Transportation Program Albuquerque (NTPA) with the responsibility of retrieving reports and information pertaining to packaging and transportation (P and T) incidents from the centralized ORPS database. These selected reports are analyzed for safety concerns, trends, potential impact on P and T operations, and ''lessons learned'' in P and T safety. To support this analysis and trending, the Safety Metrics Indicator Program (SMIP) was established by the NTPA in fiscal year (FY) 1998. Its chief goal is to augment historical reporting of occurrence-based information by providing (1) management notification of those incidents that require attention, (2) an accurate picture of contractors' P and T-related performance, and (3) meaningful statistics on occurrences at particular sites, including comparisons among different contractor sites and between DOE and the private sector. This annual report contains information on those P and T-related occurrences reported to the ORPS during the period from October 1, 1999, through September 30, 2000. Only those incidents that occur in preparation for transport, during transport, and during unloading of hazardous material are considered as packaging- or transportation-related occurrences

  12. 25 CFR 42.6 - When does due process require a formal disciplinary hearing?

    Science.gov (United States)

    2010-04-01

    ... 25 Indians 1 2010-04-01 2010-04-01 false When does due process require a formal disciplinary... RIGHTS § 42.6 When does due process require a formal disciplinary hearing? Unless local school policies and procedures provide for less, a formal disciplinary hearing is required before a suspension in...

  13. 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

  14. 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

  15. 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

  16. DOE high-level waste tank safety program Final report, Task 002

    International Nuclear Information System (INIS)

    1998-01-01

    The overall objective of the work on Task 002 was to provide LANL with support to the DOE High-Level Waste Tank Safety program. The objective of the work was to develop safety documentation in support of the unsafe tank mitigation activities at Hanford. The work includes the development of safety assessment and an environmental assessment. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective are provided. The two tasks were: Task 2.1--safety assessment for instrumentation insertion; and Task 2.2--environmental assessment

  17. 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...

  18. 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.

  19. DOE contractor radiation safety CBT [computer based training] course

    International Nuclear Information System (INIS)

    Gardner, P.R.

    1986-01-01

    Westinghouse Hanford Company developed a generic Radiation Worker safety CBT course for Department of Energy contractors. Task analysis concentrated on actual and potential tasks and included visits to fourteen different contractor sites. Team Design and Prototype verification formed the major portion of the development phase. Lesson entry was accomplished using the WISE author system from WICAT Systems, Inc. The course features graded task simulations for both Pretest and Final; fourteen Topics in five Lessons, each Topic keyed to ''Critical Acts'' and Questions in the Pretest and Final; Automatic, Intensive, and Manual modes of instruction available for each Lesson; Practical Problems and Sample Questions associated with each Topic; and provisions for local configuration in several areas. The course is deliverable on IBM PC compatible equipment. 2 refs

  20. 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

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

    International Nuclear Information System (INIS)

    1992-03-01

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

  2. Child Passenger Safety Training for Pediatric Interns: Does it Work?

    Science.gov (United States)

    Morrissey, Dina; Riese, Alison; Violano, Pina; Lapidus, Garry; Baird, Janette; Mello, Michael J

    2016-03-01

    Evaluate the efficacy of a child passenger safety (CPS) educational intervention on the CPS-related knowledge, attitude and anticipatory guidance behaviors of pediatric interns. All subjects were surveyed at baseline and 6 months. Intervention interns attended a CPS training module which included viewing an educational video, observing a car seat inspection appointment, hands-on practice and completion of a post-intervention survey. All 16 intervention interns completed the initial survey, the intervention and the immediate-post questionnaire. Thirteen (81%) completed the 6-month follow-up. The baseline survey was completed by 27/40 (67%) of control interns, 28/40 (70%) submitted a follow-up. The proportion of intervention interns who self-reported giving CPS guidance at all well-child visits increased by 31.3% (95% CI 6.1,56.5%); the control group had no change. Similar results were seen with self-reported knowledge and attitude. A CPS training module increases pediatric interns' knowledge, improves attitudes, and self-reported behaviors regarding CPS-related anticipatory guidance.

  3. 12 CFR 617.7630 - Does this Federal requirement affect any state property laws?

    Science.gov (United States)

    2010-01-01

    ... 12 Banks and Banking 6 2010-01-01 2010-01-01 false Does this Federal requirement affect any state property laws? 617.7630 Section 617.7630 Banks and Banking FARM CREDIT ADMINISTRATION FARM CREDIT SYSTEM BORROWER RIGHTS Right of First Refusal § 617.7630 Does this Federal requirement affect any state property...

  4. 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

  5. 78 FR 9902 - DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford...

    Science.gov (United States)

    2013-02-12

    ... Safety Board, Hanford Tank Farms Flammable Gas Safety Strategy; Correction AGENCY: Department of Energy. ACTION: Notice; Correction SUMMARY: The Department of Energy (DOE) published a document in the Federal... Facilities Safety Board, Hanford Tank Farms Flammable Gas Safety Strategy. This document corrects an error in...

  6. 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

  7. 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

  8. Environment, safety, and health. Status of DOE's reorganization of its safety oversight function

    International Nuclear Information System (INIS)

    Bannerman, Carl J.; Cannon, Doris E.; Jones, Gary L.; Ulrich, Timothy W.

    1990-01-01

    Several major events that preceded the Secretary's decision to restructure DOE's management of its nuclear facilities were identified. The proposed restructuring plan, in concept, is designed to set in place an oversight framework, which will provide confidence in DOE's ability to operate its nuclear facilities in a safe manner. Further, on the basis of the previous work in this area as well as other independent studies, several issues were identified that may affect the success of the proposed restructuring plan

  9. 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

  10. Cytochrome oxidase assembly does not require catalytically active cytochrome C.

    Science.gov (United States)

    Barrientos, Antoni; Pierre, Danielle; Lee, Johnson; Tzagoloff, Alexander

    2003-03-14

    Cytochrome c oxidase (COX), the terminal enzyme of the mitochondrial respiratory chain, catalyzes the transfer of electrons from reduced cytochrome c to molecular oxygen. COX assembly requires the coming together of nuclear- and mitochondrial-encoded subunits and the assistance of a large number of nuclear gene products acting at different stages of maturation of the enzyme. In Saccharomyces cerevisiae, expression of cytochrome c, encoded by CYC1 and CYC7, is required not only for electron transfer but also for COX assembly through a still unknown mechanism. We have attempted to distinguish between a functional and structural requirement of cytochrome c in COX assembly. A cyc1/cyc7 double null mutant strain was transformed with the cyc1-166 mutant gene (Schweingruber, M. E., Stewart, J. W., and Sherman, F. (1979) J. Biol. Chem. 254, 4132-4143) that expresses stable but catalytically inactive iso-1-cytochrome c. The COX content of the cyc1/cyc7 double mutant strain harboring non-functional iso-1-cytochrome c has been characterized spectrally, functionally, and immunochemically. The results of these studies demonstrate that cytochrome c plays a structural rather than functional role in assembly of cytochrome c oxidase. In addition to its requirement for COX assembly, cytochrome c also affects turnover of the enzyme. Mutants containing wild type apocytochrome c in mitochondria lack COX, suggesting that only the folded and mature protein is able to promote COX assembly.

  11. 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.

  12. 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.

  13. 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

  14. 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)

  15. Does this adult patient with suspected bacteremia require blood cultures?

    Science.gov (United States)

    Coburn, Bryan; Morris, Andrew M; Tomlinson, George; Detsky, Allan S

    2012-08-01

    Clinicians order blood cultures liberally among patients in whom bacteremia is suspected, though a small proportion of blood cultures yield true-positive results. Ordering blood cultures inappropriately may be both wasteful and harmful. To review the accuracy of easily obtained clinical and laboratory findings to inform the decision to obtain blood cultures in suspected bacteremia. A MEDLINE and EMBASE search (inception to April 2012) yielded 35 studies that met inclusion criteria for evaluating the accuracy of clinical variables for bacteremia in adult immunocompetent patients, representing 4566 bacteremia and 25,946 negative blood culture episodes. Data were extracted to determine the prevalence and likelihood ratios (LRs) of findings for bacteremia. The pretest probability of bacteremia varies depending on the clinical context, from low (eg, cellulitis: 2%) to high (eg, septic shock: 69%). Elevated temperatures alone do not accurately predict bacteremia (for ≥38°C [>100.3°F], LR, 1.9 [95% CI, 1.4-2.4]; for ≥38.5°C [>101.2°F], LR, 1.4 [95% CI, 1.1-2.0]), nor does isolated leukocytosis (LR, cultures should not be ordered for adult patients with isolated fever or leukocytosis without considering the pretest probability. SIRS and the decision rule may be helpful in identifying patients who do not need blood cultures. These conclusions do not apply to immunocompromised patients or when endocarditis is suspected.

  16. What does it require to open a franchising clothing business?

    OpenAIRE

    Ranta, Heidi-Helena; Pikarinen, Helmi

    2013-01-01

    The purpose of the study was to find out how to open a franchising clothing business in Finland and what is required from the new franchisee. The goal of this thesis work was to make preparation work for a potential future business. This thesis was to find out the possibility if a franchising business could be a possible future career for the authors. The information for this thesis was gathered from literature, the Internet and by interviewing the franchisees and franchisors from differe...

  17. IEEE Std 600: IEEE trial-use standard requirements for organizations that conduct qualification testing of safety systems equipment for use in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The purpose of this standard is to provide requirements for establishing a program for conducting qualification tests of safety systems equipment used in nuclear power generating stations. Compliance with the requirements of this standard does not assure the adequacy of the qualification tests performed. This standard applies to organizations that conduct qualification tests on equipment that has a definable safety function and is an identifiable part of a safety system for use in nuclear power generating stations. It requires a technical program, a quality assurance program, and a demonstrated ability to meet specified technical requirements. It does not apply to materials tests, production tests, normal performance testing, qualification by analysis, qualification by operating experience, or reliability tests such as diesel-generator multiple start tests. The intent of this standard is to achieve greater consistency, reliability, and reproducibility of test results and to provide adequate control of qualification testing of safety systems equipment

  18. 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

  19. 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

  20. 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)

  1. 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

  2. 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

  3. Requirements for shipment of DOE radioactive mixed waste

    International Nuclear Information System (INIS)

    Gablin, K.; No, Hyo; Herman, J.

    1993-01-01

    There are several sources of radioactive mixed waste (RMW) at Argonne National Laboratory which, in the past, were collected at waste tanks and/or sludge tanks. They were eventually pumped out by special pumps and processed in an evaporator located in the waste operations area in Building No. 306. Some of this radioactive mixed waste represents pure elementary mercury. These cleaning tanks must be manually cleaned up because the RMW material was too dense to pump with the equipment in use. The four tanks being discussed in this report are located in Building No. 306. They are the Acid Waste Tank, IMOX/FLOC Tanks, Evaporation Feed Tanks, and Waste Storage Tanks. All of these tanks are characterized and handled separately. This paper discusses the process and the requirements for characterization and the associated paperwork for Argonne Waste to be shipped to Westinghouse Hanford Company for storage

  4. Core Problem: Does the CV Parent Body Magnetization require differentiation?

    Science.gov (United States)

    O'Brien, T.; Tarduno, J. A.; Smirnov, A. V.

    2016-12-01

    Evidence for the presence of past dynamos from magnetic studies of meteorites can provide key information on the nature and evolution of parent bodies. However, the suggestion of a past core dynamo for the CV parent body based on the study of the Allende meteorite has led to a paradox: a core dynamo requires differentiation, evidence for which is missing in the meteorite record. The key parameter used to distinguish core dynamo versus external field mechanisms is absolute field paleointensity, with high values (>>1 μT) favoring the former. Here we explore the fundamental requirements for absolute field intensity measurement in the Allende meteorite: single domain grains that are non-interacting. Magnetic hysteresis and directional data define strong magnetic interactions, negating a standard interpretation of paleointensity measurements in terms of absolute paleofield values. The Allende low field magnetic susceptibility is dominated by magnetite and FeNi grains, whereas the magnetic remanence is carried by an iron sulfide whose remanence-carrying capacity increases with laboratory cycling at constant field values, indicating reordering. The iron sulfide and FeNi grains are in close proximity, providing mineralogical context for interactions. We interpret the magnetization of Allende to record the intense early solar wind with metal-sulfide interactions amplifying the field, giving the false impression of a higher field value in some prior studies. An undifferentiated CV parent body is thus compatible with Allende's magnetization. Early solar wind magnetization should be the null hypothesis for evaluating the source of magnetization for chondrites and other meteorites.

  5. Does multifocal papillary micro-carcinoma require radioiodine ablation?

    International Nuclear Information System (INIS)

    Punda, A.; Markovic, V.; Eterovic, D.

    2015-01-01

    Full text of publication follows. Background: the thyroid carcinomas smaller than 1 cm (micro-carcinomas) comprise a significant fraction of papillary carcinomas. Excluding clinical micro-carcinomas, which present as metastatic disease, the micro-carcinomas diagnosed by ultrasound/FNAC or incidentally have very good prognosis. However, whether or not these papillary micro-carcinomas require post-surgical radioiodine ablation remains a matter of debate. Hypothesis: multi-focality is present in majority of clinical papillary micro-carcinomas and this characteristic can be used to identify the subset of non-clinical micro-carcinomas with greater malignant potential. Methods: the data on types of differentiated thyroid carcinomas diagnosed in the period 2008-2011 in the University Hospital Split were collected. Results: there were 359 patients with thyroid carcinoma, 329 (92%) of which had papillary carcinoma. About 61% (202/329) of papillary carcinomas were micro-carcinomas; most of them were diagnosed by ultrasound/FNAC (134/202= 66%), the rest were incidentalomas (48/202=24%) and clinical micro carcinomas (20/202=10%). Sixty percent (12/20) of patients with clinical micro-carcinoma and 23 patients with non-clinical micro-carcinoma (23/182=13%) had multifocal disease. Conclusion: multifocal disease is a frequent characteristic of clinical papillary thyroid micro-carcinomas, suggesting that multi-focality presents an early stage of non-clinical micro-carcinomas with more aggressive behaviour. Thus multifocal, but not uni-focal papillary micro-carcinomas may require radioiodine ablation. (authors)

  6. US Department of Energy DOE Nevada Operations Office, Nevada Test Site: Underground safety and health standards

    International Nuclear Information System (INIS)

    1993-05-01

    The Nevada Test Site Underground Safety and Health Standards Working Group was formed at the direction of John D. Stewart, Director, Nevada Test Site Office in April, 1990. The objective of the Working Group was to compile a safety and health standard from the California Tunnel Safety Orders and OSHA for the underground operations at the NTS, (excluding Yucca Mountain). These standards are called the NTS U/G Safety and Health Standards. The Working Group submits these standards as a RECOMMENDATION to the Director, NTSO. Although the Working Group considers these standards to be the most integrated and comprehensive standards that could be developed for NTS Underground Operations, the intent is not to supersede or replace any relevant DOE orders. Rather the intent is to collate the multiple safety and health references contained in DOE Order 5480.4 that have applicability to NTS Underground Operations into a single safety and heath standard to be used in the underground operations at the NTS. Each portion of the standard was included only after careful consideration by the Working Group and is judged to be both effective and appropriate. The specific methods and rationale used by the Working Group are outlined as follows: The letter from DOE/HQ, dated September 28, 1990 cited OSHA and the CTSO as the safety and health codes applicable to underground operations at the NTS. These mandated codes were each originally developed to be comprehensive, i.e., all underground operations of a particular type (e.g., tunnels in the case of the CTSO) were intended to be adequately regulated by the appropriate code. However, this is not true; the Working Group found extensive and confusing overlap in the codes in numerous areas. Other subjects and activities were addressed by the various codes in cursory fashion or not at all

  7. US Department of Energy DOE Nevada Operations Office, Nevada Test Site: Underground safety and health standards

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    The Nevada Test Site Underground Safety and Health Standards Working Group was formed at the direction of John D. Stewart, Director, Nevada Test Site Office in April, 1990. The objective of the Working Group was to compile a safety and health standard from the California Tunnel Safety Orders and OSHA for the underground operations at the NTS, (excluding Yucca Mountain). These standards are called the NTS U/G Safety and Health Standards. The Working Group submits these standards as a RECOMMENDATION to the Director, NTSO. Although the Working Group considers these standards to be the most integrated and comprehensive standards that could be developed for NTS Underground Operations, the intent is not to supersede or replace any relevant DOE orders. Rather the intent is to collate the multiple safety and health references contained in DOE Order 5480.4 that have applicability to NTS Underground Operations into a single safety and heath standard to be used in the underground operations at the NTS. Each portion of the standard was included only after careful consideration by the Working Group and is judged to be both effective and appropriate. The specific methods and rationale used by the Working Group are outlined as follows: The letter from DOE/HQ, dated September 28, 1990 cited OSHA and the CTSO as the safety and health codes applicable to underground operations at the NTS. These mandated codes were each originally developed to be comprehensive, i.e., all underground operations of a particular type (e.g., tunnels in the case of the CTSO) were intended to be adequately regulated by the appropriate code. However, this is not true; the Working Group found extensive and confusing overlap in the codes in numerous areas. Other subjects and activities were addressed by the various codes in cursory fashion or not at all.

  8. 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)

  9. 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.

  10. 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

  11. What does respect for the patient's autonomy require?

    Science.gov (United States)

    Cheng, Kam-Yuen

    2013-11-01

    Personal autonomy presupposes the notion of rationality. What is not so clear is whether, and how, a compromise of rationality to various degrees will diminish a person's autonomy. In bioethical literature, three major types of threat to the rationality of a patient's medical decision are identified: insufficient information, irrational beliefs/desires, and influence of different framing effects. To overcome the first problem, it is suggested that patients be provided with information about their diseases and treatment choices according to the objective standard. I shall explain how this should be finessed. Regarding the negative impact of irrational beliefs/desires, some philosophers have argued that holding irrational beliefs can still be an expression of autonomy. I reject this argument because the degree of autonomy of a decision depends on the degree of rationality of the beliefs or desires on which the decision is based. Hence, to promote patient autonomy, we need to eliminate irrational beliefs by the provision of evidence and good arguments. Finally, I argue that the way to smooth out the framing effects is to present the same information in different perspectives: it is too often assumed that medical information can always be given in a complete and unadorned manner. This article concludes with a cautionary note that the protection of patient autonomy requires much more time and effort than the current practice usually allows. © 2012 John Wiley & Sons Ltd.

  12. Extinction of relapsed fear does not require the basolateral amygdala.

    Science.gov (United States)

    Lingawi, Nura W; Westbrook, R Frederick; Laurent, Vincent

    2017-03-01

    It is well established that extinguished fears are restored with the passage of time or a change in physical context. These fear restoration phenomena are believed to mimic the conditions under which relapse occurs in patients that have been treated for anxiety disorders by means of cue-exposure therapy. Here, we used a rodent model to extinguish relapsed fear and assess whether this new extinction prevents further relapse. We found that activity in the basolateral amygdala (BLA) is required to initially extinguish conditioned fear, but this activity was not necessary to subsequently extinguish relapsed fear. That is, extinction of spontaneously recovered or renewed fear was spared by BLA inactivation. Yet, this BLA-independent learning of extinction did not protect against further relapse: extinction of relapsed fear conducted without BLA activity was still likely to return after the passage of time or a shift in physical context. These findings have important clinical implications. They indicate that pharmacological agents with anxiolytic properties may disrupt initial cue-exposure therapy but may be useful when therapy is again needed due to relapse. However, they also suggest that these agents will not protect against further relapse, implying the need for developing drugs that target other brain regions involved in fear inhibition. Copyright © 2017 Elsevier Inc. All rights reserved.

  13. Richland Operations (DOE-RL) Environmental Safety Health (ES and H) FY 2000 and FY 2001 Execution Commitment Summary

    Energy Technology Data Exchange (ETDEWEB)

    REEP, I.E.

    2000-12-01

    All sites in the U.S. Department of Energy (DOE) Complex prepare this report annually for the DOE Office of Environment, Safety and Health (EH). The purpose of this report is to provide a summary of the previous and current year's Environment, Safety and Health (ES&H) execution commitments and the Safety and Health (S&H) resources that support these activities. The fiscal year (FY) 2000 and 2001 information and data contained in the Richland Operations Environment, Safefy and Health Fiscal Year 2002 Budget-Risk Management Summary (RL 2000a) were the basis for preparing this report. Fiscal year 2001 activities are based on the President's Amended Congressional Budget Request of $689.6 million for funding Ofice of Environmental Management (EM) $44.0 million for Fast Flux Test Facility standby less $7.0 million in anticipated DOE, Headquarters holdbacks for Office of Nuclear Energy, Science and Technology (NE); and $55.3 million for Safeguards and Security (SAS). Any funding changes as a result of the Congressional appropriation process will be reflected in the Fiscal Year 2003 ES&H Budget-Risk Management Summary to be issued in May 2001. This report provides the end-of-year status of FY 2000 ES&H execution commitments, including actual S&H expenditures, and describes planned FY 2001 ES&H execution commitments and the S&H resources needed to support those activities. This requirement is included in the ES&H guidance contained in the FY 2002 Field Budget Call (DOE 2000).

  14. 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

  15. 45 CFR 261.16 - Does the imposition of a penalty affect an individual's work requirement?

    Science.gov (United States)

    2010-10-01

    ... OF HEALTH AND HUMAN SERVICES ENSURING THAT RECIPIENTS WORK What Are the Provisions Addressing Individual Responsibility? § 261.16 Does the imposition of a penalty affect an individual's work requirement... 45 Public Welfare 2 2010-10-01 2010-10-01 false Does the imposition of a penalty affect an...

  16. 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

  17. Development of safety-related regulatory requirements for nuclear power in developing countries. Key issue paper no. 4

    International Nuclear Information System (INIS)

    Han, K.I.

    2000-01-01

    In implementing a national nuclear power program, balanced regulatory requirements are necessary to ensure nuclear safety and cost competitive nuclear power, and to help gain public acceptance. However, this is difficult due to the technology-intensive nature of the nuclear regulatory requirements, the need to reflect evolving technology and the need for cooperation among multidisciplinary technical groups. This paper suggests approaches to development of balanced nuclear regulatory requirements in developing countries related to nuclear power plant safety, radiation protection and radioactive waste management along with key technical regulatory issues. It does not deal with economic or market regulation of electric utilities using nuclear power. It suggests that national regulatory requirements be developed using IAEA safety recommendations as guidelines and safety requirements of the supplier country as a main reference after careful planning, manpower buildup and thorough study of international and supplier country's regulations. Regulation making is not recommended before experienced manpower has been accumulated. With an option that the supplier country's regulations may be used in the interim, the lack of complete national regulatory requirements should not deter introduction of nuclear power in developing countries. (author)

  18. 40 CFR 141.571 - What records does subpart T require my system to keep?

    Science.gov (United States)

    2010-07-01

    ... records does subpart T require my system to keep? Your system must keep several types of records based on... Benchmarking(§§ 141.540-141.544) Benchmark (including raw data and analysis) Indefinitely. ...

  19. 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...

  20. 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...

  1. Quantifying and Addressing the DOE Material Reactivity Requirements with Analysis and Testing of Hydrogen Storage Materials & Systems

    Energy Technology Data Exchange (ETDEWEB)

    Khalil, Y. F. [United Technologies Research Center (UTRC), East Hartford, CT (United States)

    2012-04-30

    The objective of this project is to examine safety aspects of candidate hydrogen storage materials and systems being developed in the DOE Hydrogen Program. As a result of this effort, the general DOE safety target will be given useful meaning by establishing a link between the characteristics of new storage materials and the satisfaction of safety criteria. This will be accomplished through the development and application of formal risk analysis methods, standardized materials testing, chemical reactivity characterization, novel risk mitigation approaches and subscale system demonstration. The project also will collaborate with other DOE and international activities in materials based hydrogen storage safety to provide a larger, highly coordinated effort.

  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. 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)

  4. Cross-index to DOE-prescribed occupational safety codes and standards

    International Nuclear Information System (INIS)

    1982-01-01

    A compilation of detailed information from more than three hundred and fifty DOE-prescribed or OSHA-referenced industrial safety codes and standards is presented. Condensed data from individual code portions are listed according to reference code, section, paragraph and page. A glossary of letter initials/abbreviations for the organizations or documents whose codes or standards are contained in this Cross-Index, is listed

  5. 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)

  6. 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...

  7. 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...

  8. 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

  9. Evaluation of replacement tritium facility (RTF) compliance with DOE safety goals using probabilistic consequence assessment methodology

    International Nuclear Information System (INIS)

    O'Kula, K.R.; East, J.M.; Moore, M.L.

    1993-01-01

    The Savannah River Site (SRS), operated by the Westinghouse Savannah River Company (WSRC) for the US Department of Energy (DOE), is a major center for the processing of nuclear materials for national defense, deep-space exploration, and medical treatment applications in the United States. As an integral part of the DOE's effort to modernize facilities, implement improved handling and processing technology, and reduce operational risk to the general public and onsite workers, transition of tritium processing at SRS from the Consolidated Tritium Facility to the Replacement Tritium Facility (RTF) began in 1993. To ensure that operation of new DOE facilities such as RTF present minimum involuntary and voluntary risks to the neighboring public and workers, indices of risk have been established to serve as target levels or safety goals of performance for assessing nuclear safety. These goals are discussed from a historical perspective in the initial part of this paper. Secondly, methodologies to quantify risk indices are briefly described. Lastly, accident, abnormal event, and normal operation source terms from RTF are evaluated for consequence assessment purposes relative to the safety targets

  10. 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.

  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. 30 CFR 250.195 - What notification does MMS require on the production status of wells?

    Science.gov (United States)

    2010-07-01

    ... the well in a production status. You must confirm oral notification by telefax or e-mail within those... production status of wells? 250.195 Section 250.195 Mineral Resources MINERALS MANAGEMENT SERVICE, DEPARTMENT... Information and Reporting Requirements § 250.195 What notification does MMS require on the production status...

  15. 13 CFR 123.11 - Does SBA require collateral for any of its disaster loans?

    Science.gov (United States)

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Does SBA require collateral for any of its disaster loans? 123.11 Section 123.11 Business Credit and Assistance SMALL BUSINESS...? Generally, SBA will not require that you pledge collateral to secure a disaster home loan or a physical...

  16. 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

  17. 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)

  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. 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

  20. 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

  1. 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

  2. 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.

  3. Integrating total quality management principles with the requirements of DOE Order 5700.6C

    Energy Technology Data Exchange (ETDEWEB)

    Hedges, D. [Scientific Ecology Group, Inc. (United States)

    1993-03-01

    The Department of Energy has recently required its field offices, contractors, and subcontractors to implement DOE Order 5700.6C, ``Quality Assurance,`` for all work on waste management contracts. The order restructures the 18 criteria of NQA-1 and focuses on the role of management in achieving and assuring quality, performance of activities to achieve and assure quality, and management`s assessment of its performance for the purpose of identifying improvements to be made. The DOE order also introduces elements of the total quality management (TQM) philosophy, which were not present in DOE Order 5700.6B. The research community within DOE has recently issued a document entitled DOE Order 5700.6C Implementation Guide, which is more explicit about the integration of TQM principles with the implementation of DOE Order 5700.6C in research facilities. The Environmental Protection Agency is sponsoring a quality assurance standard (ANSI/ASQC E-4) to replace EPA`s QAMS 005/80. The new standard is consistent with DOE Order 5700.6C, and it also stresses the integration of TQM principles within the quality assurance process. This paper discusses the intent and philosophy of the 10 criteria of the new DOE order, the status of ANSI/ASQC E-4, and how to effectively integrate TQM principles into the quality assurance process as the conversion is made from NQA-1 to DOE Order 5700.6C. The purpose and value of DOE Order 5700.6C Implementation Guide for research will also be discussed.

  4. 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

  5. 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)

  6. Safety requirements to be met in final storage of heat-producing waste an evaluation of the BMU draft

    International Nuclear Information System (INIS)

    Thomauske, B.

    2008-01-01

    The German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) on August 12, 2008 published a July 29, 2008 draft 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). The main changes made by the BMU are the introduction of a phased procedure in building repositories. A phased plans approval procedure under the Atomic Energy Act has been foreseen by the Ministry for this purpose. In addition, the draft provides for the introduction of a risk-based goal of protection. To ensure retrievability of the waste, the casks are to have a demonstrated service life of 500 years. The BMU draft safety requirements are unable to bring the safety criteria of 1983 up to the current state of the art. Here are the key points of criticism: - A risk-based goal of protection is introduced. The yardstick to be applied is to be defined in a guideline yet to be elaborated. As a consequence, the draft lacks substance. - As in licensing of nuclear facilities, the licensing procedure provides for a phased plans approval procedure for exploration. This analogy does not exist, as exploration is not the first phase of the plant to be built but a measure which is a precondition for obtaining a permit for construction and operation. - The information contained in the draft indicates that, contrary to international recommendations, it tightens the goal of protection by more than one order of magnitude. - The requirements to be met by the casks because of retrievability impose constraints on solutions optimized for safety in emplacement technology. - The risk-based approach is not mature and is

  7. 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

  8. 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

  9. 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.

  10. 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

  11. Meeting the requirements for a DOE environmental restoration project. The Fernald strategy

    International Nuclear Information System (INIS)

    Vanoss, R.L.; Risenhoover, G.M.

    1994-01-01

    Environmental Restoration (ER) of five Operable Units (OU) at Fernald Environmental Management Project (FEMP) includes compliance with the requirements of Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), Resource Conservation Recovery Act (RCRA), National Environmental Policy Act (NEPA), and DOE Orders. Each regulatory driver has differing procedural requirements for documenting calculations, decisions, and actions involved in site cleanup. Integration of documentation and avoidance of duplication can save time and money. Such savings are being achieved by OU specific application of supporting studies, revised procedures, and guidance documents. Each OU is seeking appropriate opportunities to produce single documents that simultaneously fulfill the important requirements of the other regulations and DOE orders. These opportunities are evaluated at all phases of decision making, remedial design, and remedial action. Three essential processes precede environmental restoration/remedial action at a DOE site/project: 1. Completion of decision-making documents required by governing or applicable statutes. 2. Completion of important scientific and engineering analyses of remedial alternatives, and design and implementation of the remedial solution established in the CERCLA Record of Decision (ROD). 3. Preparation of DOE-mandated documentation to record engineering evaluations and cost estimates required for budgeting, decision making, and project management. Methodology and requirements for each process have developed from long, successful practice, but independently of each other. FERMCO, as new DOE contractor at Fernald and first Environmental Restoration Management Contractor (ERMC), is committed to a process of Continuous Performance Improvement (CPI). A major reevaluation of documentation and processes for support of environmental decision-making and design of cleanup activities to remediate the five OUs at the FEMP is being undertaken

  12. Regulations for the Safe Transport of Radioactive Material. 2012 Edition. Specific Safety Requirements (Arabic Edition)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    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 conventions

  13. Regulations for the Safe Transport of Radioactive Material. 2012 Edition. Specific Safety Requirements (Chinese Edition)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-03-01

    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 conventions

  14. Regulations for the Safe Transport of Radioactive Material. 2012 Edition. Specific Safety Requirements

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-10-15

    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

  15. Regulations for the Safe Transport of Radioactive Material. 2012 Edition. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2012-01-01

    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

  16. Regulations for the Safe Transport of Radioactive Material. 2012 Edition. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2013-01-01

    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

  17. Regulations for the Safe Transport of Radioactive Material. 2012 Edition. Specific Safety Requirements (Chinese Edition)

    International Nuclear Information System (INIS)

    2013-01-01

    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 conventions

  18. Regulations for the Safe Transport of Radioactive Material. 2012 Edition. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    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 conventions

  19. 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)

  20. 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)

  1. 10 CFR 905.14 - Does Western require annual IRP progress reports?

    Science.gov (United States)

    2010-01-01

    ... Section 905.14 Energy DEPARTMENT OF ENERGY ENERGY PLANNING AND MANAGEMENT PROGRAM Integrated Resource Planning § 905.14 Does Western require annual IRP progress reports? Yes, customers must submit IRP progress... projected goals and implementation schedules, and energy and capacity benefits and renewable energy...

  2. 13 CFR 123.513 - Does SBA require collateral on its Military Reservist EIDL?

    Science.gov (United States)

    2010-01-01

    ..., or both. SBA will not decline a loan if you do not have a particular amount of collateral so long as SBA is reasonably sure that you can repay the loan. If you refuse to pledge the available collateral... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Does SBA require collateral on its...

  3. 32 CFR 555.9 - Reporting requirements for work in support of DOE.

    Science.gov (United States)

    2010-07-01

    ... 32 National Defense 3 2010-07-01 2010-07-01 true Reporting requirements for work in support of DOE. 555.9 Section 555.9 National Defense Department of Defense (Continued) DEPARTMENT OF THE ARMY MILITARY.... This notification shall include: (1) A brief statement of the problem. (2) Nature of corrective action...

  4. Spent fuel storage requirements. An update of DOE/RL-85-2

    International Nuclear Information System (INIS)

    1986-10-01

    Utility projections of spent fuel storage capacities indicate that some commercial light water reactors (LWRs) have inadequate capacity to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. The estimates in this report cover the period from the present through the year 2000. Although the DOE objective is to begin accepting spent fuel for final disposal in 1998, types of fuel and the receipt rates to be shipped are not yet known. Hence, this report makes no assumption regarding such fuel shipments. The report also assesses the possible impacts of increased fuel exposure and spent fuel transshipment on the requirements for additional storage capacity

  5. 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...

  6. 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

  7. 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

    waste. Upon approval of the permit modification, B696S rooms 1007, 1008, and 1009 will be able to store hazardous and mixed waste for up to 1 year. Furthermore, an additional drum crusher and 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 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., 90 Sr, 137 Cs, 3 H. Chapter 5 of the DSA documents the derivation of TSRs and develops the operational limits that protect the safety envelope defined for this facility. The DSA is applicable to the handling of radioactive waste stored and treated in the B695 Segment of the DWTF. Section 5 of the TSR, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the B695 Segment of the DWTF. A basis explanation follows each of the requirements described in Section 5.5, Specific Administrative Controls. The basis explanation does not constitute an additional requirement, but is intended as an expansion of the logic and reasoning behind development of the requirement. Programmatic Administrative Controls are addressed in Section 5.6

  8. 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

    mixed waste. Upon approval of the permit modification, B696S rooms 1007, 1008, and 1009 will be able to store hazardous and mixed waste for up to 1 year. Furthermore, an additional drum crusher and 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 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, {sup 3}H. Chapter 5 of the DSA documents the derivation of TSRs and develops the operational limits that protect the safety envelope defined for this facility. The DSA is applicable to the handling of radioactive waste stored and treated in the B695 Segment of the DWTF. Section 5 of the TSR, Administrative Controls, contains those Administrative Controls necessary to ensure safe operation of the B695 Segment of the DWTF. A basis explanation follows each of the requirements described in Section 5.5, Specific Administrative Controls. The basis explanation does not constitute an additional requirement, but is intended as an expansion of the logic and reasoning behind development of the requirement. Programmatic Administrative Controls are addressed in Section 5.6.

  9. Lessons learned in the implementation of Integrated Safety Management at DOE Order Compliance Sites vs Necessary and Sufficient Sites

    International Nuclear Information System (INIS)

    Hill, R.L.

    2000-01-01

    This paper summarizes the development and implementation of Integrated Safety Management (ISM) at an Order Compliance Site (Savannah River Site) and a Necessary and Sufficient Site (Nevada Test Site). A discussion of each core safety function of ISM is followed by an example from an Order Compliance Site and a Necessary and Sufficient Site. The Savannah River Site was the first DOE site to have a DOE Headquarters-validated and approved ISM System. The NTS is beginning the process of verification and validation. This paper defines successful strategies for integrating Environment, Safety, and Health management into work under various scenarios

  10. 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.)

  11. 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

  12. 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.

  13. 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...

  14. 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...

  15. 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...

  16. 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)

  17. 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

  18. “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.

  19. Safety First

    Science.gov (United States)

    Taft, Darryl

    2011-01-01

    Ned Miller does not take security lightly. As director of campus safety and emergency management at the Des Moines Area Community College (DMACC), any threat requires serious consideration. As community college administrators adopt a more proactive approach to campus safety, many institutions are experimenting with emerging technologies, including…

  20. 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.

  1. 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

  2. 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

  3. 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

  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. This

  5. 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

  6. 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

  7. 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

  8. 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

  9. 76 FR 42686 - DOE Response to Recommendation 2011-1 of the Defense Nuclear Facilities Safety Board, Safety...

    Science.gov (United States)

    2011-07-19

    ... examples of a failed safety culture.'' The Department disagrees with this categorization and believes the... Safety Board, Safety Culture at the Waste Treatment and Immobilization Plant AGENCY: Department of Energy... Recommendation 2011-1, concerning Safety Culture at the Waste Treatment and Immobilization Plant, to the...

  10. Spent fuel storage requirements. An update of DOE/RL-83-1

    International Nuclear Information System (INIS)

    1984-05-01

    Spent fuel storage capacities at some commercial light water reactors (LWRs) are inadequate to handle projected spent fuel discharges. This report presents estimates of potential near-term requirements for additional LWR spent fuel storage capacity, based on information voluntarily supplied by utilities operating commercial nuclear power plants. These estimates provide information needed for planning the Department of Energy's (DOE) Federal Interim Storage (FIS) Program and the spent fuel research, development, and demonstration (RD and D) activities to be carried out under the DOE's Commercial Spent Fuel Management (CSFM) Program, in conjunction with the requirements of the Nuclear Waste Policy Act of 1982. This report is the latest in a series published by the DOE on LWR spent fuel storage requirements. Since the planning needs of the CSFM program focus on the near-term management of spent fuel inventories from commercial nuclear power reactors, the estimates in this report cover the ten-year period from the present through 1983. The report also assesses the possible impacts of using various concepts to reduce the requirements for additional storage capacity

  11. 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.

  12. Does company size matter? Validation of an integrative model of safety behavior across small and large construction companies.

    Science.gov (United States)

    Guo, Brian H W; Yiu, Tak Wing; González, Vicente A

    2018-02-01

    Previous safety climate studies primarily focused on either large construction companies or the construction industry as a whole, while little is known about whether company size has significant effects on workers' understanding of safety climate measures and relationships between safety climate factors and safety behavior. Thus, this study aims to: (a) test the measurement equivalence (ME) of a safety climate measure across workers from small and large companies; (b) investigate if company size alters the causal structure of the integrative model developed by Guo, Yiu, and González (2016). Data were collected from 253 construction workers in New Zealand using a safety climate measure. This study used multi-group confirmatory factor analyses (MCFA) to test the measurement equivalence of the safety climate measure and structure invariance of the integrative model. Results indicate that workers from small and large companies understood the safety climate measure in a similar manner. In addition, it was suggested that company size does not change the causal structure and mediational processes of the integrative model. Both measurement equivalence of the safety climate measure and structural invariance of the integrative model were supported by this study. Practical applications: Findings of this study provided strong support for a meaningful use of the safety climate measure across construction companies in different sizes. Safety behavior promotion strategies designed based on the integrative model may be well suited for both large and small companies. Copyright © 2017 National Safety Council and Elsevier Ltd. All rights reserved.

  13. 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

  14. 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

  15. 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

  16. Successful intraosseous infusion in the critically ill patient does not require a medullary cavity.

    LENUS (Irish Health Repository)

    McCarthy, Gerard

    2012-02-03

    OBJECTIVES: To demonstrate that successful intraosseous infusion in critically ill patients does not require bone that contains a medullary cavity. DESIGN: Infusion of methyl green dye via standard intraosseous needles into bones without medullary cavity-in this case calcaneus and radial styloid-in cadaveric specimens. SETTING: University department of anatomy. PARTICIPANTS: Two adult cadaveric specimens. MAIN OUTCOME MEASURES: Observation of methyl green dye in peripheral veins of the limb in which the intraosseous infusion was performed. RESULTS: Methyl green dye was observed in peripheral veins of the chosen limb in five out of eight intraosseous infusions into bones without medullary cavity-calcaneus and radial styloid. CONCLUSIONS: Successful intraosseous infusion does not always require injection into a bone with a medullary cavity. Practitioners attempting intraosseous access on critically ill patients in the emergency department or prehospital setting need not restrict themselves to such bones. Calcaneus and radial styloid are both an acceptable alternative to traditional recommended sites.

  17. Status of cask procurement strategy to satisfy DOE/OCRWM requirements

    International Nuclear Information System (INIS)

    Teer, B.R.; Nolan, D.J.; Lake, W.H.

    1993-01-01

    The Nuclear Waste Policy Act requires the development of a safe and efficient system to transport spent nuclear fuel to and within the Federal Waste Management System. This paper describes the DOE/OCRWM strategy to develop and procure a major component of the Transportation System-the transport cask systems. The original initiative to develop high-capacity innovative designs and its current status is described. The follow-on phase to design and procure proven technology cask systems is also discussed

  18. 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)

  19. Proposed risk evaluation guidelines for use by the DOE-AL Nuclear Explosive Safety Division in evaluating proposed shipments of nuclear components

    International Nuclear Information System (INIS)

    Just, R.A.; Love, A.F.

    1997-10-01

    The licensing requirements of 10 CFR 71 (US Code of Federal Regulations) are the primary criteria used to license proposed US Department of Energy (DOE) shipments of nuclear components. However, if a shipment cannot meet 10 CFR 71 requirements, a Transportation System Risk Assessment (TSRA) is prepared to document: (1) the degree of compliance of proposed DOE shipments of nuclear components with applicable federal regulations, and (2) the risk associated with the proposed shipments. The Nuclear Explosive Safety Division (NESD) of the Department of Energy, Albuquerque Area Office (DOE-AL) is responsible for evaluating TSRAs and for preparing Safety Evaluation Reports (SERs) to authorize the off-site transport. Hazards associated with the transport may include the presence of fissile material, chemically and radiologically toxic uranium, and ionizing radiation. The Nuclear Regulatory Commission (NRC) has historically considered only radiological hazards in licensing the transport of radiological material because the US Department of Transportation considers licensing requirements of nonradiological (i.e., chemically toxic) hazards. The requirements of 10 CFR 71 are based primarily on consideration of radiological hazards. For completeness, this report provides information for assessing the effects of chemical toxicity. Evaluating the degree of compliance with the requirements of 10 CFR 71 is relatively straightforward. However, there are few precedents associated with developing TSRA risk assessments for packages that do not comply with all of the requirements of 10 CFR 71. The objective of the task is to develop Risk Evaluation Guidelines for DOE-AL to use when evaluating a TSRA. If the TSRA shows that the Risk Evaluation Guidelines are not exceeded, then from a risk perspective the TSRA should be approved if there is evidence that the ALARA (as low as reasonably achievable) principle has been applied

  20. 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)

  1. Public involvement in environmental, safety and health issues at the DOE Nuclear Weapons Complex

    International Nuclear Information System (INIS)

    Taylor, Laura L.; Morgan, Robert P.

    1992-01-01

    The state of public involvement in environmental, safety, and health issues at the DOE Nuclear Weapons Complex is assessed through identification of existing opportunities for public involvement and through interviews with representatives of ten local citizen groups active in these issues at weapons facilities in their communities. A framework for analyzing existing means of public involvement is developed. On the whole, opportunities for public involvement are inadequate. Provisions for public involvement are lacking in several key stages of the decision-making process. Consequently, adversarial means of public involvement have generally been more effective than cooperative means in motivating change in the Weapons Complex. Citizen advisory boards, both on the local and national level, may provide a means of improving public involvement in Weapons Complex issues. (author)

  2. Cross-index to DOE-prescribed occupational safety codes and standards

    International Nuclear Information System (INIS)

    1981-01-01

    This Cross-Index volume is the 1981 compilation of detailed information from more than three hundred and fifty DOE prescribed or OSHA referenced industrial safety codes and standards and is revised yearly to provide information from current codes. Condensed data from individual code portions are listed according to reference code, section, paragraph and page. Each code is given a two-digit reference code number or letter in the Contents section (pages C to L) of this volume. This reference code provides ready identification of any code listed in the Cross-Index. The computerized information listings are on the left-hand portion of Cross-Index page; in order to the right of the listing are the reference code letters or numbers, the section, paragraph and page of the referenced code containing expanded information on the individual listing

  3. 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...

  4. 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

  5. 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...

  6. 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.)

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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).

  12. Low level radiation: how does the linear without threshold model provide the safety of Canadian

    International Nuclear Information System (INIS)

    Anon.

    2010-01-01

    The linear without threshold model is a model of risk used worldwide by the most of health organisms of nuclear regulation in order to establish dose limits for workers and public. It is in the heart of the approach adopted by the Canadian commission of nuclear safety (C.C.S.N.) in matter of radiation protection. The linear without threshold model presumes reasonably it exists a direct link between radiation exposure and cancer rate. It does not exist scientific evidence that chronicle exposure to radiation doses under 100 milli sievert (mSv) leads harmful effects on health. Several scientific reports highlighted scientific evidences that seem indicate a low level of radiation is less harmful than the linear without threshold predicts. As the linear without threshold model presumes that any radiation exposure brings risks, the ALARA principle obliges the licensees to get the radiation exposure at the lowest reasonably achievable level, social and economical factors taken into account. ALARA principle constitutes a basic principle in the C.C.S.N. approach in matter of radiation protection; On the radiation protection plan, C.C.S.N. gets a careful approach that allows to provide health and safety of Canadian people and the protection of their environment. (N.C.)

  13. Development of a stand-alone microcomputer based DOE contractor generic radiation worker safety course

    International Nuclear Information System (INIS)

    Klos, D.B.; Gardner, P.R.

    1985-01-01

    Westinghouse Hanford Company (WHC) has developed Computer Based Training (CBT) materials for radiation and industrial safety. First released for general Fast Flux Test Facility in November, 1984. This course has now been taken by nearly 350 people. Completion times for new personnel average around eight hours. The next project undertaken was construction of a Radiation Worker Safety course generic enough for use by all contractors at the Hanford site. The design process of the Hanford site course indicated that the quantity of ''DOE common material'' may be sufficient to warrant consideration of a larger target population. Specifically, the course will be designed to run on an IBM-PC or compatible computer having 256K RAM, a standard IBM color graphics card or equivalent, a color graphics monitor, and two floppy disk drives or one hard disk. The target student population includes those who routinely work in Radiation Areas, especially crafts people. We are not targeting Health Physics personnel, except, possibly, for introductory training, nor are we directing the course toward ''casual'' or escorted workers

  14. Cross-Index to DOE-prescribed industrial safety codes and standards

    International Nuclear Information System (INIS)

    1980-01-01

    This Cross-Index volume is the 1980 compilation of detailed information from more than two hundred and ninety Department of Energy (DOE) prescribed or Occupational Health and Safety Administration (OSHA) referenced industrial safety codes and standards. The compilation of this material was conceived and initiated in 1973, and is revised yearly to provide information from current codes. Condensed data from individual code portions are listed according to reference code, section, paragraph, and page. Each code is given a two-digit reference code number or letter in the Contents section. This reference code provides ready identification of any code listed in the Cross-Index. The computerized information listings are on the left-hand portion of Cross-Index page; in order to the right of the listing are the reference code letters or numbers, the section, paragraph, and page of the referenced code containing expanded information on the individual listing. Simplified How to Use directions are listed. A glossary of letter initials/abbreviations for the organizations or documents, whose codes or standards are contained in this Cross-Index, is included

  15. Environment, safety, and health at DOE facilities. Annual report, calendar year 1981

    International Nuclear Information System (INIS)

    1982-10-01

    During 1981 the US Department of Energy's (DOE) safety and property protection performance remained excellent in all reported categories with loss rates generally less than one-third of comparable private-industry rates. The 1981 occupational-injury rates of 1.0 lost-workday cases and 13.5 lost workdays per 200,000 work hours were better than 1980 rates of 1.1 lost-workday cases and 17.5 lost workdays per 200,000 work hours. The recorded occupational illness rate, based on only 61 cases in 1981, was 0.04 cases per 200,000 work hours compared to 0.06 cases per 200,000 work hours for 1980. Property losses during 1981 totaled $4.7 million, with the largest loss of $1.25 million resulting from a fire at a Bonneville Power Administration lighting substation in Ellenberg, Washington. Fire losses during 1981 totaled $2.01 million, resulting in a fire loss rate of 0.38 cent per $100 property valuation - about one-eight the best class private-sector rate. Non-fire losses during 1981 totaled $2.65 million. The resulting non-fire loss rate of 0.48 cent per $100 property valuation was approx. one-third (36%) the 1980 loss rate of 1.34 cents per $100 property valuation. The 82,873 monitored Department of Energy federal and contractor employees received a total radiation dose of 6902 rem in 1981. The total amount of effluent radioactivity released from all DOE facilities in 1981 was 1.9 million curies, compared to 1.4 million curies in 1980. The largest probable whole-body radiation dose received by a maximally exposed public individual from any of these facilities was the same as in 1980: 15 millirem, or 3% of the DOE Radiation Protection Guide standard. A major accomplishemet during 1981 was the successful pilot testing of the Computerized Accident/Incident Reporting System (CAIRS), which will provide more-comprehensive safety information than the present data system

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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…

  4. 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)

  5. 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.

  6. The Development, Content, Design, and Conduct of the 2011 Piloted US DOE Nuclear Criticality Safety Program Criticality Safety Engineering Training and Education Project

    International Nuclear Information System (INIS)

    Hopper, Calvin Mitchell

    2011-01-01

    In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT and SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT and SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National

  7. 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

  8. DOE Integrated Safeguards and Security (DISS) historical document archival and retrieval analysis, requirements and recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Guyer, H.B.; McChesney, C.A.

    1994-10-07

    The overall primary Objective of HDAR is to create a repository of historical personnel security documents and provide the functionality needed for archival and retrieval use by other software modules and application users of the DISS/ET system. The software product to be produced from this specification is the Historical Document Archival and Retrieval Subsystem The product will provide the functionality to capture, retrieve and manage documents currently contained in the personnel security folders in DOE Operations Offices vaults at various locations across the United States. The long-term plan for DISS/ET includes the requirement to allow for capture and storage of arbitrary, currently undefined, clearance-related documents that fall outside the scope of the ``cradle-to-grave`` electronic processing provided by DISS/ET. However, this requirement is not within the scope of the requirements specified in this document.

  9. Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

    International Nuclear Information System (INIS)

    Widia Lastana Istanto

    2011-01-01

    Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

  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. Does lean management improve patient safety culture? An extensive evaluation of safety culture in a radiotherapy institute.

    Science.gov (United States)

    Simons, Pascale A M; Houben, Ruud; Vlayen, Annemie; Hellings, Johan; Pijls-Johannesma, Madelon; Marneffe, Wim; Vandijck, Dominique

    2015-02-01

    The importance of a safety culture to maximize safety is no longer questioned. However, achieving sustainable culture improvements are less evident. Evidence is growing for a multifaceted approach, where multiple safety interventions are combined. Lean management is such an integral approach to improve safety, quality and efficiency and therefore, could be expected to improve the safety culture. This paper presents the effects of lean management activities on the patient safety culture in a radiotherapy institute. Patient safety culture was evaluated over a three year period using triangulation of methodologies. Two surveys were distributed three times, workshops were performed twice, data from an incident reporting system (IRS) was monitored and results were explored using structured interviews with professionals. Averages, chi-square, logistical and multi-level regression were used for analysis. The workshops showed no changes in safety culture, whereas the surveys showed improvements on six out of twelve dimensions of safety climate. The intention to report incidents not reaching patient-level decreased in accordance with the decreasing number of reports in the IRS. However, the intention to take action in order to prevent future incidents improved (factorial survey presented β: 1.19 with p: 0.01). Due to increased problem solving and improvements in equipment, the number of incidents decreased. Although the intention to report incidents not reaching patient-level decreased, employees experienced sustained safety awareness and an increased intention to structurally improve. The patient safety culture improved due to the lean activities combined with an organizational restructure, and actual patient safety outcomes might have improved as well. Copyright © 2014 Elsevier Ltd. All rights reserved.

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. NERSC Cyber Security Challenges That Require DOE Development andSupport

    Energy Technology Data Exchange (ETDEWEB)

    Draney, Brent; Campbell, Scott; Walter, Howard

    2007-01-16

    Traditional security approaches do not adequately addressall the requirements of open, scientific computing facilities. Many ofthe methods used for more restricted environments, including almost allcorporate/commercial systems, do not meet the needs of today's science.Use of only the available "state of the practice" commercial methods willhave adverse impact on the ability of DOE to accomplish its sciencegoals, and impacts the productivity of the DOE Science community. Inparticular, NERSC and other high performance computing (HPC) centers havespecial security challenges that are unlikely to be met unless DOE fundsdevelopment and support of reliable and effective tools designed to meetthe cyber security needs of High Performance Science. The securitychallenges facing NERSC can be collected into three basic problem sets:network performance and dynamics, application complexity and diversity,and a complex user community that can have transient affiliations withactual institutions. To address these problems, NERSC proposes thefollowing four general solutions: auditing user and system activityacross sites; firewall port configuration in real time;cross-site/virtual organization identity management and access control;and detecting security issues in application middleware. Solutions arealsoproposed for three general long term issues: data volume,application complexity, and information integration.

  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. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. Assessment of the requirements for DOE's annual report to congress on low-level radioactive waste

    International Nuclear Information System (INIS)

    1987-10-01

    The Low-level Radioactive Waste Policy Amendments Act of 1985 (PL99-240; LLRWPAA) requires the Department of Energy (DOE) to ''submit to Congress on an annual basis a report which: (1) summarizes the progress of low-level waste disposal siting and licensing activities within each compact region, (2) reviews the available volume reduction technologies, their applications, effectiveness, and costs on a per unit volume basis, (3) reviews interim storage facility requirements, costs, and usage, (4) summarizes transportation requirements for such wastes on an inter- and intra-regional basis, (5) summarizes the data on the total amount of low-level waste shipped for disposal on a yearly basis, the proportion of such wastes subjected to volume reduction, the average volume reduction attained,, and the proportion of wastes stored on an interim basis, and (6) projects the interim storage and final disposal volume requirements anticipated for the following year, on a regional basis (Sec. 7(b)).'' This report reviews and assesses what is required for development of the annual report specified in the LLRWPAA. This report addresses each of the subject areas set out in the LLRWPAA

  7. Analysis of waste treatment requirements for DOE mixed wastes: Technical basis

    International Nuclear Information System (INIS)

    1995-02-01

    The risks and costs of managing DOE wastes are a direct function of the total quantities of 3wastes that are handled at each step of the management process. As part of the analysis of the management of DOE low-level mixed wastes (LLMW), a reference scheme has been developed for the treatment of these wastes to meet EPA criteria. The treatment analysis in a limited form was also applied to one option for treatment of transuranic wastes. The treatment requirements in all cases analyzed are based on a reference flowsheet which provides high level treatment trains for all LLMW. This report explains the background and basis for that treatment scheme. Reference waste stream chemical compositions and physical properties including densities were established for each stream in the data base. These compositions are used to define the expected behavior for wastes as they pass through the treatment train. Each EPA RCRA waste code was reviewed, the properties, chemical composition, or characteristics which are of importance to waste behavior in treatment were designated. Properties that dictate treatment requirements were then used to develop the treatment trains and identify the unit operations that would be included in these trains. A table was prepared showing a correlation of the waste physical matrix and the waste treatment requirements as a guide to the treatment analysis. The analysis of waste treatment loads is done by assigning wastes to treatment steps which would achieve RCRA compliant treatment. These correlation's allow one to examine the treatment requirements in a condensed manner and to see that all wastes and contaminant sets are fully considered

  8. IDENTIFICATION OF DOE'S POST-CLOSURE MONITORING NEEDS AND REQUIREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    M.A. Ebadian, Ph.D.

    1999-01-01

    The 2006 plan sets an ambitious agenda for the U.S. Department of Energy (DOE), Office of Environmental Management (EM) and the remediation of sites contaminated by decades of nuclear weapons production activities. The plan's primary objective is to reduce overall clean up costs by first eliminating the environmental problems that are most expensive to control and safely maintain. In the context of the 2006 Plan, closure refers to the completion of area or facility specific cleanup projects. The cleanup levels are determined by the planned future use of the site or facility. Use restrictions are still undecided for most sites but are highly probable to exclude residential or agricultural activities. Most of the land will be remediated to ''industrial use'' levels with access restrictions and some areas will be closed-off through containment. Portions of the site will be reserved for waste disposal, either as a waste repository or the in-situ immobilization of contaminated soil and groundwater, and land use will be restricted to waste disposal only. The land used for waste disposal will require monitoring and maintenance activities after closure. Most of the land used for industrial use may also require such postclosure activities. The required postclosure monitoring and maintenance activities will be imposed by regulators and stakeholders. Regulators will not approve closure plans without clearly defined monitoring methods using approved technologies. Therefore, among all other more costly and labor-intensive closure-related activities, inadequate planning for monitoring and lack of appropriate monitoring technologies can prevent closure. The purpose of this project is to determine, document, and track the current and evolving postclosure monitoring requirements at DOE-EM sites. This information will aid CMST-CP in guiding its postclosure technology development and deployment efforts.

  9. 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...

  10. 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.

  11. 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...

  12. 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.)

  13. 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...

  14. 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

  15. 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.

  16. Feed tank transfer requirements

    International Nuclear Information System (INIS)

    Freeman-Pollard, J.R.

    1998-01-01

    This document presents a definition of tank turnover; DOE responsibilities; TWRS DST permitting requirements; TWRS Authorization Basis (AB) requirements; TWRS AP Tank Farm operational requirements; unreviewed safety question (USQ) requirements; records and reporting requirements, and documentation which will require revision in support of transferring a DST in AP Tank Farm to a privatization contractor for use during Phase 1B

  17. 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.

  18. 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

  19. 25 CFR 39.219 - What happens if a residential program does not maintain residency levels required by this subpart?

    Science.gov (United States)

    2010-04-01

    ... 25 Indians 1 2010-04-01 2010-04-01 false What happens if a residential program does not maintain residency levels required by this subpart? 39.219 Section 39.219 Indians BUREAU OF INDIAN AFFAIRS..., Student Counts, and Verifications Residential Programs § 39.219 What happens if a residential program does...

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. Hazardous Substance Release Reporting Under CERCLA, EPCR {section}304 and DOE Emergency Management System (EMS) and DOE Occurrence Reporting Requirements. Environmental Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Traceski, T.T.

    1994-06-01

    Releases of various substances from DOE facilities may be subject to reporting requirements under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and the Emergency Planning and Community Right-to-Know Act (EPCRA), as well as DOE`s internal ``Occurrence Reporting and Processing of Operations Information`` and the ``Emergency Management System`` (EMS). CERCLA and EPCPA are Federal laws that require immediate reporting of a release of a Hazardous Substance (HS) and an Extremely Hazardous Substance (EHS), respectively, in a Reportable Quantity (RQ) or more within a 24-hour period. This guidance uses a flowchart, supplemental information, and tables to provide an overview of the process to be followed, and more detailed explanations of the actions that must be performed, when chemical releases of HSs, EHSs, pollutants, or contaminants occur at DOE facilities. This guidance should be used in conjunction with, rather than in lieu of, applicable laws, regulations, and DOE Orders. Relevant laws, regulations, and DOE Orders are referenced throughout this guidance.

  6. 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

  7. Guidance for the design and management of a maintenance plan to assure safety and improve the predictability of a DOE nuclear irradiation facility. Final report

    International Nuclear Information System (INIS)

    Booth, R.S.; Kryter, R.C.; Shepard, R.L.; Smith, O.L.; Upadhyaya, B.R.; Rowan, W.J.

    1994-10-01

    A program is recommended for planning the maintenance of DOE nuclear facilities that will help safety and enhance availability throughout a facility's life cycle. While investigating the requirements for maintenance activities, a major difference was identified between the strategy suitable for a conventional power reactor and one for a research reactor facility: the latter should provide a high degree of predicted availability (referred to hereafter as ''predictability'') to its users, whereas the former should maximize total energy production. These differing operating goals necessitate different maintenance strategies. A strategy for scheduling research reactor facility operation and shutdown for maintenance must balance safety, reliability,and predicted availability. The approach developed here is based on three major elements: (1) a probabilistic risk analysis of the balance between assured reliability and predictability (presented in Appendix C), (2) an assessment of the safety and operational impact of maintenance activities applied to various components of the facility, and (3) a data base of historical and operational information on the performance and requirements for maintenance of various components. These factors are integrated into a set of guidelines for designing a new highly maintainable facility, for preparing flexible schedules for improved maintenance of existing facilities, and for anticipating the maintenance required to extend the life of an aging facility. Although tailored to research reactor facilities, the methodology has broader applicability and may therefore be used to improved the maintenance of power reactors, particularly in anticipation of peak load demands

  8. An in fiber experimental approach to photonic quantum digital signatures that does not require quantum memory

    Science.gov (United States)

    Collins, Robert J.; Donaldon, Ross J.; Dunjko, Vedran; Wallden, Petros; Clarke, Patrick J.; Andersson, Erika; Jeffers, John; Buller, Gerald S.

    2014-10-01

    Classical digital signatures are commonly used in e-mail, electronic financial transactions and other forms of electronic communications to ensure that messages have not been tampered with in transit, and that messages are transferrable. The security of commonly used classical digital signature schemes relies on the computational difficulty of inverting certain mathematical functions. However, at present, there are no such one-way functions which have been proven to be hard to invert. With enough computational resources certain implementations of classical public key cryptosystems can be, and have been, broken with current technology. It is nevertheless possible to construct information-theoretically secure signature schemes, including quantum digital signature schemes. Quantum signature schemes can be made information theoretically secure based on the laws of quantum mechanics, while classical comparable protocols require additional resources such as secret communication and a trusted authority. Early demonstrations of quantum digital signatures required quantum memory, rendering them impractical at present. Our present implementation is based on a protocol that does not require quantum memory. It also uses the new technique of unambiguous quantum state elimination, Here we report experimental results for a test-bed system, recorded with a variety of different operating parameters, along with a discussion of aspects of the system security.

  9. 1994 annual report on waste generation and waste minimization progress as required by DOE Order 5400.1

    International Nuclear Information System (INIS)

    Irwin, E.F.; Poligone, S.E.

    1995-01-01

    The Y-12 Plant serves as a key manufacturing technology center for the development and demonstration of unique materials, components, and services of importance to the Department of Energy (DOE) and the nation. This is accomplished through the reclamation and storage of nuclear materials, manufacture of nuclear materials, manufacture of components for the nation's defense capabilities, support to national security programs, and services provided to other customers as approved by DOE. We are recognized by our people, the community, and our customers as innovative, responsive, and responsible. We are a leader in worker health and safety, environmental protection, and stewardship of our national resources. As a DOE facility, Y-12 also supports DOE's waste minimization mission. Data contained in this report represents waste generation in Tennessee

  10. 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...

  11. 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...

  12. How does a modal shift from short car trips to cycling affect road safety?

    NARCIS (Netherlands)

    Schepers, J. P.; Heinen, E.

    Governments aim to promote a shift from car to bicycle, but concerns about road safety seem to represent an important argument against this encouragement. This study examines the road safety impact of a modal shift from short car trips to cycling in Dutch municipalities. The road safety effect is

  13. 10 CFR 830.202 - Safety basis.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 4 2010-01-01 2010-01-01 false Safety basis. 830.202 Section 830.202 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.202 Safety basis. (a) The contractor responsible for a hazard category 1, 2, or 3 DOE nuclear facility must establish and maintain the safety basis...

  14. 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

  15. 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.

  16. 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)

  17. [Optimal intravascular brachytherapy: safety and radiation protection, reliability and precision guaranteed by guidelines, recommendations and regulatory requirements].

    Science.gov (United States)

    Quast, Ulrich; Kaulich, Theodor W; Lorenz, Joachim

    2002-02-01

    The success of intravascular brachytherapy relies entirely on the interdisciplinary approach. Interventional cardiologists, radiation oncologists and medical physicists must form a team from day 1. All members of the team need special knowledge and regular training in the field of vascular radiation therapy. Optimization of intravascular brachytherapy requires the use of standardized methods of dose specification, recording and reporting. This also implies using standardized methods of source calibration in terms of absorbed dose to water and having methods for simple internal control of the dosimetric quantities of new or replaced sources. Guidance is offered by international recommendations (AAPM TG 60, DGMP Report 16, NCS and EVA GEC-ESTRO). LEGAL REQUIREMENTS FOR RADIATION PROTECTION--WHAT'S NEW?: In Europe, new legal requirements on radiation protection issues have to be fulfilled. For Germany, the revised "Strahlenschutzverordnung" has been released recently. Nearly all organizational and medical processes are affected. For intravascular brachytherapy, several changes of requirements have to be considered. However, to follow these requirements does not cause serious problems. DGMP REPORT 16: GUIDELINES FOR MEDICAL PHYSICAL ASPECTS OF INTRAVASCULAR BRACHYTHERAPY: Evaluation of clinical results by comparison of intravascular brachytherapy treatment parameters is possible only if the prescribed dose and the applied dose distribution are reported clearly, completely and uniformly. The DGMP guidelines thus recommend to prescribe the dose to water at the system related reference point PRef at 2 mm radial distance for intracoronary application (and at 5 mm for peripheral vessels). The mean dose at 1 mm tissue depth (respectively at 2 mm) should be reported in addition. To safely define the planning target volume from the injured length, safety margins of at least 5 mm (10 mm) have to be taken into account on both ends. Safety margins have also to be considered for

  18. 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

  19. 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.

  20. 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.

  1. 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)

  2. 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

  3. Two Cultures in Modern Science and Technology: For Safety and Validity Does Medicine Have to Update?

    Science.gov (United States)

    Becker, Robert E

    2016-01-11

    Two different scientific cultures go unreconciled in modern medicine. Each culture accepts that scientific knowledge and technologies are vulnerable to and easily invalidated by methods and conditions of acquisition, interpretation, and application. How these vulnerabilities are addressed separates the 2 cultures and potentially explains medicine's difficulties eradicating errors. A traditional culture, dominant in medicine, leaves error control in the hands of individual and group investigators and practitioners. A competing modern scientific culture accepts errors as inevitable, pernicious, and pervasive sources of adverse events throughout medical research and patient care too malignant for individuals or groups to control. Error risks to the validity of scientific knowledge and safety in patient care require systemwide programming able to support a culture in medicine grounded in tested, continually updated, widely promulgated, and uniformly implemented standards of practice for research and patient care. Experiences from successes in other sciences and industries strongly support the need for leadership from the Institute of Medicine's recommended Center for Patient Safely within the Federal Executive branch of government.

  4. 24 CFR 1000.242 - When does the requirement for exemption from taxation apply to affordable housing activities?

    Science.gov (United States)

    2010-04-01

    ... exemption from taxation apply to affordable housing activities? 1000.242 Section 1000.242 Housing and Urban... ACTIVITIES Indian Housing Plan (IHP) § 1000.242 When does the requirement for exemption from taxation apply to affordable housing activities? The requirement for exemption from taxation applies only to rental...

  5. Postearthquake safety evaluation of buildings at DOE (Department of Energy) facilities

    International Nuclear Information System (INIS)

    Gallagher, R.

    1989-01-01

    New postearthquake building safety evaluation procedures have been developed. The procedures cover inspection and safety assessment of the principal types of building construction found in the US, including wood, masonry, tilt-up, concrete, and steel frame structures. Guidelines are also provided for appraising the structural safety significance of ground movements resulting from geologic hazards and for the inspection of nonstructural elements for falling and other hazards

  6. Does applying technology throughout the medication use process improve patient safety with antineoplastics?

    Science.gov (United States)

    Bubalo, Joseph; Warden, Bruce A; Wiegel, Joshua J; Nishida, Tess; Handel, Evelyn; Svoboda, Leanne M; Nguyen, Lam; Edillo, P Neil

    2014-12-01

    Medical errors, in particular medication errors, continue to be a troublesome factor in the delivery of safe and effective patient care. Antineoplastic agents represent a group of medications highly susceptible to medication errors due to their complex regimens and narrow therapeutic indices. As the majority of these medication errors are frequently associated with breakdowns in poorly defined systems, developing technologies and evolving workflows seem to be a logical approach to provide added safeguards against medication errors. This article will review both the pros and cons of today's technologies and their ability to simplify the medication use process, reduce medication errors, improve documentation, improve healthcare costs and increase provider efficiency as relates to the use of antineoplastic therapy throughout the medication use process. Several technologies, mainly computerized provider order entry (CPOE), barcode medication administration (BCMA), smart pumps, electronic medication administration record (eMAR), and telepharmacy, have been well described and proven to reduce medication errors, improve adherence to quality metrics, and/or improve healthcare costs in a broad scope of patients. The utilization of these technologies during antineoplastic therapy is weak at best and lacking for most. Specific to the antineoplastic medication use system, the only technology with data to adequately support a claim of reduced medication errors is CPOE. In addition to the benefits these technologies can provide, it is also important to recognize their potential to induce new types of errors and inefficiencies which can negatively impact patient care. The utilization of technology reduces but does not eliminate the potential for error. The evidence base to support technology in preventing medication errors is limited in general but even more deficient in the realm of antineoplastic therapy. Though CPOE has the best evidence to support its use in the

  7. Does employee safety influence customer satisfaction? Evidence from the electric utility industry.

    Science.gov (United States)

    Willis, P Geoffrey; Brown, Karen A; Prussia, Gregory E

    2012-12-01

    Research on workplace safety has not examined implications for business performance outcomes such as customer satisfaction. In a U.S. electric utility company, we surveyed 821 employees in 20 work groups, and also had access to archival safety data and the results of a customer satisfaction survey (n=341). In geographically-based work units where there were more employee injuries (based on archival records), customers were less satisfied with the service they received. Safety climate, mediated by safety citizenship behaviors (SCBs), added to the predictive power of the group-level model, but these two constructs exerted their influence independently from actual injuries. In combination, two safety-related predictor paths (injuries and climate/SCB) explained 53% of the variance in customer satisfaction. Results offer preliminary evidence that workplace safety influences customer satisfaction, suggesting that there are likely spillover effects between the safety environment and the service environment. Additional research will be needed to assess the specific mechanisms that convert employee injuries into palpable results for customers. Better safety climate and reductions in employee injuries have the potential to offer payoffs in terms of what customers experience. Copyright © 2012 National Safety Council and Elsevier Ltd. All rights reserved.

  8. 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 ...

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. Change in neighborhood traffic safety: does it matter in terms of physical activity?

    NARCIS (Netherlands)

    Jongeneel-Grimen, Birthe; Busschers, Wim; Droomers, Mariël; van Oers, Hans A. M.; Stronks, Karien; Kunst, Anton E.

    2013-01-01

    There is limited evidence on the causality of previously observed associations between neighborhood traffic safety and physical activity (PA). This study aims to contribute to this evidence by assessing the extent to which changes over time in neighborhood traffic safety were associated with PA.

  15. Does safety climate moderate the influence of staffing adequacy and work conditions on nurse injuries?

    Science.gov (United States)

    Mark, Barbara A; Hughes, Linda C; Belyea, Michael; Chang, Yunkyung; Hofmann, David; Jones, Cheryl B; Bacon, Cynthia T

    2007-01-01

    Hospital nurses have one of the highest work-related injury rates in the United States. Yet, approaches to improving employee safety have generally focused on attempts to modify individual behavior through enforced compliance with safety rules and mandatory participation in safety training. We examined a theoretical model that investigated the impact on nurse injuries (back injuries and needlesticks) of critical structural variables (staffing adequacy, work engagement, and work conditions) and further tested whether safety climate moderated these effects. A longitudinal, non-experimental, organizational study, conducted in 281 medical-surgical units in 143 general acute care hospitals in the United States. Work engagement and work conditions were positively related to safety climate, but not directly to nurse back injuries or needlesticks. Safety climate moderated the relationship between work engagement and needlesticks, while safety climate moderated the effect of work conditions on both needlesticks and back injuries, although in unexpected ways. DISCUSSION AND IMPACT ON INDUSTRY: Our findings suggest that positive work engagement and work conditions contribute to enhanced safety climate and can reduce nurse injuries.

  16. Does perceived neighborhood walkability and safety mediate the association between education and meeting physical activity guidelines?

    Science.gov (United States)

    Pratt, Michael; Yin, Shaoman; Soler, Robin; Njai, Rashid; Siegel, Paul Z; Liao, Youlian

    2015-04-09

    The role of neighborhood walkability and safety in mediating the association between education and physical activity has not been quantified. We used data from the 2010 and 2012 Communities Putting Prevention to Work Behavioral Risk Factor Surveillance System and structural equation modeling to estimate how much of the effect of education level on physical activity was mediated by perceived neighborhood walkability and safety. Neighborhood walkability accounts for 11.3% and neighborhood safety accounts for 6.8% of the effect. A modest proportion of the important association between education and physical activity is mediated by perceived neighborhood walkability and safety, suggesting that interventions focused on enhancing walkability and safety could reduce the disparity in physical activity associated with education level.

  17. Overview of the U.S. DOE Hydrogen Safety, Codes and Standards Program. Part 4: Hydrogen Sensors; Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Buttner, William J.; Rivkin, Carl; Burgess, Robert; Brosha, Eric; Mukundan, Rangachary; James, C. Will; Keller, Jay

    2016-12-01

    Hydrogen sensors are recognized as a critical element in the safety design for any hydrogen system. In this role, sensors can perform several important functions including indication of unintended hydrogen releases, activation of mitigation strategies to preclude the development of dangerous situations, activation of alarm systems and communication to first responders, and to initiate system shutdown. The functionality of hydrogen sensors in this capacity is decoupled from the system being monitored, thereby providing an independent safety component that is not affected by the system itself. The importance of hydrogen sensors has been recognized by DOE and by the Fuel Cell Technologies Office's Safety and Codes Standards (SCS) program in particular, which has for several years supported hydrogen safety sensor research and development. The SCS hydrogen sensor programs are currently led by the National Renewable Energy Laboratory, Los Alamos National Laboratory, and Lawrence Livermore National Laboratory. The current SCS sensor program encompasses the full range of issues related to safety sensors, including development of advance sensor platforms with exemplary performance, development of sensor-related code and standards, outreach to stakeholders on the role sensors play in facilitating deployment, technology evaluation, and support on the proper selection and use of sensors.

  18. EVALUATION OF STATE-OF-THE-ART MANIPULATORS AND REQUIREMENTS FOR DOE ROBOTICS APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    BLACK, DEREK; GRUPINSKI, STEPHEN

    1998-10-08

    This report provides an overview of applications within the DOE complex which could benefit from the use of modular robotics technology during remediation operations. Each application area contains one or more specific tasks which are presently conducted by humans under hazardous conditions or which are deemed highly impractical, or are altogether impossible without automation. Five major areas were investigated for specific needs with respect to automation. Information was collected on Mixed Waste Operations, Contaminant Automated Analysis, Tanks, Decontamination and Dismantlement and Automated Plutonium Processing. During this investigation, information was gathered from available literature, telephone interviews with informed personnel and on-site visits. This data serves to provide design requirements and guidelines for the design of a family of modular actuators, which will be used to construct manipulators suited to each task. In addition, a survey of existing modular manipulator designs is presented. This survey addresses modular manipulators developed inside government labs and in universities for such applications as space exploration or controls research. It also addresses efforts at commercially viable industrial manipulators which have been built. This survey of robotic systems provides the reader with a glimpse into what technology currently exists in the way of modular manipulator automation and, to a degree, where this technology may be applicable or, more often, where these systems are unsuited to EM applications. From the information gathered during this study, it is possible to sufficiently define the requirements of one manipulator system which can be used to conduct automated transfer operations within Plutonium gloveboxes. This manipulator will be constructed from ARM Automation actuator modules and will provide this application with a viable option for automation within these gloveboxes. The design issues surrounding this manipulator and its

  19. TYLCV-Is movement in planta does not require V2 protein

    Energy Technology Data Exchange (ETDEWEB)

    Hak, Hagit [Institute of Plant Sciences, Agricultural Research Organization, The Volcani Center, Bet Dagan (Israel); Department of Biological Chemistry, The Alexander Silberman Institute of Life Sciences, The Hebrew University of Jerusalem (Israel); Levy, Yael; Chandran, Sam A.; Belausov, Eduard [Institute of Plant Sciences, Agricultural Research Organization, The Volcani Center, Bet Dagan (Israel); Loyter, Abraham [Department of Biological Chemistry, The Alexander Silberman Institute of Life Sciences, The Hebrew University of Jerusalem (Israel); Lapidot, Moshe [Institute of Plant Sciences, Agricultural Research Organization, The Volcani Center, Bet Dagan (Israel); Gafni, Yedidya, E-mail: ygafni@volcani.agri.gov.il [Institute of Plant Sciences, Agricultural Research Organization, The Volcani Center, Bet Dagan (Israel)

    2015-03-15

    Tomato yellow leaf curl virus (TYLCV), a major tomato pathogen causing extensive crop losses, is a whitefly-transmitted geminivirus. V2 mutants of TYLCV-Is and related viruses tend to induce symptomless infection with attenuated viral DNA levels, while accumulating close to wild-type DNA levels in protoplasts, suggesting V2 as a movement protein. The discovery of plant-silencing mechanisms and viral silencing suppressors, V2 included, led us to reconsider V2's involvement in viral movement. We studied two mutant versions of the virus, one impaired in V2 silencing-suppression activity, and another carrying a non-translatable V2. While both mutant viruses spread in the infected plant to newly emerged leaves at the same rate as the wild-type virus, their DNA-accumulation levels were tenfold lower than in the wild-type virus. Thus, we suggest that the setback in virus proliferation, previously ascribed to a movement impediment, is due to lack of silencing-suppression activity. - Highlights: • TYLCV-Is V2 protein is localized in distinct microbodies throughout the cell cytoplasm, around the nucleus and in association with cytoplasmic strands but is not associated with the plasmodesmata. • Disruption of RNA-silencing suppression activity of TYLCV-Is V2 protein causes low titer of the virus in the infected plants. • The movement of TYLCV-Is in planta does not require a functional V2 protein.

  20. DOE standard: Radiological control

    Energy Technology Data Exchange (ETDEWEB)

    1999-07-01

    The Department of Energy (DOE) has developed this Standard to assist line managers in meeting their responsibilities for implementing occupational radiological control programs. DOE has established regulatory requirements for occupational radiation protection in Title 10 of the Code of Federal Regulations, Part 835 (10 CFR 835), ``Occupational Radiation Protection``. Failure to comply with these requirements may lead to appropriate enforcement actions as authorized under the Price Anderson Act Amendments (PAAA). While this Standard does not establish requirements, it does restate, paraphrase, or cite many (but not all) of the requirements of 10 CFR 835 and related documents (e.g., occupational safety and health, hazardous materials transportation, and environmental protection standards). Because of the wide range of activities undertaken by DOE and the varying requirements affecting these activities, DOE does not believe that it would be practical or useful to identify and reproduce the entire range of health and safety requirements in this Standard and therefore has not done so. In all cases, DOE cautions the user to review any underlying regulatory and contractual requirements and the primary guidance documents in their original context to ensure that the site program is adequate to ensure continuing compliance with the applicable requirements. To assist its operating entities in achieving and maintaining compliance with the requirements of 10 CFR 835, DOE has established its primary regulatory guidance in the DOE G 441.1 series of Guides. This Standard supplements the DOE G 441.1 series of Guides and serves as a secondary source of guidance for achieving compliance with 10 CFR 835.

  1. DOE standard: Radiological control

    International Nuclear Information System (INIS)

    1999-07-01

    The Department of Energy (DOE) has developed this Standard to assist line managers in meeting their responsibilities for implementing occupational radiological control programs. DOE has established regulatory requirements for occupational radiation protection in Title 10 of the Code of Federal Regulations, Part 835 (10 CFR 835), ''Occupational Radiation Protection''. Failure to comply with these requirements may lead to appropriate enforcement actions as authorized under the Price Anderson Act Amendments (PAAA). While this Standard does not establish requirements, it does restate, paraphrase, or cite many (but not all) of the requirements of 10 CFR 835 and related documents (e.g., occupational safety and health, hazardous materials transportation, and environmental protection standards). Because of the wide range of activities undertaken by DOE and the varying requirements affecting these activities, DOE does not believe that it would be practical or useful to identify and reproduce the entire range of health and safety requirements in this Standard and therefore has not done so. In all cases, DOE cautions the user to review any underlying regulatory and contractual requirements and the primary guidance documents in their original context to ensure that the site program is adequate to ensure continuing compliance with the applicable requirements. To assist its operating entities in achieving and maintaining compliance with the requirements of 10 CFR 835, DOE has established its primary regulatory guidance in the DOE G 441.1 series of Guides. This Standard supplements the DOE G 441.1 series of Guides and serves as a secondary source of guidance for achieving compliance with 10 CFR 835

  2. 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...

  3. 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)

  4. 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

  5. Account of requirements for modernization in VPBER-600 enhanced safety reactor instrumentation and control system development

    International Nuclear Information System (INIS)

    Shashkin, S.L.; Pobedonostsev, A.B.; Drumov, V.V.; Chudin, A.G.

    1993-01-01

    Nuclear power plant (NPP) with VPBER-600 reactor is a station of new generation. The specified term of reactor plant operation is 60 years and taking into account that the proposed term of starting the first power unit is on the turn of centuries one can definitely state that for Russia conditions VPBER-600 is a plant of 21 century. Such far removed term for NPP now in the stage of development as it can seem does not put the problems of modernization as first order tasks. But open-quotes...who does not think about future lives in the past.close quotes It is that the NPP instrumentation and control (I ampersand C) systems are in the most degree subjected to the influence of factors which favor their modifications. These factors can be arbitrarily divided into two groups: (1) inner factors, i.e. changes (failures, aging, etc) in I ampersand C components as well as changes dictated by technological reasons (change of equipment composition, control algorithms, operation modes); (2) outer factors, i.e. intensive development of information technologies and rapid improvement of electronic components. This presentation addresses the problem of modernization of the safety instrumentation for this next generation facility, and the research effort it will entail. The system is designed to allow for modernization, and the relatively easy adoption of new instrumentation and technology as it becomes available

  6. Does the accreditation of private dental practices work? Time to rethink how accreditation can improve patient safety.

    Science.gov (United States)

    Jean, Gillian

    2017-10-09

    Accreditation to demonstrate engagement with the National Safety and Quality Health Service Standards (Standards) is compulsory for most hospital and healthcare settings, but to date remains voluntary for private dental practices (PDPs). The regulatory framework governing the dental profession lacks a proactive element to drive improvements in quality and safety of care, and an accreditation scheme can strengthen existing regulation. The current model of accreditation operating in accordance with the Australian Health Service Safety and Quality Accreditation Scheme (Scheme) is based on the Standards, which were written for a hospital model of healthcare service. The majority of PDPs are small office-based businesses with clear leadership structure and employing six staff or fewer. The Scheme is overly bureaucratic given the simplicity of the PDP business model. This article considers whether accreditation has a proven track record of improving quality of service and offers opinions about how a more appropriate safety management program for PDPs may look. What is known about the topic? There has been minimal research about the impact of accreditation schemes in improving patient safety in PDP. What does this paper add? This paper proposes a redesign of the Scheme to make it more relevant to PDPs. The paper offers strategies to minimise duplication of purpose between accreditation and existing legislation; and to strengthen critical elements of accreditation to improve effects on patient safety. What are the implications for practitioners? A redesigned accreditation scheme will support dental practitioners to implement a quality assurance system with improved efficiency, reduced administrative burden, and optimised patient safety.

  7. 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

  8. 76 FR 61350 - DOE Response to Defense Nuclear Facilities Safety Board's Request for Clarification on...

    Science.gov (United States)

    2011-10-04

    ...., that the independence, stature, and leadership experience of the implementation team that will be... technical competence, objectivity, experience in safety management, executive leadership, and a clear... the power of perceptions fully into account. 4. The independence, public stature, and leadership...

  9. Worker Safety and Health Issues Associated with the DOE Environmental Cleanup Program: Insights From the DOE Laboratory Directors' Environmental and Occupational/Public health Standards Steering Group

    International Nuclear Information System (INIS)

    M.C. Edelson; Samuel C. Morris; Joan M. Daisey

    2001-01-01

    The U.S. Department of Energy (DOE) Laboratory Directors' Environmental and Occupational/Public Health Standards Steering Group (or ''SSG'') was formed in 1990. It was felt then that ''risk'' could be an organizing principle for environmental cleanup and that risk-based cleanup standards could rationalize clean up work. The environmental remediation process puts workers engaged in cleanup activities at risk from hazardous materials and from the more usual hazards associated with construction activities. In a real sense, the site remediation process involves the transfer of a hypothetical risk to the environment and the public from isolated contamination into real risks to the workers engaged in the remediation activities. Late in its existence the SSG, primarily motivated by its LANL representative, Dr. Harry Ettinger, actively investigated issues associated with worker health and safety during environmental remediation activities. This paper summarizes the insights noted by the SSG. Most continue to be pertinent today

  10. Does Perceived Neighborhood Walkability and Safety Mediate the Association Between Education and Meeting Physical Activity Guidelines?

    OpenAIRE

    Pratt, Michael; Yin, Shaoman; Soler, Robin; Njai, Rashid; Siegel, Paul Z.; Liao, Youlian

    2015-01-01

    The role of neighborhood walkability and safety in mediating the association between education and physical activity has not been quantified. We used data from the 2010 and 2012 Communities Putting Prevention to Work Behavioral Risk Factor Surveillance System and structural equation modeling to estimate how much of the effect of education level on physical activity was mediated by perceived neighborhood walkability and safety. Neighborhood walkability accounts for 11.3% and neighborhood safet...

  11. DOES FOOD SAFETY CONFLICT WITH FOOD SECURITY? THE SAFE CONSUMPTION OF FOOD

    OpenAIRE

    Kinsey, Jean D.

    2004-01-01

    This paper concludes by saying no, food safety and security reinforce each other. It combines food safety and food security into the concept of "safe food consumption." Unsafe food consumption occurs when food contains known substances that lead to short or long term illness or death (botulism) and suspect substances that are believed to lead to delayed diseases (pesticides). It also occurs when hunger or over eating contribute to long-term illness and shorter life expectancy. The costs of il...

  12. 76 FR 13397 - DOE Response to Recommendation 2010-2 of the Defense Nuclear Facilities Safety Board, Pulse Jet...

    Science.gov (United States)

    2011-03-11

    ... Treatment and Immobilization Plant (WTP) will achieve their design and operating requirements. DOE has... meeting; in the October 7-8, 2010 public hearing on WTP; and in our supplement to the public hearing... implementation of these commitments is on-going as part of WTP project plans that supports scheduled testing to...

  13. 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)

  14. 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.)

  15. Safety class methodology

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  16. 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

  17. 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

  18. 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)

  19. Lessons learned from on-site safety assessments performed by DOE in response to the Tomsk accident

    International Nuclear Information System (INIS)

    Witmer, F.E.

    1995-01-01

    In response to the accident, in April 1993, at the nuclear fuel reprocessing plant of the Siberian chemical Combine, Tomsk, Russia, the U.S. Department of Energy (DOE) initiated concurrent efforts to understand the causes for the accident and to review potential vulnerabilities for similar occurrences across the DOE radiochemical complex. Because the accident occurred in the feed adjustment stage of a Purex type process, US facilities which contained significant inventories of TBP, organic diluent and nitric acid were evaluated by expert teams. From accident conditions, prior experience, modeling and experimental programs and confirmatory dialogue with the Russians, enhanced understanding was achieved and vulnerabilities (e.g., lack of safety analysis, organic layering, inadvertent acid addition, use of aromatic diluents, uncertain venting capability, no mitigative/emergency procedures, etc.) were identified and corrected

  20. Merger of two dispatch centres: does it improve quality and patient safety?

    Science.gov (United States)

    Moser, Alexandre; Mettler, Annette; Fuchs, Vincent; Hanhart, Walter; Robert, Claude-François; Della Santa, Vincent; Dami, Fabrice

    2017-04-13

    Dispatch centres (DCs) are considered an essential but expensive component of many highly developed healthcare systems. The number of DCs in a country, region, or state is usually based on local history and often related to highly decentralised healthcare systems. Today, current technology (Global Positioning System or Internet access) abolishes the need for closeness between DCs and the population. Switzerland went from 22 DCs in 2006 to 17 today. This study describes from a quality and patient safety point of view the merger of two DCs. The study analysed the performance (over and under-triage) of two medical DCs for 12 months prior to merging and for 12 months again after the merger in 2015. Performance was measured comparing the priority level chosen by dispatcher and the severity of cases assessed by paramedics on site using the National Advisory Committee for Aeronautics (NACA) score. We ruled that NACA score > 3 (injuries/diseases which can possibly lead to deterioration of vital signs) to 7 (lethal injuries/diseases) should require a priority dispatch with lights and siren (L&S). While NACA score  3. Prior to merging, Dispatch A had a sensitivity/specificity regarding the use of lights and sirens and severity of cases of 86%/48% with over- and under-triage rates of 78% and 5%, respectively. Dispatch B had sensitivity and specificity of 92%/20% and over- and under-triage rates of 84% and 7%, respectively. After they merged, global sensitivity/specificity reached 87%/67%, and over- and under-triage rates were 71% and 3%, respectively CONCLUSIONS: A part the potential cost advantage achieved by the merger of two DCs, it can improve the quality of services to the population, reducing over- and under-triage and the use of lights and sirens and therefore, the risk of accidents. This is especially the case when a DC with poor triage performance merges with a high-performing DC.

  1. 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.)

  2. 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.

  3. 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)

  4. DOE enforcement program roles and responsibilities: DOE handbook

    International Nuclear Information System (INIS)

    1995-08-01

    The Price-Anderson Act provides indemnification to DOE contractors who manage and conduct nuclear activities in the DOE complex. The government acts as an insurer for these contractors against any findings of liability from the nuclear activities of the contractor within the scope of its contract. 10 CFR Part 820 establishes the legal framework for implementing DOE's Nuclear Safety Enforcement Program. Integration with other DOE organizations and programs would assure that the enforcement process properly considers the actual or potential safety significance of a violation when determining an appropriate enforcement sanction. Achieving a proactive contractor compliance assurance rather than a heavy enforcement hand, will require a foundation of cooperation and teamwork across DOE organizations. This handbook identifies the areas of interface for the DOE Enforcement Program and provides guidance on roles and responsibilities for the key DOE organizational areas. It complements DOE-HDBK-1087-95 and 1089-95

  5. 14 CFR 382.97 - To which aircraft does the requirement to provide boarding and deplaning assistance through the...

    Science.gov (United States)

    2010-01-01

    ... provide boarding and deplaning assistance through the use of lifts apply? 382.97 Section 382.97... REGULATIONS NONDISCRIMINATION ON THE BASIS OF DISABILITY IN AIR TRAVEL Boarding, Deplaning, and Connecting Assistance § 382.97 To which aircraft does the requirement to provide boarding and deplaning assistance...

  6. 43 CFR 404.60 - Does this rule contain an information collection that requires approval by the Office of...

    Science.gov (United States)

    2010-10-01

    ... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Does this rule contain an information collection that requires approval by the Office of Management and Budget (OMB)? 404.60 Section 404.60 Public Lands: Interior Regulations Relating to Public Lands BUREAU OF RECLAMATION, DEPARTMENT OF THE INTERIOR RECLAMATION RURAL WATER SUPPLY PROGRAM...

  7. 49 CFR 40.208 - What problem requires corrective action but does not result in the cancellation of a test?

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 1 2010-10-01 2010-10-01 false What problem requires corrective action but does not result in the cancellation of a test? 40.208 Section 40.208 Transportation Office of the Secretary of Transportation PROCEDURES FOR TRANSPORTATION WORKPLACE DRUG AND ALCOHOL TESTING PROGRAMS Problems...

  8. 41 CFR 102-118.25 - Does GSA still require my agency to submit its overall transportation policies for approval?

    Science.gov (United States)

    2010-07-01

    ... Public Contracts and Property Management Federal Property Management Regulations System (Continued) FEDERAL MANAGEMENT REGULATION TRANSPORTATION 118-TRANSPORTATION PAYMENT AND AUDIT General Introduction... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Does GSA still require...

  9. DOE spent nuclear fuel -- Nuclear criticality safety challenges and safeguards initiatives

    International Nuclear Information System (INIS)

    Hopper, C.M.

    1994-01-01

    The field of nuclear criticality safety is confronted with growing technical challenges and the need for forward-thinking initiatives to address and resolve issues surrounding economic, safe and secure packaging, transport, interim storage, and long-term disposal of spent nuclear fuel. These challenges are reflected in multiparameter problems involving optimization of packaging designs for maximizing the density of material per package while ensuring subcriticality and safety under variable normal and hypothetical transport and storage conditions and for minimizing costs. Historic and recently revealed uncertainties in basic data used for performing nuclear subcriticality evaluations and safety analyses highlight the need to be vigilant in assessing the validity and range of applicability of calculational evaluations that represent extrapolations from ''benchmark'' data. Examples of these uncertainties are provided. Additionally, uncertainties resulting from the safeguarding of various forms of fissionable materials in transit and storage are discussed

  10. Does a reactor need a safety backfit. Case study on communicating decision and risk analysis information to managers

    Energy Technology Data Exchange (ETDEWEB)

    Brown, R.V.; Ulvila, J.W.

    1988-06-01

    An approach to communicating decision and risk analysis findings to managers is illustrated in a real case context. This article consists essentially of a report prepared for senior managers of the Nuclear Regulatory Commission to help them make a reactor safety decision. It illustrates the communication of decision analysis findings relating to technical risks, costs, and benefits in support of a major risk management decision: whether or not to require a safety backfit. Its focus is on the needs of decision makers, and it introduces some novel communication devices.

  11. 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...

  12. Better Oversight Needed for Safety and Health Activities at DOE’s Nuclear Facilities.

    Science.gov (United States)

    1981-08-04

    States Acca !3ion Tor Kl. ; BRA&I DTIC TAi \\ • inounced Jusl Lfii I "• ’ DiStril on/ r\\ D i Avai3 i - ’•ity Coi a j...m*d——I—^1 I —* Even DOE’s audit report tracking system failed to adequately document corrective actions taken or planned. This system was...designed by DOE to ensure that corrective actions on audit report recommendations are responsive, timely and complete. Audit re- ports, such as ours

  13. Safety assessment and surveillance of decommissioning operations at DOE's nuclear facilities

    International Nuclear Information System (INIS)

    Cowgill, M.G.; Prochnow, D.; Worthington, P.R.

    1995-01-01

    A description is provided of a systematic approach currently being developed and deployed at the Department of Energy to obtain assurance that post-operational activities at nuclear facilities will be conducted in a safe manner. Using this approach, personnel will have available a formalized set of safety principles and associated question sets to assist them in the conducting of safety assessments and surveillance. Information gathered through this means will also be analyzed to determine if there are any generic complex-wide strengths or deficiencies associated with decommissioning activities and to which attention should be drawn

  14. 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.

  15. 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.

  16. Multimegawatt Space Reactor Safety

    International Nuclear Information System (INIS)

    Stanley, M.L.

    1989-01-01

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

  17. Detrimental effects of subsequent increases in the safety and quality requirements using the Grohnde nuclear power station as an example

    International Nuclear Information System (INIS)

    Boettcher, D.

    1983-01-01

    The excellent operational availability and freedom from faults of German nuclear powerstations should give one the courage to take further sensible steps. From the operator's view these include: - Refusal to accept backfitting to a different state of science and technology. Instead of this, orderly introduction of new solutions after careful testing, unless meeting an emergency requires immediate action. - Further support of efforts at standardization of the industry with the possibility of transferring experience. - Reducing multiple inspections (the previous occurrence of multiple inspections in manufacture and erection in a system hides the danger of routine and creeping delegation of responsibility and attention among those concerned). - Limiting the extent of structural and repeat tests to the essential minimum, particularly where there are hold-ups caused during manufacture and erection, which prevent optimum economic construction. - Dispensing with complete documentation of every activity by the applicant, manufacturer, authority and expert. This may contribute to providing proofs for legal processes, but does not contribute to obtaining greater safety. (orig./RW) [de

  18. Comparative analysis of DOE Order 5820.2A, NRC, and EPA radioactive and mixed waste management requirements

    International Nuclear Information System (INIS)

    1991-07-01

    As directed by DOE-Headquarters and DOE-Idaho, the Radioactive Waste Technical Support Program (TSP) drafted an analysis of DOE Order 5820. 2A on ''Radioactive Waste Management'' to develop guidelines and criteria for revising the Order. This comparative matrix is a follow up to the earlier analysis. This matrix comparing the requirements of DOE Order 5820.2A with Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA) regulations was prepared at the request of EM-30. The matrix compares DOE Order 5820.2A with the following: NRC regulations in 10 CFR Part 61 on ''Licensing Requirements for Land Disposal of Radioactive Waste''; NRC regulations in 10 CFR Part 60 on ''Disposal of High-Level Radioactive Waste in Geologic Repositories''; EPA regulations in 40 CFR Part 191 on ''Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Waste''; EPA regulations in 40 FR Part 192 on ''Health and Environmental Protection Standards for Uranium and Thorium Mill Tailing''; and EPA regulations under the Resource Conservation and Recovery Act

  19. Commercial Vehicle Safety Alliance (CVSA)/Department of Energy (DOE) cooperative agreement final report

    International Nuclear Information System (INIS)

    Slavich, Antoinette; Daust, James E.

    1999-01-01

    This S and T product is a culmination of the activities, including research of the Commercial Vehicle Safety Alliance (CVSA) in developing and implementing inspection procedures and the out-of-service criteria for states and tribes to use when inspecting HRCQ and Transuranic shipments of radioactive materials. The report also contains the results of a pilot study to test the procedures

  20. Does worker participation improve health and safety? Findings from the Netherlands

    NARCIS (Netherlands)

    Popma, J.R.

    2009-01-01

    Worker participation, both at a sectoral and organisational level, has been assigned a prominent role in Dutch occupational safety and health (OSH) policy since the 1980s. This emphasis on worker participation stems from the belief that such participation may enhance the effectiveness of OSH policy