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Sample records for safety design requires

  1. 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.

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

    International Nuclear Information System (INIS)

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    2004-01-01

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

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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.

  9. 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.

  10. 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)

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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...

  3. 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)

  4. 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.

  5. 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.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. Design requirements for the new reactor

    International Nuclear Information System (INIS)

    Koski, S.

    2005-01-01

    This presentation deals with the safety related design requirements specified for the new nuclear power plant to be built in Finland (FINS). The legislation, codes and standards, on which the design requirements are based, can be arranged into a hierarchical pyramid as follows: The safety related design criteria are based on the three uppermost hierarchical levels: Finnish legislation (e.g. decisions of the State Council) Basic Regulations (75-INSAG-3, USNRC General Design Criteria) Process oriented nuclear documents (YVL- guides or corresponding US/German rules). The European Utility Requirements (EUR) document was used as the starting point for the writing of the design requirements document. The structure and headlines of EUR could be kept, but in many cases the contents had to be deleted and rewritten to correspond to the requirement level of the above codes and standards. This was the case, for example, with the requirements concerning safety classification or application of failure criteria. In the presentation, the most important safety related design criteria are reviewed, with an emphasis on those requirements which exceed the requirement level applied on the existing plant units. Some hints are also given on the main differences between Finnish and international safety requirements. (orig.)

  15. Tool-based requirement traceability between requirement and design artifacts

    CERN Document Server

    Turban, Bernhard

    2013-01-01

    Processes for developing safety-critical systems impose special demands on ensuring requirements traceability. Achieving valuable traceability information, however, is especially difficult concerning the transition from requirements to design. Bernhard Turban analyzes systems and software engineering theories cross-cutting the issue (embedded systems development, systems engineering, software engineering, requirements engineering and management, design theory and processes for safety-critical systems). As a solution, the author proposes a new tool approach to support designers in their thinkin

  16. 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

  17. 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

  18. 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)

  19. Preliminary safety design analysis of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Kwon, Y. M.; Kim, K. D. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This document first introduces a set of safety design requirements and accident evaluation criteria established for the conceptual design of KALIMER and then summarizes some of the preliminary results of engineering and design analyses performed for the safety of KALIMER. 19 refs., 19 figs., 6 tabs. (Author)

  20. Engineering design guidelines for nuclear criticality safety

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1988-08-01

    This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

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

    International Nuclear Information System (INIS)

    2009-01-01

    ) Facilities where the mining and processing of radioactive ores (such as ores of uranium and thorium) are carried out. 'Activities' includes: (a) production, use, import and export of radiation sources for industrial, research, medical and other purposes; (b) transport of radioactive material; (c) decommissioning and dismantling of facilities and the closure of repositories for radioactive waste; (d) close-out of facilities where the mining and processing of radioactive ore was carried out; (e) activities for radioactive waste management such as the discharge of effluents; (f) remediation of sites affected by residues from past activities. Safety assessment plays an important role throughout the lifetime of the facility or activity whenever decisions on safety issues are made by the designers, the constructors, the manufacturers, the operating organization or the regulatory body. Stages in the lifetime of a facility or activity where a safety assessment is carried out, updated and used by the designers, the operating organization and the regulatory body include: (a) site evaluation for the facility or activity; (b) development of the design; (c) construction of the facility or implementation of the activity; (d) commissioning of the facility or activity; (e) commencement of operation of the facility or conduct of the activity; (f) normal operation of the facility or normal conduct of the activity; (g) modification of the design or operation; (h) periodic safety reviews;(i) life extension of the facility beyond its original design life; (j) changes in ownership or management of the facility; (k) decommissioning and dismantling of a facility; (l) closure of a repository for the disposal of radioactive waste and the post-closure phase; (m) remediation of a site and release from regulatory control. The publication is structured as follows: An introduction is followed by Section 2 which provides the basis for requiring a safety assessment to be carried out, derived from the

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

    International Nuclear Information System (INIS)

    2009-01-01

    ) Facilities where the mining and processing of radioactive ores (such as ores of uranium and thorium) are carried out. 'Activities' includes: (a) production, use, import and export of radiation sources for industrial, research, medical and other purposes; (b) transport of radioactive material; (c) decommissioning and dismantling of facilities and the closure of repositories for radioactive waste; (d) close-out of facilities where the mining and processing of radioactive ore was carried out; (e) activities for radioactive waste management such as the discharge of effluents; (f) remediation of sites affected by residues from past activities. Safety assessment plays an important role throughout the lifetime of the facility or activity whenever decisions on safety issues are made by the designers, the constructors, the manufacturers, the operating organization or the regulatory body. Stages in the lifetime of a facility or activity where a safety assessment is carried out, updated and used by the designers, the operating organization and the regulatory body include: (a) site evaluation for the facility or activity; (b) development of the design; (c) construction of the facility or implementation of the activity; (d) commissioning of the facility or activity; (e) commencement of operation of the facility or conduct of the activity; (f) normal operation of the facility or normal conduct of the activity; (g) modification of the design or operation; (h) periodic safety reviews; (i) life extension of the facility beyond its original design life; (j) changes in ownership or management of the facility; (k) decommissioning and dismantling of a facility; (l) closure of a repository for the disposal of radioactive waste and the post-closure phase; (m) remediation of a site and release from regulatory control. The publication is structured as follows: An introduction is followed by Section 2 which provides the basis for requiring a safety assessment to be carried out, derived from the

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

    International Nuclear Information System (INIS)

    2010-01-01

    ) Facilities where the mining and processing of radioactive ores (such as ores of uranium and thorium) are carried out. 'Activities' includes: (a) production, use, import and export of radiation sources for industrial, research, medical and other purposes; (b) transport of radioactive material; (c) decommissioning and dismantling of facilities and the closure of repositories for radioactive waste; (d) close-out of facilities where the mining and processing of radioactive ore was carried out; (e) activities for radioactive waste management such as the discharge of effluents; (f) remediation of sites affected by residues from past activities. Safety assessment plays an important role throughout the lifetime of the facility or activity whenever decisions on safety issues are made by the designers, the constructors, the manufacturers, the operating organization or the regulatory body. Stages in the lifetime of a facility or activity where a safety assessment is carried out, updated and used by the designers, the operating organization and the regulatory body include: (a) site evaluation for the facility or activity; (b) development of the design; (c) construction of the facility or implementation of the activity; (d) commissioning of the facility or activity; (e) commencement of operation of the facility or conduct of the activity; (f) normal operation of the facility or normal conduct of the activity; (g) modification of the design or operation; (h) periodic safety reviews; (i) life extension of the facility beyond its original design life; (j) changes in ownership or management of the facility; (k) decommissioning and dismantling of a facility; (l) closure of a repository for the disposal of radioactive waste and the post-closure phase; (m) remediation of a site and release from regulatory control. The publication is structured as follows: An introduction is followed by Section 2 which provides the basis for requiring a safety assessment to be carried out, derived from the

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

    International Nuclear Information System (INIS)

    2009-01-01

    installed; (i) Facilities where the mining and processing of radioactive ores (such as ores of uranium and thorium) are carried out. 'Activities' includes: (a) production, use, import and export of radiation sources for industrial, research, medical and other purposes; (b) transport of radioactive material; (c) decommissioning and dismantling of facilities and the closure of repositories for radioactive waste; (d) close-out of facilities where the mining and processing of radioactive ore was carried out; (e) activities for radioactive waste management such as the discharge of effluents; (f) remediation of sites affected by residues from past activities. Safety assessment plays an important role throughout the lifetime of the facility or activity whenever decisions on safety issues are made by the designers, the constructors, the manufacturers, the operating organization or the regulatory body. Stages in the lifetime of a facility or activity where a safety assessment is carried out, updated and used by the designers, the operating organization and the regulatory body include: (a) site evaluation for the facility or activity; (b) development of the design; (c) construction of the facility or implementation of the activity; (d) commissioning of the facility or activity; (e) commencement of operation of the facility or conduct of the activity; (f) normal operation of the facility or normal conduct of the activity; (g) modification of the design or operation; (h) periodic safety reviews;(i) life extension of the facility beyond its original design life; (j) changes in ownership or management of the facility; (k) decommissioning and dismantling of a facility; (l) closure of a repository for the disposal of radioactive waste and the post-closure phase; (m) remediation of a site and release from regulatory control. The publication is structured as follows: An introduction is followed by Section 2 which provides the basis for requiring a safety assessment to be carried out, derived

  5. Safety design guides for fire protection 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 establishes design requirements to ensure the radiological risk to the public due to fire is acceptable and operating personnel are adequately protected from the hazards of fires. This safety design guide also specifies the safety criteria for fire protection to be applied to mitigate fires and recommends the fire protection program to be established to initiate, coordinate and document the design activities associated with fire protection. The requirements for fire protection outlined in this safety design guide shall be satisfied in the design stage and the change status of the regulatory requirements, code and standards should be traced and incorporated into this safety design guide accordingly. 1 fig., (Author) .new

  6. 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.)

  7. NUCLEAR SAFETY DESIGN BASES FOR LICENSE APPLICATION

    International Nuclear Information System (INIS)

    Garrett, R.J.

    2005-01-01

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111 [DIRS 156605] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113 [DIRS 156605] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

  8. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    Energy Technology Data Exchange (ETDEWEB)

    RYAN GW

    2007-09-24

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents.

  9. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    International Nuclear Information System (INIS)

    RYAN GW

    2007-01-01

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents

  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. EC6 safety design improvements

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.; Lee, A.G.; Soulard, M. [Candu Energy Inc., Mississauga, ON (Canada)

    2014-07-01

    The Enhanced CANDU 6 (EC6) builds on the proven high performance design such as the Qinshan CANDU 6 reactor, and has made improvements to safety, operational performance, and has incorporated extensive operational feedback. Completion of all three phases of the pre-licensing design review by the Canadian Regulator - the Canadian Nuclear Safety Commission has provided a higher level of assurance that the EC6 reference design has taken modern regulatory requirements and expectations into account and further confirmed that there are no fundamental barriers to licensing the EC6 design in Canada. The EC6 design is based on the defence-in-depth principles in INSAG-10 and provides further safety features that address the lessons learned from Fukushima. With these safety features, the EC6 design has strengthened accident prevention as the first priority in the defence-in-depth strategy, as outlined in INSAG-10. As well, the EC6 design has incorporated further mitigation measures to provide additional protection of the public and the environment if the preventive measures fail. The EC6 design has an appropriate combination of inherent, passive safety characteristics, engineered features and administrative safety measures to effectively prevent and mitigate severe accident progressions. A strong contributor to the robustness and redundancy of CANDU design is the two-group separation philosophy. This ensures a high degree of independence between safety systems as well as physical separation and functional independence in how fundamental safety functions are provided. This paper will describe the following safety features based on the application of defence-in-depth and design approach to prevent beyond design basis events progressing to severe accidents and to mitigate the consequences if it occurs: Improved steam generator heat sink via a more reliable emergency heat removal system; Increased time before manual field actions are required via enhanced capacity of

  12. Nuclear Safety Design Base for License Application

    International Nuclear Information System (INIS)

    R.J. Garrett

    2005-01-01

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111(b) [DIRS 173273] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113(b) and (c) [DIRS 173273] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

  13. 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

  14. 49 CFR 229.206 - Design requirements.

    Science.gov (United States)

    2010-10-01

    ...-climber, emergency egress, emergency interior lighting, and interior configuration design requirements set... 49 Transportation 4 2010-10-01 2010-10-01 false Design requirements. 229.206 Section 229.206..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Locomotive Crashworthiness Design...

  15. 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)

  16. Development of ABWR-2 and its safety design

    International Nuclear Information System (INIS)

    Takafumi, Anegawa; Kenji, Tateiwa

    2002-01-01

    This paper reports the current status of development project on ABWR-II, a next generation reactor design based on ABWR, and its safety design. This project was initiated over a decade ago and has completed three phases to date. In Phase I (1991-92), basic design requirements were discussed and several plant concepts were studied. In Phase II (1993-95), key design features were selected in order to establish a reference reactor concept. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements were made to fulfill the design requirements. By adopting large electric output (1 700 MW), large fuel bundle, modified ECCS, and passive heat removal systems, among other design features, we achieved a design concept capable of increasing both economic competitiveness and safety performance. Main focus of this paper will be on the safety design, safety performance, and further research needs related to safety. (authors)

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

    International Nuclear Information System (INIS)

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

    1996-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  19. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  1. PHWR safety: design, siting and construction

    International Nuclear Information System (INIS)

    Sharma, V.K.

    2002-01-01

    In all activities associated with NPPs viz. siting, design, construction, commissioning and operation, safety is given overriding importance. The safety design principles of PHWRs are based on defence-in-depth approach, physical and functional separation between process and safety systems and also among various safety systems, redundancy to meet single failure criteria and postulation of a number of design basis events for which the plant must be designed. Apart from engineered safety systems, PHWRs have inherent characteristics which contribute to safety. In siting of a NPP, it is required to ensure that the given site does not pose undue radiological hazard to public and the environment both during normal operation as well as during and following an accident condition. For this purpose, all site related external events, both natural and man induced, are assessed for their effect on the plant and are considered as part of the design basis. Possible radiological impact of the NPP on environment and surrounding population is assessed and ensured to be within acceptable limits. During construction phase, it is essential that the NPP be built in accordance with design intent and with required quality of workmanship to ensure that the NPP will remain safe during all states of operation. This is achieved through careful execution and QA activities encompassing all aspects of component fabrication at manufacturer works, civil construction, site erection, assembly, and commissioning. Future trends in nuclear safety will continue to be based on existing principles which have proved to be sound. These will be further strengthened by features such as increasing use of passive means of performing safety functions and a more explicit treatment of severe accidents. (author)

  2. Comparative Study on Research Reactor Design Requirements between IAEA and Korea

    International Nuclear Information System (INIS)

    Chang, Won Joon; Yune, Young Gill; Song, Myung Ho; Cho, Seung Ho

    2013-01-01

    This study has identified the gaps in the safety requirements for design of research reactors of Korea comparing with those of the IAEA. The review results showed that the gaps have arisen mainly from the following aspects: - The differences in the characteristics of design and operation between power reactor and research reactor - Enhancement of the level of safety of nuclear reactor facility - Consideration of advanced safety technologies. The review results will be utilized to reflect the IAEA safety requirements for design of research reactors into those of Korea, which will contribute to enhancing the level of safety and improving the technical standards of research reactors of Korea. The IAEA safety standards encompass international consensus to strengthen the nuclear safety and to reflect the latest advancement of nuclear safety technologies. Also, they provide reliable means to ensure the effective fulfillment of obligations under the various international safety conventions. Many countries have adopted the IAEA safety standards as their national standards in nuclear regulations. Since Korea has exported research reactor technologies abroad these days and will continue to export them in the future, it is desirable to harmonize domestic safety requirements for research reactor with those of the IAEA. The KINS (Korea Institute of Nuclear Safety) has performed a review of the IAEA safety requirements for design of research reactors comparing with those of Korea. The purpose of this comparative study is to harmonize the safety requirements for the design of research reactors of Korea with those of the IAEA as a member state of the IAEA, and to encompass global efforts to enhance the nuclear safety and to reflect the latest advancement of nuclear safety technologies into the safety requirements for the design of research reactors of Korea. Design requirements for structures, systems, and components of research reactors important to safety, which are required to

  3. 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

  4. 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.

  5. Cold vacuum drying facility design requirements

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    1999-07-01

    This document provides the detailed design requirements for the Spent Nuclear Fuel Project Cold Vacuum Drying Facility. Process, safety, and quality assurance requirements and interfaces are specified.

  6. Cold vacuum drying facility design requirements

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1999-01-01

    This document provides the detailed design requirements for the Spent Nuclear Fuel Project Cold Vacuum Drying Facility. Process, safety, and quality assurance requirements and interfaces are specified

  7. 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

  8. Safety design guides for grouping and separation 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 grouping and separation describes the philosophy of physical and functional separation for systems, structures and components in CANDU 9 plants and provides the requirements for the implementation of the philosophy in the detailed plant design. The separation of the safety systems is to ensure that common cause events and functional interconnections between systems do not impair the capability to perform the required safety functions for accident conditions. The separation requirements are also applied to the design by grouping the plant systems into two basic groups. Group 1 includes the power production systems and Group 2 includes the safety related systems required for the mitigation of serious process failure. The Group 2 is further separated into subgroups to ensure that events that could cause failure of a special safety system in one subgroup can be mitigated by the other subgroup. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 6 figs. (Author) .new

  9. Safety design guides for grouping and separation for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

    This safety design guide for grouping and separation describes the philosophy of physical and functional separation for systems, structures and components in CANDU 9 plants and provides the requirements for the implementation of the philosophy in the detailed plant design. The separation of the safety systems is to ensure that common cause events and functional interconnections between systems do not impair the capability to perform the required safety functions for accident conditions. The separation requirements are also applied to the design by grouping the plant systems into two basic groups. Group 1 includes the power production systems and Group 2 includes the safety related systems required for the mitigation of serious process failure. The Group 2 is further separated into subgroups to ensure that events that could cause failure of a special safety system in one subgroup can be mitigated by the other subgroup. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 6 figs. (Author) .new.

  10. General design safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide provides the safety principles and the approach that have been used to implement the Code in the Safety Guides. These safety principles and the approach are tied closely to the safety analyses needed to assist the design process, and are used to verify the adequacy of nuclear power plant designs. This Guide also provides a framework for the use of other design Safety Guides. However, although it explains the principles on which the other Safety Guides are based, the requirements for specific applications of these principles are mostly found in the other Guides

  11. System Design and the Safety Basis

    International Nuclear Information System (INIS)

    Ellingson, Darrel

    2008-01-01

    The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities

  12. Safety design

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Shiozawa, Shusaku

    2004-01-01

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

  13. Multinational Design Evaluation Programme (MDEP) - Safety Goals

    International Nuclear Information System (INIS)

    Vaughan, G.J.

    2011-01-01

    One of the aims of the NEA's Multinational Design Evaluation Programme (MDEP) is to work towards greater harmonisation of regulatory requirements. To achieve this aim, it is necessary that there is a degree of convergence on the safety goals that are required to be met by designers and operators. The term 'safety goals' is defined to cover all health and safety requirements which must be met: these may be deterministic rules and/or probabilistic targets. They should cover the safety of workers, public and the environment in line with the IAEA's Basic Safety Objective; encompassing safety in normal operation through to severe accidents. MDEP is also interested in how its work can be extended to future reactors, which may use significantly different technology to the almost ubiquitous LWRs used today and in the next generation, building on the close co-operation within MDEP between the regulators who are currently engaged in constructing or carrying out design reviews on new designs. For two designs this work has involved several regulators sharing their safety assessments and in some cases issuing statements on issues that need to be addressed. Work is also progressing towards joint regulatory position statements on specific assessment areas. Harmonisation of safety goals will enhance the cooperation between regulators as further developments in design and technology occur. All regulators have safety goals, but these are expressed in many different ways and exercises in comparing them frequently are done at a very low level eg specific temperatures in the reactor vessel of a specific reactor type. The differences in the requirements from different regulators are difficult to resolve as the goals are derived using different principles and assumptions and are often for a specific technology. Therefore a different approach is being investigated, starting with the top-level safety goals and try to derive a structure and means of deriving lower tier

  14. Code on the safety of nuclear research reactors: Design

    International Nuclear Information System (INIS)

    1992-01-01

    The main objective of this publication is to provide a safety basis for the design of a research reactor and for the assessment of the design. Another objective is to cover certain aspects related to regulatory supervision, siting and quality assurance, as far as these are related to activities for the design of a research reactor. These objectives are expressed in terms of requirements and recommendations for the design of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop specific regulations and safety criteria for its research reactor programme.

  15. 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)

  16. The role of probabilistic safety assessment in the design

    International Nuclear Information System (INIS)

    Green, A.; Ingham, E.L.

    1989-01-01

    The use of probabilistic safety assessment (PSA) for Heysham 2 and Torness marked a major change in the design approach to nuclear safety within the U.K. Design Safety Guidelines incorporating probabilistic safety targets required that design justification would necessitate explicit consideration of the consequence of accidents in relation to their frequency. The paper discusses these safety targets and their implications, the integration of PSA into the design process and an outline of the methodology. The influence of PSA on the design is discussed together with its role in the overall demonstration of reactor safety. (author)

  17. DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

    The purpose of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) Design Package 1E, Surface Facilities, (for a list of design items included in the package 1E system safety analysis see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the Design Package 1E structures/systems/components(S/S/Cs) in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions.

  18. 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)

  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. Analysis of regulatory requirement for beyond design basis events of SMART

    International Nuclear Information System (INIS)

    Kim, W. S.; Seol, K. W.

    2000-01-01

    To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review

  1. 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?

  2. 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

  3. 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.

  4. Safety design guides for environmental qualification 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 describes the safety philosophy and requirements for the environmental qualification of safety related systems and components for CANDU 9. The environmental qualification program identifies the equipments to be qualified and conditions to be used for qualification and provides comprehensive set of documentation to ensure that the qualification is complete and can be maintained for the life of the plant. A summary of the system, components and structures requiring environmental qualification is provided in the table for the guidance of the system design, and this table will be subject to change or confirmation by the environmental qualification program. Also, plant ares subject to harsh environment is provided in the figure. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 tab., 5 figs. (Author) .new

  5. 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)

  6. 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

  7. Design requirements for new nuclear reactor facilities in Canada

    International Nuclear Information System (INIS)

    Shim, S.; Ohn, M.; Harwood, C.

    2012-01-01

    The Canadian Nuclear Safety Commission (CNSC) has been establishing the regulatory framework for the efficient and effective licensing of new nuclear reactor facilities. This regulatory framework includes the documentation of the requirements for the design and safety analysis of new nuclear reactor facilities, regardless of size. For this purpose, the CNSC has published the design and safety analysis requirements in the following two sets of regulatory documents: 1. RD-337, Design of New Nuclear Power Plants and RD-310, Safety Analysis for Nuclear Power Plants; and 2. RD-367, Design of Small Reactor Facilities and RD-308, Deterministic Safety Analysis for Small Reactor Facilities. These regulatory documents have been modernized to document past practices and experience and to be consistent with national and international standards. These regulatory documents provide the requirements for the design and safety analysis at a high level presented in a hierarchical structure. These documents were developed in a technology neutral approach so that they can be applicable for a wide variety of water cooled reactor facilities. This paper highlights two particular aspects of these regulatory documents: The use of a graded approach to make the documents applicable for a wide variety of nuclear reactor facilities including nuclear power plants (NPPs) and small reactor facilities; and, Design requirements that are new and different from past Canadian practices. Finally, this paper presents some of the proposed changes in RD-337 to implement specific details of the recommendations of the CNSC Fukushima Task Force Report. Major changes were not needed as the 2008 version of RD-337 already contained requirements to address most of the lessons learned from the Fukushima event of March 2011. (author)

  8. 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

  9. 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

  10. 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)

  11. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  12. Proposal for a technology-neutral safety approach for new reactor designs

    International Nuclear Information System (INIS)

    2007-09-01

    Many states are considering an expansion of their nuclear power generation programmes. Many of the technologies and concepts are new and innovative. The current design and licensing rules are applicable to mostly large water reactors and there are no accepted rules in place for design, safety assessment and licensing for new innovative nuclear power plants. This TECDOC proposes a (new) safety approach and a methodology to generate technology-neutral (i.e. independent of reactor technology) safety requirements and a 'safe design' for advanced and innovative reactors. The experience gained in decades of design and licensing, combined with the development of risk-based concepts, has provided insights that will form the basis for new safety rules and requirements. Many lessons learned acknowledge the importance of such concepts as safety goals and defence in depth and the benefits of integrating risk insights early in an iterative design process. A new safety approach will incorporate many of the new developments in these concepts. For example, the probabilistic elements of defence in depth will help define the cumulative provisions to compensate for uncertainty and incompleteness of our knowledge of accident initiation and progression. This TECDOC also identifies areas of work, which will require further definition, research and development and guidance on application. This publication is to be used as a guide to developing a new technology-neutral safety approach, and as a guide in the application of methodologies to define the safety requirements for an innovative reactor designs. The method proposes an integration of deterministic and probabilistic considerations with established principles and concepts such as safety goals and defence in depth. The TECDOC recommends that the structure of the new technology-neutral main pillars for the design and licensing of innovative nuclear reactors be developed following a top-down approach to reflect a newer risk-informed and

  13. ALS beamline design requirements: A guide for beamline designers

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-06-01

    This manual is written as a guide for researchers in designing beamlines and endstations acceptable for use at the ALS. It contains guidelines and policies related to personnel safety and equipment and vacuum protection. All equipment and procedures must ultimately satisfy the safety requirements set aside in the Lawrence Berkeley National Laboratory (LBNL) Health and Safety Manual (PUB-3000) which is available from the ALS User Office or on the World WideWeb from the LBNL Homepage (http:// www.lbl.gov).

  14. 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

  15. Safety design analyses of Korea Advanced Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Park, C.K.

    2000-01-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

  16. Safety aspects and shield design of a Poton irradiator

    International Nuclear Information System (INIS)

    Mehta, S.K.; Nayak, A.R.; Bongirwar, D.R.; Modi, R.K.; Ramkumar, M.S.

    1998-01-01

    An irradiation plant, POTON, for irradiation of potatoes and onions is being set up at Nashik. Shield design and safety features of this plant incorporate some novel and innovative features like a compact cell, curved cell boundaries for smooth conveyor movement though the cell labyrinth and conform to ICRP and AERB design safety requirements. The safety features include multiple safety interlocks, audio-visual alarms, scram switches and trip wire for avoiding accidental exposures. (author)

  17. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    Keen, Barry A.

    1990-01-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  18. Advanced gas cooled reactors - Designing for safety

    Energy Technology Data Exchange (ETDEWEB)

    Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

    1990-07-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

  19. 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.

  20. Design requirement on HYPER blanket fuel assembly

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

    2000-07-01

    This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER

  1. 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

  2. Panel 1: Safety design criteria

    International Nuclear Information System (INIS)

    Yllera, Javier

    2013-01-01

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

  3. Safety design guides for containment extension 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 containment extension describes the containment isolation philosophy and containment extension requirements. The metal extensions and components falling within the scope of ASME Section III are classified in accordance with the CAN/CSA-N285.0 and CAN/CSA-N285.3. The special consideration for the leak monitoring capability, seismic qualification and inspection requirements for containment extensions, etc., are defined in this design guide. In addition, the containment isolation systems are defined and summarized schematically in appendix A. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. (Author) .new

  4. 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.)

  5. Design requirement on KALIMER control rod assembly duct

    International Nuclear Information System (INIS)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J.

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

  6. Design requirement on KALIMER control rod assembly duct

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

  7. Conceptual safety design analysis of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Suk, S. D.; Park, C. K.

    1999-01-01

    The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

  8. 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)

  9. Requirements for Space Settlement Design

    Science.gov (United States)

    Gale, Anita E.; Edwards, Richard P.

    2004-02-01

    When large space settlements are finally built, inevitably the customers who pay for them will start the process by specifying requirements with a Request for Proposal (RFP). Although we are decades away from seeing the first of these documents, some of their contents can be anticipated now, and provide insight into the variety of elements that must be researched and developed before space settlements can happen. Space Settlement Design Competitions for High School students present design challenges in the form of RFPs, which predict basic requirements for space settlement attributes in the future, including structural features, infrastructure, living conveniences, computers, business areas, and safety. These requirements are generically summarized, and unique requirements are noted for specific space settlement locations and applications.

  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. 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

  12. 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

  13. Safety criteria for design of nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

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

  14. European passive plant program preliminary safety analyses to support system design

    International Nuclear Information System (INIS)

    Saiu, Gianfranco; Barucca, Luciana; King, K.J.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

  15. Analysis of effect of safety classification on DCS design in nuclear power plants

    International Nuclear Information System (INIS)

    Gou Guokai; Li Guomin; Wang Qunfeng

    2011-01-01

    By analyzing the safety classification for the systems and functions of nuclear power plants based on the general design requirements for nuclear power plants, especially the requirement of availability and reliability of I and C systems, the characteristics of modem DCS technology and I and C products currently applied in nuclear power field are interpreted. According to the requirements on the safety operation of nuclear power plants and the regulations for safety audit, the effect of different safety classifications on DCS design in nuclear power plants is analyzed, by considering the actual design process of different DCS solutions in the nuclear power plants under construction. (authors)

  16. Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

    International Nuclear Information System (INIS)

    Holcomb, David Eugene

    2015-01-01

    Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate

  17. 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

  18. 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

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

    International Nuclear Information System (INIS)

    Jones, G.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

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

  1. 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

  2. Design of concrete structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety standard for civil engineering structures important to safety of nuclear facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design of concrete structures important to safety

  3. 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.

  4. 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.

  5. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design

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

    International Nuclear Information System (INIS)

    1984-12-01

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

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

    International Nuclear Information System (INIS)

    Moon, D.W.

    1988-10-01

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

  8. Conceptual design of safety instrumentation for PFBR

    International Nuclear Information System (INIS)

    Muralikrishna, G.; Seshadri, U.; Raghavan, K.

    1996-01-01

    Instrumentation systems enable monitoring of the process which in turn enables control and shutdown of the process as per the requirements. Safety Instrumentation due to its vital importance has a stringent role and this needs to be designed methodically. This paper presents the details of the conceptual design for PFBR. (author). 4 figs, 3 tabs

  9. Design aspects of safety critical instrumentation of nuclear installations

    Energy Technology Data Exchange (ETDEWEB)

    Swaminathan, P. [Electronics Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)]. E-mail: swamy@igcar.ernet.in

    2005-07-01

    Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

  10. Design aspects of safety critical instrumentation of nuclear installations

    International Nuclear Information System (INIS)

    Swaminathan, P.

    2005-01-01

    Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

  11. 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

  12. Design of agricultural product quality safety retrospective supervision system of Jiangsu province

    Science.gov (United States)

    Wang, Kun

    2017-08-01

    In store and supermarkets to consumers can trace back agricultural products through the electronic province card to query their origin, planting, processing, packaging, testing and other important information and found that the problems. Quality and safety issues can identify the responsibility of the problem. This paper designs a retroactive supervision system for the quality and safety of agricultural products in Jiangsu Province. Based on the analysis of agricultural production and business process, the goal of Jiangsu agricultural product quality safety traceability system construction is established, and the specific functional requirements and non-functioning requirements of the retroactive system are analyzed, and the target is specified for the specific construction of the retroactive system. The design of the quality and safety traceability system in Jiangsu province contains the design of the overall design, the trace code design and the system function module.

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

    International Nuclear Information System (INIS)

    Reitsma, Frederik; Silady, Fred; Kunitomi, Kazuhiko

    2014-01-01

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

  14. 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.

  15. 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

  16. 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

  17. ARIES-AT safety design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D.A. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States)]. E-mail: David.Petti@inl.gov; Merrill, B.J. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Moore, R.L. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Longhurst, G.R. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); El-Guebaly, L. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Mogahed, E. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Henderson, D. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Wilson, P. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Abdou, A. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2006-01-15

    ARIES-AT is a 1000 MWe conceptual fusion power plant design with a very low projected cost of electricity. The design contains many innovative features to improve both the physics and engineering performance of the system. From the safety and environmental perspective, there is greater depth to the overall analysis than in past ARIES studies. For ARIES-AT, the overall spectrum of off-normal events to be examined has been broadened. They include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant, and in-vessel off-normal events that mobilize in-vessel inventories (e.g., tritium and tokamak dust) and bypass primary confinement such as a loss of vacuum and an in-vessel loss of coolant with bypass. This broader examination of accidents improves the robustness of the design from the safety perspective and gives additional confidence that the facility can meet the no-evacuation requirement under average weather conditions. We also provide a systematic assessment of the design to address key safety functions such as confinement, decay heat removal, and chemical energy control. In the area of waste management, both the volume of the component and its hazard are used to classify the waste. In comparison to previous ARIES designs, the overall waste volume is less because of the compact design.

  18. Engineered barrier systems (EBS): design requirements and constraints

    International Nuclear Information System (INIS)

    2004-01-01

    A joint NEA-EC workshop entitled 'Engineered Barrier Systems: Design Requirements and Constraints' was organised in Turku, Finland on 26-29 August 2003 and hosted by Posiva Oy. The main objectives of the workshop were to promote interaction and collaboration among experts responsible for engineering design and safety assessment in order to develop a greater understanding of how to achieve the integration needed for the successful design of engineered barrier systems, and to clarify the role that an EBS can play in the overall safety case for a repository. These proceedings present the outcomes of this workshop. (author)

  19. Safety regulation for the design approval of special form radioactive sources

    International Nuclear Information System (INIS)

    Cho, Woon-Kap

    2009-01-01

    Several kinds of special form radioactive sources for industrial, medical applications are being produced in Korea. Special form radioactive sources should meet strict safety requirements specified in the domestic safety regulations and the design of the sources should be certified by the regulatory authority, the Ministry of Education, Science and Technology (MEST). Several safety tests such as impact, percussion, heating, and leak tests are performed on the sources according to the domestic regulations and the international safety standards such as ANSI N542-1977 and ISO 2919-1999(E). As a regulatory expert body, Korea Institute of Nuclear Safety (KINS) assesses various types of application documents, such as safety analysis report, quality assurance program, and other documents evidencing fulfillment of requirements for design approval of the special form radioactive sources, submitted by a legal person who intends to produce special form radioactive sources and then reports the assessment result to MEST. A design approval certificate is issued to the applicant by MEST on the basis of a technical evaluation report presented by KINS.

  20. Safety and security aspects in design of digital safety I and C in nuclear power plants

    International Nuclear Information System (INIS)

    Ding, Yongjian; Waedt, Karl

    2016-01-01

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  1. Safety and security aspects in design of digital safety I and C in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Yongjian [University of Applied Sciences Magdeburg-Stendal, Magdeburg (Germany). Inst. of Electrical Engineering; Waedt, Karl [Areva GmbH, Erlangen (Germany). PEAS-G

    2016-05-15

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  2. Directives and general design requirements for a small PWR

    International Nuclear Information System (INIS)

    Arrieta, L.A.

    1992-08-01

    A proposal of directives and general requirements for the development of a small PWR conceptual design is presented. These directives address the main safety, performance and economic design aspects. The purpose is to use this work as a base for a wide discussion, involving all project participants, culminating with the definition of the final directives and general requirements. (author)

  3. Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

    International Nuclear Information System (INIS)

    Duckwitz, Noel

    2011-01-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets,' safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, 'Facility Safety,' and the expectations of DOE-STD-1189-2008, 'Integration of Safety into the Design Process,' provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  4. 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

  5. Range Flight Safety Requirements

    Science.gov (United States)

    Loftin, Charles E.; Hudson, Sandra M.

    2018-01-01

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

  6. 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.

  7. Evaluation of the Ventilation and Air Cleaning System Design Concepts for Safety Requirements during Fire Conditions in Nuclear Applications

    International Nuclear Information System (INIS)

    Rashad, S.; El-Fawal, M.; Kandil, M.

    2013-01-01

    The ventilation and air cleaning system in the nuclear or radiological installations is one of the essential nuclear safety concerns. It is responsible for confining the radioactive materials involved behind suitable barriers during normal and abnormal conditions. It must be designed to prevent the release of harmful products (radioactive gases, or airborne radioactive materials) from the system or facility, impacting the public or workers, and doing environmental damage. There are two important safety functions common to all ventilation and air cleaning system in nuclear facilities. They are: a) the requirements to maintain the pressure of the ventilated volume below that of surrounding, relatively non-active areas, in order to inhibit the spread of contamination during normal and abnormal conditions, and b) the need to treat the ventilated gas so as to minimize the release of any radioactive or toxic materials. Keeping the two important safety functions is achieved by applying the fire protection for the ventilation system to achieve safety and adequate protection in nuclear applications facilities during fire and accidental criticality conditions.The main purpose of this research is to assist ventilation engineers and experts in nuclear installations for safe operation and maintaining ventilation and air cleaning system during fire accident in nuclear facilities. The research focuses on fire prevention and protection of the ventilation systems in nuclear facilities. High-Efficiency particulate air (HEPA) filters are extremely susceptible to damage when exposed to the effects of fire, smoke, and water; it is the intent of this research to provide the designer with the experience gained over the years from hard lessons learned in protecting HEPA filters from fire. It describes briefly and evaluates the design safety features, constituents and working conditions of ventilation and air cleaning system in nuclear and radioactive industry.This paper provides and

  8. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

  9. External Events Excluding Earthquakes in the Design of Nuclear Power Plants. Safety Guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide provides recommendations and guidance on design for the protection of nuclear power plants from the effects of external events (excluding earthquakes), i.e. events that originate either off the site or within the boundaries of the site but from sources that are not directly involved in the operational states of the nuclear power plant units. In addition, it provides recommendations on engineering related matters in order to comply with the safety objectives and requirements established in the IAEA Safety Requirements publication, Safety of Nuclear Power Plants: Design. It is also applicable to the design and safety assessment of items important to the safety of land based stationary nuclear power plants with water cooled reactors. Contents: 1. Introduction; 2. Application of safety criteria to the design; 3. Design basis for external events; 4. Aircraft crash; 5. External fire; 6. Explosions; 7. Asphyxiant and toxic gases; 8. Corrosive and radioactive gases and liquids; 9. Electromagnetic interference; 10. Floods; 11. Extreme winds; 12. Extreme meteorological conditions; 13. Biological phenomena; 14. Volcanism; 15. Collisions of floating bodies with water intakes and UHS components; Annex I: Aircraft crashes; Annex II: Detonation and deflagration; Annex III: Toxicity limits.

  10. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

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

  12. A series of student design projects for improving and modernizing safety helmets

    NARCIS (Netherlands)

    Beurden, van K.M.M. (Karin); Boer, de J. (Johannes); Stilma, M. (Margot); Teeuw, W.B. (Wouter)

    2014-01-01

    The Saxion Research Centre for Design and Technology employs many students during research projects. This paper discusses a series of student design projects on safety helmets in the Safety@Work project. At construction sites workers are required to wear personal protective equipment during their

  13. Requirements on software lifecycle process (RSLP) for KALIMER digital computer-based MMIS design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Kwon, Kee Choon; Kim, Jang Yeol [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-04-01

    Digital Man Machine Interface System (MMIS) systems of Korea Advanced Liquid MEtal Reactor (KALIMER) may share code, data transmission, data, and process equipment to a greater degree than analog systems. Although this sharing is the basis for many of the advantages of digital systems, it also raises a key concern: a design using shared data or code has the potential to propagate a common-cause or common-mode failure via software errors, thus defeating the redundancy achieved by the hardware architectural structure. Greater sharing of process equipment among functions within a channel increases the consequences of the failure of a single hardware module and reduces the amount of diversity available within a single safety channel. The software safety plan describes the safety analysis implementation tasks that are to be carried out during the software life cycle. Documentation should exist that shows that the safety analysis activities have been successfully accomplished for each life cycle activity group. In particular, the documentation should show that the system safety requirement have been adequately addressed for each life cycle activity group, that no new hazards have been introduced, and that the software requirements, design elements, and code elements that can affect safety have been identified. Because the safety of software can be assured through both the process Verification and Validation (V and V) itself and the V and V of all the intermediate and final products during the software development lifecycle, the development of KALIMER Software Safety Framework (KSSF) must be established. As the first activity for establishing KSSF, we have developed this report, Requirement on Software Life-cycle Process (RSLP) for designing KALIMER digital MMIS. This report is organized as follows. Section I describes the background, definitions, and references of RSLP. Section II describes KALIMER safety software categorization. In Section III, we define the

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Design requirement on KALIMER blanket fuel assembly duct

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

    1998-03-01

    This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs

  19. 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)

  20. Development of design and safety analysis supporting system for casks

    International Nuclear Information System (INIS)

    Ohsono, Katsunari; Higashino, Akira; Endoh, Shuji

    1993-01-01

    Mitsubishi heavy Industries has developed a design and safety analysis supporting system 'CADDIE' (Cask Computer Aided Design, Drawing and Integrated Evaluation System), with the following objectives: (1) Enhancement of efficiency of the design and safety analysis (2) Further advancement of design quality (3) Response to the diversification of design requirements. The features of this system are as follows: (1) The analysis model data common to analyses is established, and it is prepared automatically from the model made by CAD. (2) The input data for the analysis code is available by simple operation of conversation type from the analysis model data. (3) The analysis results are drawn out in diagrams by output generator, so as to facilitate easy observation. (4) The data of material properties, fuel assembly data, etc. required for the analyses are made available as a data base. (J.P.N.)

  1. ARIES-RS safety design and analysis

    International Nuclear Information System (INIS)

    Steiner, D.; El-Guebaly, L.; Herring, S.; Khater, H.; Mogahed, E.; Thayer, R.; Tillack, M.S.

    1997-01-01

    The ARIES-RS safety design and analysis focused on achieving two objectives: (1) The avoidance of sheltering or evacuation in the event of an accident; and (2) the generation of only low-level waste, no greater than Class C. The ARIES-RS baseline design employs V-4Cr-4Ti as the blanket structural material and a low activation ferritic steel in the reflector and shield. In the event of a LOCA, the baseline design first wall maximum temperature falls in the range of 1100-1200 C. For this temperature range, the hazard assessment indicates that the dose at the site boundary will be less than 1 rem per year. Thus, no sheltering or evacuation would be required in the event of a LOCA. Although the baseline design satisfies the first safety objective noted above, a first wall maximum temperature of ∝1100-1200 C would likely compromise the integrity of the vanadium blanket structure and would require blanket replacement following such a temperature excursion. To avoid this situation, a modified blanket design incorporating supplemental heat removal is also proposed. Preliminary analysis of this modified design suggests that the first wall maximum temperature can be kept below the temperature range of concern, ∝1000-1100 C, in the event of a LOCA. When the ferritic steel used in the reflector and shield is one reduced in Ir and Ag impurities, all in-vessel components qualify for near-surface shallow land burial as Class C low-level waste. (orig.)

  2. Safety design features for current UK advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

    1981-01-15

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

  3. Safety design features for current UK advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.; Cobb, E.C.

    1981-01-01

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

  4. Interaction between periodic in-service inspection requirements and design

    International Nuclear Information System (INIS)

    Prot, A.C.; Saglio, R.

    1979-03-01

    After reviewing the requirements specific of periodic In-Service Inspection related to safety problems, especially for the pressure vessels, and taking into account the experience gained with several PWR reactors, the authors show these requirements could lead to modify the primary circuit design

  5. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

  6. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    International Nuclear Information System (INIS)

    Park, Jaekwan; Suh, Yongsuk

    2014-01-01

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

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

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, P.; Boucau, J.; Cantineau, B.; Doumont, C.; Mared, A.

    2000-01-01

    DART is the acronym for Design Analysis Re-engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

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

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, B.; Boucau, J.; Cantineau, B.; Mared, A.

    2001-01-01

    DART is the acronym for Design Analysis Re-Engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can then be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

  9. 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)

  10. 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

  11. An aspect-oriented approach for designing safety-critical systems

    Science.gov (United States)

    Petrov, Z.; Zaykov, P. G.; Cardoso, J. P.; Coutinho, J. G. F.; Diniz, P. C.; Luk, W.

    The development of avionics systems is typically a tedious and cumbersome process. In addition to the required functions, developers must consider various and often conflicting non-functional requirements such as safety, performance, and energy efficiency. Certainly, an integrated approach with a seamless design flow that is capable of requirements modelling and supporting refinement down to an actual implementation in a traceable way, may lead to a significant acceleration of development cycles. This paper presents an aspect-oriented approach supported by a tool chain that deals with functional and non-functional requirements in an integrated manner. It also discusses how the approach can be applied to development of safety-critical systems and provides experimental results.

  12. UK experience of safety requirements for thermal reactor stations

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. Integrating Safeguards and Security with Safety into Design

    International Nuclear Information System (INIS)

    Bean, Robert S.; Hockert, John W.; Hebditch, David J.

    2009-01-01

    There is a need to minimize security risks, proliferation hazards, and safety risks in the design of new nuclear facilities in a global environment of nuclear power expansion, while improving the synergy of major design features and raising operational efficiency. In 2008, the U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA) launched the Next Generation Safeguards Initiative (NGSI) covering many safeguards areas. One of these, launched by NNSA with support of the DOE Office of Nuclear Energy, was a multi-laboratory project, led by the Idaho National Laboratory (INL), to develop safeguards by design. The proposed Safeguards-by-Design (SBD) process has been developed as a structured approach to ensure the timely, efficient, and cost effective integration of international safeguards and other nonproliferation barriers with national material control and accountability, physical security, and safety objectives into the overall design process for the nuclear facility lifecycle. A graded, iterative process was developed to integrate these areas throughout the project phases. It identified activities, deliverables, interfaces, and hold points covering both domestic regulatory requirements and international safeguards using the DOE regulatory environment as exemplar to provide a framework and guidance for project management and integration of safety with security during design. Further work, reported in this paper, created a generalized SBD process which could also be employed within the licensed nuclear industry and internationally for design of new facilities. Several tools for integrating safeguards, safety, and security into design are discussed here. SBD appears complementary to the EFCOG TROSSI process for security and safety integration created in 2006, which focuses on standardized upgrades to enable existing DOE facilities to meet a more severe design basis threat. A collaborative approach is suggested.

  15. 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

  16. 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.

  17. 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

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

    International Nuclear Information System (INIS)

    Yllera, Javier

    2013-01-01

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

  19. Meeting Human Reliability Requirements through Human Factors Design, Testing, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Boring

    2007-06-01

    In the design of novel systems, it is important for the human factors engineer to work in parallel with the human reliability analyst to arrive at the safest achievable design that meets design team safety goals and certification or regulatory requirements. This paper introduces the System Development Safety Triptych, a checklist of considerations for the interplay of human factors and human reliability through design, testing, and modeling in product development. This paper also explores three phases of safe system development, corresponding to the conception, design, and implementation of a system.

  20. 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.

  1. 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

  2. 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)

  3. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  4. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  5. Basic design requirements for indigenous irradiator

    International Nuclear Information System (INIS)

    Anwar Abd Rahman; Rosli Darmawan; Mohd Arif Hamzah; Fadil Ismail; Muhd Nor Atan

    2007-01-01

    Most of the irradiators owned by Nuclear Malaysia are imported from other countries. The irradiators are used for various applications such as Research and Development, agriculture and industry. There is a plan to develop locally made multi-purpose gamma irradiator in 9th Malaysia Plan which equipped with better safety features. This paper will discuss the basic requirements for the design of the irradiator. (Author)

  6. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    International Nuclear Information System (INIS)

    Duckwitz, Noel

    2011-01-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets,' safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, 'Facility Safety,' and the expectations of DOE-STD-1189-2008, 'Integration of Safety into the Design Process,' provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  7. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

    2004-01-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  8. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  9. Safety design guide for pipe rupture protection 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 pipe rupture protection identifies high-energy systems in which pipe ruptures must be postulated to occur, as well as systems that must be protected from the dynamic effects of such ruptures. Dynamic effects considered in this SDG consist of pipe whip (including missiles generated by pipe ruptures, if any) and jet impingement, Requirements for protection against the dynamic effects of a postulated pipe rupture and method of protection of essential structures, systems and components are specified for these effects. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 5 refs. (Author) .new

  10. Metrics design for safety assessment

    NARCIS (Netherlands)

    Luo, Yaping; van den Brand, M.G.J.

    2016-01-01

    Context:In the safety domain, safety assessment is used to show that safety-critical systems meet the required safety objectives. This process is also referred to as safety assurance and certification. During this procedure, safety standards are used as development guidelines to keep the risk at an

  11. Protection against internal fires and explosions in the design of nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Experience of the past two decades in the operation of nuclear power plants and modern analysis techniques confirm that fire may be a real threat to nuclear safety and should receive adequate attention from the beginning of the design process throughout the life of the plant. Within the framework of the NUSS programme, a Safety Guide on fire protection had therefore been developed to enlarge on the general requirements given in the Code. Since its first publication in 1979, there has been considerable development in protection technology and analysis methods and after the Chernobyl accident it was decided to revise the existing Guide. This Safety Guide supplements the requirements established in Safety of Nuclear Power Plants: Design. It supersedes Safety Series No. 50-SG-D2 (Rev. 1), Fire Protection in Nuclear Power Plants: A Safety Guide, issued in 1992.The present Safety Guide is intended to advise designers, safety assessors and regulators on the concept of fire protection in the design of nuclear power plants and on recommended ways of implementing the concept in some detail in practice

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

    International Nuclear Information System (INIS)

    2004-01-01

    industrial, agricultural, medical, research and teaching applications. Facilities using radiation or radioactive material. And satellites and radiothermal generators using radiation sources or reactors. The requirements also cover emergencies arising from radiation sources of an unknown or untraceable origin. The requirements apply to actions in preparedness and response for emergencies involving hazards associated with ionizing radiation only. The requirements do not apply to preparedness or response for emergencies involving hazards associated with non-ionizing radiation such as microwave, ultraviolet or infrared radiation. The IAEA has issued Safety Fundamentals publications relating to nuclear installations and to radioactive waste management. In addition, the IAEA has issued Safety Series and Safety Standards Series publications that have established requirements relating to emergency preparedness for the safe use of radiation sources, the transport of radioactive material, legal and governmental infrastructure, the design and operation of nuclear power plants, and the design and operation of research reactors. This Safety Requirements publication incorporates, elaborates on, augments and structures all the requirements for emergency preparedness and response established in these previously issued safety standards. To this effect some requirements established in these other safety standards have been incorporated as quotations in this publication (revised text is shown in square brackets and omissions are indicated by ellipses). In other cases footnotes state where related requirements have been established

  13. Design of safety-critical systems using the complementarities of success and failure domains with a case study

    International Nuclear Information System (INIS)

    Ahmed, Rizwan; Koo, June Mo; Jeong, Yong Hoon; Heo, Gyunyoung

    2011-01-01

    A safety-critical system has to qualify the performance-related requirements and the safety-related requirements simultaneously. Conceptually, design processes should consider both of them simultaneously but the practices do not and/or cannot follow such a theoretical approach due to the limitation of design resources. From our experience, we found that safety-related functions must be simultaneously resolved with the development of performance-related functions, particularly, in case of safety-critical systems. Since, success and failure domain analyses are essential for the investigation of performance-related and safety-related requirements, respectively, we articulated our perception to Axiomatic Design (AD), Fault Tree Analysis (FTA), and TRIZ. A design evolution procedure considering feedbacks from AD to identify functional couplings, TRIZ methodology to explore uncoupling solutions and FTA to improve reliability in a systematic way is presented here. A case study regarding design of safety injection tank installed in a nuclear power plant is also included to illustrate the proposed framework. It is expected that several iterations between AD-TRIZ-FTA would result into an optimized design which could be tested against the desired performance and safety criteria.

  14. Design of safety-critical systems using the complementarities of success and failure domains with a case study

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Rizwan; Koo, June Mo [Department of Nuclear Engineering, Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of); Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.k [Department of Nuclear Engineering, Kyung Hee University, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2011-01-15

    A safety-critical system has to qualify the performance-related requirements and the safety-related requirements simultaneously. Conceptually, design processes should consider both of them simultaneously but the practices do not and/or cannot follow such a theoretical approach due to the limitation of design resources. From our experience, we found that safety-related functions must be simultaneously resolved with the development of performance-related functions, particularly, in case of safety-critical systems. Since, success and failure domain analyses are essential for the investigation of performance-related and safety-related requirements, respectively, we articulated our perception to Axiomatic Design (AD), Fault Tree Analysis (FTA), and TRIZ. A design evolution procedure considering feedbacks from AD to identify functional couplings, TRIZ methodology to explore uncoupling solutions and FTA to improve reliability in a systematic way is presented here. A case study regarding design of safety injection tank installed in a nuclear power plant is also included to illustrate the proposed framework. It is expected that several iterations between AD-TRIZ-FTA would result into an optimized design which could be tested against the desired performance and safety criteria.

  15. Design for safety: theoretical framework of the safety aspect of BIM system to determine the safety index

    Directory of Open Access Journals (Sweden)

    Ai Lin Evelyn Teo

    2016-12-01

    Full Text Available Despite the safety improvement drive that has been implemented in the construction industry in Singapore for many years, the industry continues to report the highest number of workplace fatalities, compared to other industries. The purpose of this paper is to discuss the theoretical framework of the safety aspect of a proposed BIM System to determine a Safety Index. An online questionnaire survey was conducted to ascertain the current workplace safety and health situation in the construction industry and explore how BIM can be used to improve safety performance in the industry. A safety hazard library was developed based on the main contributors to fatal accidents in the construction industry, determined from the formal records and existing literature, and a series of discussions with representatives from the Workplace Safety and Health Institute (WSH Institute in Singapore. The results from the survey suggested that the majority of the firms have implemented the necessary policies, programmes and procedures on Workplace Safety and Health (WSH practices. However, BIM is still not widely applied or explored beyond the mandatory requirement that building plans should be submitted to the authorities for approval in BIM format. This paper presents a discussion of the safety aspect of the Intelligent Productivity and Safety System (IPASS developed in the study. IPASS is an intelligent system incorporating the buildable design concept, theory on the detection, prevention and control of hazards, and the Construction Safety Audit Scoring System (ConSASS. The system is based on the premise that safety should be considered at the design stage, and BIM can be an effective tool to facilitate the efforts to enhance safety performance. IPASS allows users to analyse and monitor key aspects of the safety performance of the project before the project starts and as the project progresses.

  16. 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

  17. Design requirement for electrical system of an advanced research reactor

    International Nuclear Information System (INIS)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S.

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system

  18. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  19. Evolution of general design requirements for french pressurized water reactors

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Rollinger, F.

    1988-10-01

    The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5

  20. Design, fabrication and erection of steel structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety Standard for Civil Engineering Structures Important to Safety of Nuclear Facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design, fabrication and erection of steel structures important to safety

  1. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.

  2. Integrated Safety in Design

    DEFF Research Database (Denmark)

    Schultz, Casper Siebken; Jørgensen, Kirsten

    2014-01-01

    An on-going research project investigates the inclusion of health and safety considerations in the design phase as a means to achieve a higher level of health and safety in the construction industry. Moreover, the approach is coupled to the overall quality efforts. Two architectural firms and two...... consulting engineering firms are project participants. The hypothesis is that health and safety problems in execution can be prevented through better planning in the early stages of the construction processes and that accidents are prevented by providing safety. In the first stage of the research project...... a theoretical framework is developed from a combination of existing literature on health and safety and a mapping of existing practices based on interviews in all four companies. The interviews revealed that the basic knowledge on OHS among architects and engineers is limited. Also currently designers typically...

  3. 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

  4. 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

  5. Design and safety of the Sizewell pressurized water reactor

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the UK the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear power focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The long-term consequences can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking. (author)

  6. 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

  7. Impact of ITER liquid metal design options on safety level and licensing - Sweden

    International Nuclear Information System (INIS)

    Harfors, C.; Devell, L.; Johansson, Kjell; Lundell, B.; Rolandsson, S.

    1993-01-01

    The safety level and licensability of five design options for ITER coolant, breeding material and structural material are assessed, with emphasis on some specified accident scenarios. The safety level is assessed in terms of barrier requirements and the feasibility to construct and qualify such a barrier. The licensability in Sweden of each design option is assessed based on the indicated safety level and on a judgement of the technical feasibility to construct and qualify the ITER tokamak itself, based on the selected design option. 20 refs

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

    International Nuclear Information System (INIS)

    1985-07-01

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

  9. Safety aspects of designs for future light water reactors (evolutionary reactors)

    International Nuclear Information System (INIS)

    1993-07-01

    The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

  10. 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

  11. Research and design of hanger and support series of nuclear safety class process piping

    International Nuclear Information System (INIS)

    Mao Chengzhang; Shi Jiemin

    1995-12-01

    Hangers and supports of nuclear safety class piping are an important part of primary system piping in a nuclear power plant. They will directly affect the reliability of operation, the period at construction and the investment for a nuclear power plant. It is an absolutely necessary job for Pakistan Chashma Nuclear Power Plant Project to research and design a series of piping supports in accordance with ASME-III NF. It is also an important designing for developing nuclear power plant later in China. After working over two years, a series of piping supports of nuclear safety class which have 57 types and more than 2460 specifications have been designed. This series is perfect, and can satisfy the requirements of piping final designing for nuclear power plant. This series of hangers and supports is mainly used in the process piping of nuclear safety class 1,2,3. They can also be used in other piping of nuclear safety class and piping with aseismic requirement of non-nuclear safety class

  12. 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

  13. 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)

  14. 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.)

  15. Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

    International Nuclear Information System (INIS)

    G. L. Sharp; R. T. McCracken

    2004-01-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

  16. Safety culture in design. Final report

    International Nuclear Information System (INIS)

    Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M.; Kahlbom, U.; Rollenhagen, C.

    2013-04-01

    In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

  17. Safety culture in design. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M. [VTT Technical Research Centre of Finland, Espoo (Finland); Kahlbom, U. [Risk Pilot AB, Stockholm (Sweden); Rollenhagen, C. [Vattenfall, Stockholm, (Sweden)

    2013-04-15

    In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

  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. 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

  1. 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

  2. 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

  3. Breeder design for enhanced performance and safety characteristics

    International Nuclear Information System (INIS)

    Fischer, G.J.; Atefi, B.; Yang, J.W.; Galperin, A.; Segev, M.

    1980-01-01

    A fast breeder reactor design has been created which offers a considerably extended fuel cycle and excellent performance characteristics. An example of a core designed to operate on a ten-year fuel cycle is described in some detail. Use of metal fuel along with a moderator such as beryllium oxide dispersed throughout the core provides both design flexibility and safety advantages such as a strong Doppler feedback and limited sodium void reactivity gain. Local power variations are small for the entire cycle; control requirements are also modest, and fuel cycle costs are low

  4. Design of marine structures with improved safety for environment

    International Nuclear Information System (INIS)

    Klanac, Alan; Varsta, Petri

    2011-01-01

    The paper describes a method for design of marine structures with increased safety for environment, considering also the required investment costs as well as the aspects of risk distribution onto the maritime stakeholders. Practically, the paper seeks to answer what is the optimal amount that should be invested into certain safety measure for any given vessel. Due to the uneven distribution of risk, as well as the differing impact of costs emerging from safety improvements, stakeholders experience conflicting ranking of alternatives. To solve this multi-stakeholder decision-making problem, in which each stakeholder is a decision-maker, the method applies concepts of group decision-making theory, namely the Game Theory. The method fosters axiomatic definition of the optimum solution, arguing that the solution, or the final selected design, should satisfy the non-dominance, efficiency, and fairness. These three are thoroughly discussed in terms of structural design, especially the latter. Considering the coupling of environmental risk and structural design, the method also builds on the preference structure of four maritime stakeholders: yards, owners, oil receivers and the public, who either share the risks or directly influence structural design. Method is presented on a practical study of structural design of a tanker with a crashworthy side structure that is capable of reducing the risk of collision. The outcome of this study outlines a number of possibilities for successful improvement of tanker safety that can benefit, concurrently, all maritime stakeholders.

  5. Safety design of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

    1984-01-01

    Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

  6. Radiation protection aspects of design for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    . The IAEA takes seriously the enduring challenge for users and regulators everywhere: that of ensuring a high level of safety in the use of nuclear materials and radiation sources around the world. Their continuing utilization for the benefit of humankind must be managed in a safe manner, and the IAEA safety standards are designed to facilitate the achievement of that goal. This Safety Guide has been prepared as a part of the IAEA programme on safety standards for nuclear power plants. It includes recommendations on how to satisfy the requirements established in the Safety Requirements publication on the Safety of Nuclear Power Plants: Design. It addresses the provisions that should be made in the design of nuclear power plants in order to protect site personnel, the public and the environment against radiological hazards for operational states, decommissioning and accident conditions. The recommendations on radiation protection provided in this Safety Guide are consistent with the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS), which were jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the IAEA, the International Labour Organisation (ILO), the OECD Nuclear Energy Agency (OECD/NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO). This Safety Guide supersedes Safety Series No. 50-SG-D9, Design Aspects of Radiation Protection for Nuclear Power Plants, published in 1985. Effective radiation protection is a combination of good design, high quality construction and proper operation. Procedures that address the radiation protection aspects of operation are covered in the Safety Guide on Radiation Protection and Radioactive Waste Management in the operation of Nuclear Power Plants

  7. 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)

  8. 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

  9. 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

  10. 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)

  11. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design

  12. 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.

  13. Towards Requirements in Systems Engineering for Aerospace IVHM Design

    Science.gov (United States)

    Saxena, Abhinav; Roychoudhury, Indranil; Lin, Wei; Goebel, Kai

    2013-01-01

    Health management (HM) technologies have been employed for safety critical system for decades, but a coherent systematic process to integrate HM into the system design is not yet clear. Consequently, in most cases, health management resorts to be an after-thought or 'band-aid' solution. Moreover, limited guidance exists for carrying out systems engineering (SE) on the subject of writing requirements for designs with integrated vehicle health management (IVHM). It is well accepted that requirements are key to developing a successful IVHM system right from the concept stage to development, verification, utilization, and support. However, writing requirements for systems with IVHM capability have unique challenges that require the designers to look beyond their own domains and consider the constraints and specifications of other interlinked systems. In this paper we look at various stages in the SE process and identify activities specific to IVHM design and development. More importantly, several relevant questions are posed that system engineers must address at various design and development stages. Addressing these questions should provide some guidance to systems engineers towards writing IVHM related requirements to ensure that appropriate IVHM functions are built into the system design.

  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. 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.

  16. 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.

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  18. Risk allocation approach to reactor safety design and evaluation

    International Nuclear Information System (INIS)

    Gokcek, O.; Temme, M.I.; Derby, S.L.

    1978-01-01

    This paper describes a risk allocation technique used for determining nuclear power plant design reliability requirements. The concept of risk allocation-optimum choice of safety function reliabilities under a maximum risk constraint - is described. An example of risk allocation is presented to demonstrate the application of the methodology

  19. A proposal for safety design philosophy of HTGR for coupling hydrogen production plant

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Imai, Yoshiyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

    2013-06-01

    Japan Atomic Energy Agency (JAEA) has been conducting research and development for hydrogen production utilizing heat from High Temperature Gas-cooled Reactors (HTGRs). Towards the realization of nuclear hydrogen production, coupled hydrogen production plants should not be treated as an extension of a nuclear plant in order to open the door for the entry of non-nuclear industries as well as assuring reactor safety against postulated abnormal events initiated in the hydrogen production plants. Since hydrogen production plant utilizing nuclear heat has never been built in the world, little attention has been given to the establishment of a safety design for such system including the High Temperature engineering Test Reactor (HTTR). In the present study, requirements in order to design, construct and operate hydrogen production plants under conventional chemical plant standards are identified. In addition, design considerations for safety design of nuclear facility are suggested. Furthermore, feasibility of proposed safety design and design considerations are evaluated. (author)

  20. Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Kim, Sang Ho; Choi, Jae Young

    2013-01-01

    Highlights: • We newly propose the design concept of integrated passive safety system (IPSS). • It has five safety functions for decay heat removal and severe accident mitigation. • Simulations for IPSS show that core melt does not occur in accidents with SBO. • IPSS can achieve the passive in-vessel retention and ex-vessel cooling strategy. • The applicability of IPSS is high due to the installation outside the containment. -- Abstract: The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the

  1. Resolution of thermal-hydraulic safety and licensing issues for the system 80+trademark design

    International Nuclear Information System (INIS)

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E.

    1995-01-01

    The System 80+ trademark Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC's new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs

  2. 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

  3. Design review report for modifications to RMCS safety class equipment

    International Nuclear Information System (INIS)

    Corbett, J.E.

    1997-01-01

    This report documents the completion of the formal design review for modifications to the Rotary Mode Core Sampling (RMCS) safety class equipment. These modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to approve the Engineering Change Notices affecting safety class equipment used in the RMCS system. The conclusion reached by the review committee was that these changes are acceptable

  4. Design review report for modifications to RMCS safety class equipment

    Energy Technology Data Exchange (ETDEWEB)

    Corbett, J.E.

    1997-05-30

    This report documents the completion of the formal design review for modifications to the Rotary Mode Core Sampling (RMCS) safety class equipment. These modifications are intended to support core sampling operations in waste tanks requiring flammable gas controls. The objective of this review was to approve the Engineering Change Notices affecting safety class equipment used in the RMCS system. The conclusion reached by the review committee was that these changes are acceptable.

  5. Test design requirements for overcoring stress measurements

    International Nuclear Information System (INIS)

    Stickney, R.G.

    1985-12-01

    This document establishes the test design requirements for a series of overcoring stress measurements to be performed in the Exploratory Shaft Facility. The stress measurements will be made to determine the in situ state of stress within the candidate repository horizon and to determine the magnitude and distribution of the stresses induced by the mined openings of the facility. The overcoring technique involves the measurement of strain (or deformation) in a volume of rock as the stress acting on the rock volume is relieved. This document presents an overview of the measurements, including objectives and rationale for the measurements. A description of the measurements is included. The support requirements are identified as are constraints for the design of the measurements. Discussions on Quality Assurance and Safety are also included in the document. 13 refs

  6. 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.

  7. 29 CFR 1910.36 - Design and construction requirements for exit routes.

    Science.gov (United States)

    2010-07-01

    ... construction requirements for exit routes. (a) Basic requirements. Exit routes must meet the following design... your workplace, consult NFPA 101-2000, Life Safety Code. (c) Exit discharge. (1) Each exit discharge... route must be adequate. (1) Exit routes must support the maximum permitted occupant load for each floor...

  8. Mitigating construction safety risks using prevention through design.

    Science.gov (United States)

    Gangolells, Marta; Casals, Miquel; Forcada, Núria; Roca, Xavier; Fuertes, Alba

    2010-04-01

    Research and practice have demonstrated that decisions made prior to work at construction sites can influence construction worker safety. However, it has also been argued that most architects and design engineers possess neither the knowledge of construction safety nor the knowledge of construction processes necessary to effectively perform Construction Hazards Prevention through Design (CHPtD). This paper introduces a quantitative methodology that supports designers by providing a way to evaluate the safety-related performance of residential construction designs using a risk analysis-based approach. The methodology compares the overall safety risk level of various construction designs and ranks the significance of the various safety risks of each of these designs. The methodology also compares the absolute importance of a particular safety risk in various construction designs. Because the methodology identifies the relevance of each safety risk at a particular site prior to the construction stage, significant risks are highlighted in advance. Thus, a range of measures for mitigating safety risks can then be implemented during on-site construction. The methodology is specially worthwhile for designers, who can compare construction techniques and systems during the design phase and determine the corresponding level of safety risk without their creative talents being restricted. By using this methodology, construction companies can improve their on-site safety performance. Copyright 2010 Elsevier Ltd. All rights reserved.

  9. 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.

  10. Safety design philosophy of Mitsubishi PWRs

    International Nuclear Information System (INIS)

    Hakata, T.; Kitamura, T.

    1993-01-01

    The basic safety design philosophy of Mitsubishi pressurized water reactors (PWRs) is discussed and compared with the British PWR. PWR plants are designed in accordance with the Japanese regulatory guidelines which are similar to American and International Atomic Energy Agency (IAEA) safety criteria and are based on defence-in-depth principles. The high reliability of nuclear power plants is especially emphasized in Mitsubishi PWRs, and this has been demonstrated by the good operating experience of PWR plants in Japan. The safety system designs of six key items, which were discussed in the recent review of overseas designs by British utilities, are addressed to show the difference in the design philosophy between the United Kingdom and Japan. (Author)

  11. 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

  12. 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

  13. 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...

  14. Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

    International Nuclear Information System (INIS)

    Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

    2003-01-01

    The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

  15. Safety Consideration for a Wet Interim Spent Fuel Store at Conceptual Design Stage

    International Nuclear Information System (INIS)

    Astoux, Marion

    2014-01-01

    EDF Energy plans to build and operate two UK EPRs at the Hinkley Point C (HPC) site in Somerset, England. Spent fuel from the UK EPRs will need to be managed from the time it is discharged from the reactor until it is ultimately disposed of and this will involve storing the spent fuel for a period in the fuel building and thereafter in a dedicated interim facility until it can be emplaced within the UK Geological Disposal Facility. EDF Energy has proposed that this interim store should be located on the Hinkley Point site which is consistent with UK policy. This Interim Spent Fuel Store (ISFS) will have the capability to store for at least one hundred years the spent fuel arising from the operation of the two EPR units (sixty years operation). Therefore, specificities regarding the lifetime of the facility have to be accounted for its design. The choice of interim storage technology was considered in some depth for the HPC project and wet storage (pool) was selected. The facility is currently at conceptual design stage, although its construction will be part of main site construction phase. Safety functions and safety requirements for this storage facility have been defined, in compliance with WENRA 'Waste and Spent Fuel Storage - Safety Reference Level Report' and IAEA Specific Safety Guide no. 15 'Storage of Spent Nuclear Fuel'. EDF technical know-how, operational feedback on existing storage pools, UK regulatory context and Fukushima experience feedback have also been accounted for. Achievement of the safety functions as passively as reasonably practicable is a key issue for the design, especially in accident situations. Regarding lifetime aspects, ageing management of equipments, optimisation of the refurbishment, climate change, passivity of the facility, and long-term achievement of the safety functions are among the subjects to consider. Adequate Operational Limits and Conditions will also have to be defined, to enable the long-term achievement of the safety

  16. Appendix C: safety design rationale

    International Nuclear Information System (INIS)

    Ghose, S.

    1985-01-01

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

  17. Request from nuclear fuel cycle and criticality safety design

    International Nuclear Information System (INIS)

    Hamasaki, Manabu; Sakashita, Kiichiro; Natsume, Toshihiro

    2005-01-01

    The quality and reliability of criticality safety design of nuclear fuel cycle systems such as fuel fabrication facilities, fuel reprocessing facilities, storage systems of various forms of nuclear materials or transportation casks have been largely dependent on the quality of criticality safety analyses using qualified criticality calculation code systems and reliable nuclear data sets. In this report, we summarize the characteristics of the nuclear fuel cycle systems and the perspective of the requirements for the nuclear data, with brief comments on the recent issue about spent fuel disposal. (author)

  18. Radiation protection aspects in the design of nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    . The IAEA takes seriously the enduring challenge for users and regulators everywhere: that of ensuring a high level of safety in the use of nuclear materials and radiation sources around the world. Their continuing utilization for the benefit of humankind must be managed in a safe manner, and the IAEA safety standards are designed to facilitate the achievement of that goal. This Safety Guide has been prepared as a part of the IAEA programme on safety standards for nuclear power plants. It includes recommendations on how to satisfy the requirements established in the Safety Requirements publication on the Safety of Nuclear Power Plants: Design. It addresses the provisions that should be made in the design of nuclear power plants in order to protect site personnel, the public and the environment against radiological hazards for operational states, decommissioning and accident conditions. The recommendations on radiation protection provided in this Safety Guide are consistent with the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS), which were jointly sponsored by the Food and Agriculture Organization of the United Nations (FAO), the IAEA, the International Labour Organisation (ILO), the OECD Nuclear Energy Agency (OECD/NEA), the Pan American Health Organization (PAHO) and the World Health Organization (WHO). This Safety Guide supersedes Safety Series No. 50-SG-D9, Design Aspects of Radiation Protection for Nuclear Power Plants, published in 1985. Effective radiation protection is a combination of good design, high quality construction and proper operation. Procedures that address the radiation protection aspects of operation are covered in the Safety Guide on Radiation Protection and Radioactive Waste Management in the operation of Nuclear Power Plants

  19. Laser safety in design of near-infrared scanning LIDARs

    Science.gov (United States)

    Zhu, X.; Elgin, D.

    2015-05-01

    3D LIDARs (Light Detection and Ranging) with 1.5μm nanosecond pulse lasers have been increasingly used in different applications. The main reason for their popularity is that these LIDARs have high performance while at the same time can be made eye-safe. Because the laser hazard effect on eyes or skin at this wavelength region (industrial mining applications. We have incorporated the laser safety requirements in the LIDAR design and conducted laser safety analysis for different operational scenarios. While 1.5μm is normally said to be the eye-safe wavelength, in reality a high performance 3D LIDAR needs high pulse energy, small beam size and high pulse repetition frequency (PRF) to achieve long range, high resolution and high density images. The resulting radiant exposure of its stationary beam could be many times higher than the limit for a Class 1 laser device. Without carefully choosing laser and scanning parameters, including field-of-view, scan speed and pattern, a scanning LIDAR can't be eye- or skin-safe based only on its wavelength. This paper discusses the laser safety considerations in the design of eye-safe scanning LIDARs, including laser pulse energy, PRF, beam size and scanning parameters in two basic designs of scanning mechanisms, i.e. galvanometer based scanner and Risley prism based scanner. The laser safety is discussed in terms of device classification, nominal ocular hazard distance (NOHD) and safety glasses optical density (OD).

  20. Analysis of compatibility of current Czech initial documentation in the area of technical assurance of nuclear safety with the requirements of the EUR document

    International Nuclear Information System (INIS)

    Zdebor, J.; Zdebor, R.; Kratochvil, L.

    2001-11-01

    The publication is structured as follows: Description of existing documentation. General requirements, goals, principles and design principles: Documents being compared; Method of comparison; Results and partial evaluation of comparison of requirements between EUR and Czech regulations (basic goals and safety philosophy; quantitative safety objectives; basic design requirements; extended design requirements; external and internal threats; technical requirements; site conditions); Summary of the comparison of safety requirements. Comparison of requirements for the systems: Requirements for the nuclear reactor unit systems; Barrier systems (fuel system; reactor cooling system; containment system); Remaining systems (control systems; protection systems; coolant makeup and purification system; residual heat removal system; emergency cooling system; power systems); Common technical requirements for systems (technical requirements for systems; internal and external events). (P.A.)

  1. Design provisions for safety

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1983-01-01

    Design provisions for safety of nuclear power plants are based on a well balanced concept: the public is protected against a release of radioactive material by multiple barriers. These barriers are protected according to a 'defence-in-depth' principle. The reactor safety concept is primarily aimed at the prevention of accidents, especially fuel damage. Additionally, measures for consequence limitation are provided in order to prevent a severe release of radioactivity to the environment. However, it is difficult to judge the overall effectiveness of such devices. In a comprehensive safety analysis it has to be shown that the protection systems and safeguards work with sufficient reliability in the event of an accident. For the reliability assessment deterministic criteria (single failure, redundancy, fail-safe, demand for diversity) play an important role. Increasing efforts have been made to assess reliability quantitatively by means of probabilistic methods. It is now usual to perform reliability analyses of essential systems of nuclear power plants in the course of licensing procedures. As an additional level of emergency measures for a further reduction of hazards a reasonable amount of accident information has to be transferred. Operational experience may be considered as an important feedback to the design of plant safety features. Operator training has to include, besides skill in performing of operating procedures, the training of a flexible response to different accident situations. Experience has shown that the design provisions for safety could prevent dangerous release of the radioactive material to the environment after an accident has occurred. For future developments of reactor safety, extensive analyses of operating experience are of great importance. The main goal should be to enhance the reliability of measures for accident prevention, which prevent the core from meltdown or other damages

  2. Safety Analysis for Key Design Features of KALIMER-600 Design Concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S

    2007-02-15

    This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies.

  3. Design safety improvements of Kozloduy NPP

    International Nuclear Information System (INIS)

    Hinovski, I.

    1999-01-01

    Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

  4. Safety in the design of production lines

    DEFF Research Database (Denmark)

    Dyhrberg, Mette Bang; Broberg, Ole; Jacobsen, Peter

    2006-01-01

    This paper is a case study report on how safety considerations were handled in the process of redesigning a production line. The design process was characterized as a specification and negotiation process between engineers from the company and the supplier organization. The new production line...... in the specification material nor in their face-to-face meetings with the supplier. Safety aspects were not part of their work practice. On this basis, it was suggested that formal guidelines or procedures for integrating safety in the design of production lines would have no effect. Instead, the researchers set up...... became safer, but not as a result of any intentional plan to integrate safety aspects into the design process. Instead, the supplier’s design of a new piece of equipment had a higher built-in safety level. The engineering team in the company was aware of the importance of safety aspects neither...

  5. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  6. IAEA safety standards and approach to safety of advanced reactors

    International Nuclear Information System (INIS)

    Gasparini, M.

    2004-01-01

    The paper presents an overview of the IAEA safety standards including their overall structure and purpose. A detailed presentation is devoted to the general approach to safety that is embodied in the current safety requirements for the design of nuclear power plants. A safety approach is proposed for the future. This approach can be used as reference for a safe design, for safety assessment and for the preparation of the safety requirements. The method proposes an integration of deterministic and risk informed concepts in the general frame of a generalized concept of safety goals and defence in depth. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor including small and medium sized reactors with innovative safety features.(author)

  7. The Role of Probabilistic Design Analysis Methods in Safety and Affordability

    Science.gov (United States)

    Safie, Fayssal M.

    2016-01-01

    For the last several years, NASA and its contractors have been working together to build space launch systems to commercialize space. Developing commercial affordable and safe launch systems becomes very important and requires a paradigm shift. This paradigm shift enforces the need for an integrated systems engineering environment where cost, safety, reliability, and performance need to be considered to optimize the launch system design. In such an environment, rule based and deterministic engineering design practices alone may not be sufficient to optimize margins and fault tolerance to reduce cost. As a result, introduction of Probabilistic Design Analysis (PDA) methods to support the current deterministic engineering design practices becomes a necessity to reduce cost without compromising reliability and safety. This paper discusses the importance of PDA methods in NASA's new commercial environment, their applications, and the key role they can play in designing reliable, safe, and affordable launch systems. More specifically, this paper discusses: 1) The involvement of NASA in PDA 2) Why PDA is needed 3) A PDA model structure 4) A PDA example application 5) PDA link to safety and affordability.

  8. 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

  9. Fire Safety Design of Wood Structures

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl

    2006-01-01

    Lecture Notes on Fire Safety Design of Wood Structures including charring of wood and load bearing capacity of beams, columns, and connections.......Lecture Notes on Fire Safety Design of Wood Structures including charring of wood and load bearing capacity of beams, columns, and connections....

  10. Safety analysis for key design features of KALIMER-600 design concept

    International Nuclear Information System (INIS)

    Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S.

    2005-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed

  11. 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.)

  12. Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

    Energy Technology Data Exchange (ETDEWEB)

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E. [ABB-Combustion Engineering, Windsor, CT (United States)] [and others

    1995-09-01

    The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

  13. Steam generator design requirements for ACR-1000

    International Nuclear Information System (INIS)

    Subash, S.; Hau, K.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) has developed the ACR-1000 (Advanced CANDU Reactor-1000 ) to meet market expectations for enhanced safety of plant operation, high capacity factor, low operating cost, increased operating life, simple component replacement, reduced capital cost, and shorter construction schedule. The ACR-1000 design is based on the use of horizontal fuel channels surrounded by a heavy water moderator, the same feature as in all CANDU reactors. The major innovation in the ACR-1000 is the use of low enriched uranium fuel, and light water as the coolant, which circulates in the fuel channels. This results in a compact reactor core design and a reduction of heavy water inventory, both contributing to a significant decrease in capital cost per MWe produced. The ACR-1000 plant is a two-unit, integrated plant with each unit having a nominal gross output of about 1165 MWe with a net output of approximately 1085 MWe. The plant design is adaptable to a single unit configuration, if required. This paper focuses on the technical considerations that went into developing some of the important design requirements for the steam generators for the ACR-1000 plant and how these requirements are specified in the Technical Specification, which is the governing document for the steam generator (SG) detail design. Layout of these SGs in the plant is briefly described and their impacts on the SG design. (author)

  14. ELFR: The European Lead Fast Reactor. Design, Safety Approach and Safety Characteristics

    International Nuclear Information System (INIS)

    Alemberti, Alessandro

    2012-01-01

    • In the framework of the LEADER project, the safety approach for a Lead cooled fast reactor has been defined and, in particular, all the possible challenges to the main safety functions and their mechanisms have been specified, in order to better define the needed provisions. • On the basis of the above and taking into account the results of the safety analyses performed during previous project (ELSY), a reference configuration of the ELFR plant has been consolidated, by improving and updating the plant design features. In particular, the emerged safety concerns have been analyzed in the LEADER project and a new set of design options and safety provisions have been proposed. • The combination of favourable Lead coolant inherent characteristics and plant design features, specifically developed to face identified challenges, resulted in a very robust and forgiving design, even in very extreme conditions, as a Fukushima-like scenario

  15. The designing of launch vehicles with liquid propulsion engines ensuring fire, explosion and environmental safety requirements of worked-off stages

    Science.gov (United States)

    Trushlyakov, V.; Shatrov, Ya.; Sujmenbaev, B.; Baranov, D.

    2017-02-01

    The paper addresses the problem of the launch vehicles (LV) with main liquid propulsion engines launch technogenic impact in different environment areas. Therefore, as the study subjects were chosen the worked-off stages (WS) with unused propellant residues in tanks, the cosmodrome ecological monitoring system, the worked-off stage design and construction solutions development system and the unified system with the "WS+the cosmodrome ecological monitoring system+design and construction solutions development system" feedback allowing to form the optimal ways of the WS design and construction parameters variations for its fire and explosion hazard management in different areas of the environment. It is demonstrated that the fire hazard effects of propellant residues in WS tanks increase the ecosystem disorder level for the Vostochny cosmodrome impact area ecosystem. Applying the system analysis, the proposals on the selection of technologies, schematic and WS design and construction solutions aimed to the fire and explosion safety improvement during the LV worked-off stages with the main liquid propulsion engines operation were formulated. Among them are the following: firstly, the unused propellant residues in tanks convective gasification based on the hot gas (heat carrier) supply in WS tanks after main liquid propulsion engines cutoff is proposed as the basic technology; secondly, the obtained unused propellant residues in WS tanks gasification products (evaporated propellant residues + pressurizing agent + heat carrier) are used for WS stabilization and orientation while descending trajectory moving. The applying of the proposed technologies allows providing fire and explosion safety requirements of LV with main liquid propulsion engines practically.

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

    OpenAIRE

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

    2015-01-01

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

  17. Nuclear data for fission reactor core design and safety analysis: Requirements and status of accuracy of nuclear data

    International Nuclear Information System (INIS)

    Rowlands, J.L.

    1984-01-01

    The types of nuclear data required for fission reactor design and safety analysis, and the ways in which the data are represented and approximated for use in reactor calculations, are summarised first. The relative importance of different items of nuclear data in the prediction of reactor parameters is described and ways of investigating the accuracy of these data by evaluating related integral measurements are discussed. The use of sensitivity analysis, together with estimates of the uncertainties in nuclear data and relevant integral measurements, in assessing the accuracy of prediction of reactor parameters is described. The inverse procedure for deciding nuclear data requirements from the target accuracies for prediction of reactor parameters follows on from this. The need for assessments of the uncertainties in nuclear data evaluations and the form of the uncertainty information is discussed. The status of the accuracies of predictions and nuclear data requirements are then summarised. The reactor parameters considered include: (a) Criticality conditions, conversion and burn-up effects. (b) Energy production and deposition, decay heating, irradiation damage, dosimetry and induced radioactivity. (c) Kinetics characteristics and control, including temperature, power and coolant density coefficients, delayed neutrons and control absorbers. (author)

  18. 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

  19. 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.

  20. Nuclear safety cooperation for Soviet designed reactors

    International Nuclear Information System (INIS)

    Reisman, A.W.; Horak, W.C.

    1995-01-01

    The nuclear accident at the Chernobyl nuclear power plant in 1986 first alerted the West to the significant safety risks of Soviet designed reactors. Five years later, this concern was reaffirmed when the IAEA, as a result of a review by an international team of nuclear safety experts, announced that it did not believe the Kozloduy nuclear power plants in Bulgaria could be operated safely. To address these safety concerns, the G-7 summit in Munich in July 1992 outlined a five point program to address the safety problems of Soviet Designed Reactors: operational safety improvement; near-term technical improvements to plants based on safety assessment; enhancing regulatory regimes; examination of the scope for replacing less safe plants by the development of alternative energy sources and the more efficient use of energy; and upgrading of the plants of more recent design. As of early 1994, over 20 countries and international organizations have pledged hundreds of millions of dollars in financial assistance to improve safety. This paper summarizes these assistance efforts for Soviet designed reactors, draws lessons learned from these activities, and offers some options for better addressing these concerns

  1. 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

  2. Safety Software Guide Perspectives for the Design of New Nuclear Facilities (U)

    International Nuclear Information System (INIS)

    VINCENT, Andrew

    2005-01-01

    In June of this year, the Department of Energy (DOE) issued directives DOE O 414.1C and DOE G 414.1-4 to improve quality assurance programs, processes, and procedures among its safety contractors. Specifically, guidance entitled, ''Safety Software Guide for use with 10 CFR 830 Subpart A, Quality Assurance Requirements, and DOE O 414.1C, Quality Assurance, DOE G 414.1-4'', provides information and acceptable methods to comply with safety software quality assurance (SQA) requirements. The guidance provides a roadmap for meeting DOE O 414.1C, ''Quality Assurance'', and the quality assurance program (QAP) requirements of Title 10 Code of Federal Regulations (CFR) 830, Subpart A, Quality Assurance, for DOE nuclear facilities and software application activities. [1, 2] The order and guide are part of a comprehensive implementation plan that addresses issues and concerns documented in Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1. [3] Safety SQA requirements for DOE as well as National Nuclear Security Administration contractors are necessary to implement effective quality assurance (QA) processes and achieve safe nuclear facility operations. DOE G 414.1-4 was developed to provide guidance on establishing and implementing effective QA processes tied specifically to nuclear facility safety software applications. The Guide includes software application practices covered by appropriate national and international consensus standards and various processes currently in use at DOE facilities. While the safety software guidance is considered to be of sufficient rigor and depth to ensure acceptable reliability of safety software at all DOE nuclear facilities, new nuclear facilities are well suited to take advantage of the guide to ensure compliant programs and processes are implemented. Attributes such as the facility life-cycle stage and the hazardous nature of each facility operations are considered, along with the category and level of importance of the

  3. A study on fire design accidental loads for aluminum safety helidecks

    Directory of Open Access Journals (Sweden)

    Sang Jin Kim

    2016-11-01

    Full Text Available The helideck structure must satisfy the safety requirements associated with various environmental and accidental loads. Especially, there have been a number of fire accidents offshore due to helicopter collision (take-off and/or landing in recent decades. To prevent further accidents, a substantial amount of effort has been directed toward the management of fire in the safety design of offshore helidecks. The aims of this study are to introduce and apply a procedure for quantitative risk assessment and management of fires by defining the fire loads with an applied example. The frequency of helicopter accidents are considered, and design accidental levels are applied. The proposed procedures for determining design fire loads can be efficiently applied in offshore helideck development projects.

  4. Review on JMTR safety design for LEU core conversion

    International Nuclear Information System (INIS)

    Komori, Yoshihiro; Yokokawa, Makoto; Saruta, Toru; Inada, Seiji; Sakurai, Fumio; Yamamoto, Katsumune; Oyamada, Rokuro; Saito, Minoru

    1993-12-01

    Safety of the JMTR was fully reviewed for the core conversion to low enriched uranium fuel. Fundamental policies for the JMTR safety design were reconsidered based on the examination guide for safety design of test and research reactors, and safety of the JMTR was confirmed. This report describes the safety design of the JMTR from the viewpoint of major functions for reactor safety. (author)

  5. 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

  6. 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)

  7. Ecological Design of Cooperative Human-Machine Interfaces for Safety of Intelligent Transport Systems

    Directory of Open Access Journals (Sweden)

    Orekhov Aleksandr

    2016-01-01

    Full Text Available The paper describes research results in the domain of cooperative intelligent transport systems. The requirements for human-machine interface considering safety issue of for intelligent transport systems (ITSare analyzed. Profiling of the requirements to cooperative human-machine interface (CHMI for such systems including requirements to usability and safety is based on a set of standards for ITSs. An approach and design technique of cooperative human-machine interface for ITSs are suggested. The architecture of cloud-based CHMI for intelligent transport systems has been developed. The prototype of software system CHMI4ITSis described.

  8. Safety performance of preliminary KALIMER conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

    1999-07-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  9. Safety performance of preliminary KALIMER conceptual design

    International Nuclear Information System (INIS)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

    1999-01-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  10. 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

  11. Design of safety mechanism for an industrial manipulator based on passive compliance

    International Nuclear Information System (INIS)

    Kim, Hwi Su; Park, Jung Jun; Song, Jae Bok; Kyung, Jin Ho

    2010-01-01

    In recent years, collision safety between humans and robots has drawn much attention since human-robot cooperation is increasingly needed in various fields. Since positioning accuracy and collision safety are both important, an industrial manipulator should maintain very high stiffness for positioning accuracy in a normal situation, but exhibit very low stiffness when subjected to a collision force greater than the tolerance for human injury. To satisfy these requirements, we proposed in our previous research a safety mechanism composed of a linear spring and a double-slider mechanism for a service robot with a small payload. We modified this device to meet more stringent requirements for an industrial manipulator which usually has a payload higher than a service robot. Several experiments on static and dynamic collisions showed high stiffness of the safety mechanism in response to an external torque that was less than a predetermined threshold torque, but low stiffness that enabled absorption of the collision force when the external torque exceeded the threshold. Thus, positioning accuracy and collision safety were improved using the proposed design. Furthermore, a new safety criterion is suggested to verify the collision safety of a manipulator that uses the proposed safety mechanism

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  13. Basic national requirements for safe design, construction and operation

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1980-01-01

    Nuclear power plants have to be save. Vendors and utilities operating such plants, are convinced that their plants meet this requirement. Who, however, is establishing the safety requirements to be met by those manufacturing and operating nuclear power plants. What are the mechanisms to control whether the features provided assure the required safety level. Who controls whether the required and planned safety features are really provided. Who is eventually responsible for assuring safety after commissioning of a nuclear power plant. These fundamental questions being raised in many discussions on safety and environmental protection are dealt with in the following sections: (1) Fundamental safety requirements on nuclear power plants, in which such items as risk, legal bases and licensing procedure are discussed, (2) Surveillance during construction, in which safety analysis report, siting, safety evaluation, document examination, quality assurance, and commissioning testing are dealt with, (3) Operating tests and conditions in which recurrent inspections, environmental protection during operation, investigation of abnormal occurences and backfitting requirements as reviewed, and (4) Safety philosophy and safety policy to conclude this presentation. The German approach to nuclear safety serves as an example for an effective way of assuring safe nuclear power. (orig.)

  14. Safety issues relating to the design of fusion power facilities

    International Nuclear Information System (INIS)

    Stasko, R.R.; Wong, K.Y.; Russell, S.B.

    1986-06-01

    In order to make fusion power a viable future source of energy, it will be necessary to ensure that the cost of power for fusion electric generation is competitive with advanced fission concepts. In addition, fusion power will have to live up to its original promise of being a more radiologically benign technology than fission, and be able to demonstrate excellent operational safety performance. These two requirements are interrelated, since the selection of an appropriate safety philosophy early in the design phase could greatly reduce or eliminate the capital costs of elaborate safety related and protective sytems. This paper will briefly overview a few of the key safety issues presently recognized as critical to the ultimate achievement of licensable, environmentally safe and socially acceptable fusion power facilities. 12 refs

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

    International Nuclear Information System (INIS)

    Saito, Takehiko

    2004-01-01

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

  16. 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.

  17. Improved safety features in the design of Alto Lazio NPP

    International Nuclear Information System (INIS)

    Bava, G.; Cianciolo, T.; Del Nero, G.

    1988-01-01

    The ALTO LAZIO Nuclear Power Plant, two 1000Mwe units, is a BWR 6/MARK III located about 100 km north of Rome, on the Tyrrhenian Sea Coasts. The construction of the plant started in 1978, but it has recently been stopped by a Government decision following a national referendum, when the units were about 70% completed. This paper is mainly intended to illustrate the major safety features which have been implemented as result of specific requirements issued by the safety authority (ENEA DISP) during the construction permit stage or the subsequent licensing process. One of the tools used to identify the need for design modifications has been a comprehensive reliability analysis of safety system: in the paper the methods used and the major results obtained by this study are briefly presented. Also, the approach used in the investigation of severe accidents and major applications in the area of plant design and emergency procedures are briefly discussed; furthermore the trend toward a simpler mitigation concept is described

  18. Design of Hack-Resistant Diabetes Devices and Disclosure of Their Cyber Safety.

    Science.gov (United States)

    Sackner-Bernstein, Jonathan

    2017-03-01

    The focus of the medical device industry and regulatory bodies on cyber security parallels that in other industries, primarily on risk assessment and user education as well as the recognition and response to infiltration. However, transparency of the safety of marketed devices is lacking and developers are not embracing optimal design practices with new devices. Achieving cyber safe diabetes devices: To improve understanding of cyber safety by clinicians and patients, and inform decision making on use practices of medical devices requires disclosure by device manufacturers of the results of their cyber security testing. Furthermore, developers should immediately shift their design processes to deliver better cyber safety, exemplified by use of state of the art encryption, secure operating systems, and memory protections from malware.

  19. 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

  20. 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)

  1. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 2-Domino Hazard Index and case study.

    Science.gov (United States)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-12-15

    The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. 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...

  4. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-09-01

    The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

  5. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-01-01

    This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

  6. Future CANDU nuclear power plant design requirements document executive summary

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; S. A. Usmani

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new

  7. Enhanced CANDU 6 design assist probabilistic safety assessment results and insights

    International Nuclear Information System (INIS)

    Torabi, T.; Bettig, R.; Iliescu, P.; Robinson, J.; Santamaura, P.; Skorupska, B.; Tyagi, A.K.; Vencel, I.

    2013-01-01

    The Enhanced CANDU 6(EC6) is a 700 MWe reactor, which has evolved from the well-established CANDU line of reactors, which are heavy-water moderated, and heavy-water cooled horizontal pressure tube reactors, using natural uranium fuel. The EC6 design retains the generic CANDU design features, while incorporating innovations and state-of-the-art technologies to ensure competitiveness with other design with respect to operation, performance and economics. A design assist probabilistic safety assessment (PSA) was conducted during the design change phase of the project. The purpose of the assessment was to assess internal events during at-power operation and identify the design improvements and additional features needed to comply with the latest regulatory requirements in Canada and compete with other reactor designs, internationally. The PSA results show that the EC6 plant response to the postulated initiating events is well balanced, and the design meets its safety objectives. This paper summarizes the results and insights gained during the development of the PSA models for at-power internal events. (author)

  8. OSHA and Experimental Safety Design.

    Science.gov (United States)

    Sichak, Stephen, Jr.

    1983-01-01

    Suggests that a governmental agency, most likely Occupational Safety and Health Administration (OSHA) be considered in the safety design stage of any experiment. Focusing on OSHA's role, discusses such topics as occupational health hazards of toxic chemicals in laboratories, occupational exposure to benzene, and role/regulations of other agencies.…

  9. SEISMIC DESIGN REQUIREMENTS SELECTION METHODOLOGY FOR THE SLUDGE TREATMENT and M-91 SOLID WASTE PROCESSING FACILITIES PROJECTS

    International Nuclear Information System (INIS)

    RYAN GW

    2008-01-01

    In complying with direction from the U.S. Department of Energy (DOE), Richland Operations Office (RL) (07-KBC-0055, 'Direction Associated with Implementation of DOE-STD-1189 for the Sludge Treatment Project,' and 08-SED-0063, 'RL Action on the Safety Design Strategy (SDS) for Obtaining Additional Solid Waste Processing Capabilities (M-91 Project) and Use of Draft DOE-STD-I 189-YR'), it has been determined that the seismic design requirements currently in the Project Hanford Management Contract (PHMC) will be modified by DOE-STD-1189, Integration of Safety into the Design Process (March 2007 draft), for these two key PHMC projects. Seismic design requirements for other PHMC facilities and projects will remain unchanged. Considering the current early Critical Decision (CD) phases of both the Sludge Treatment Project (STP) and the Solid Waste Processing Facilities (M-91) Project and a strong intent to avoid potentially costly re-work of both engineering and nuclear safety analyses, this document describes how Fluor Hanford, Inc. (FH) will maintain compliance with the PHMC by considering both the current seismic standards referenced by DOE 0 420.1 B, Facility Safety, and draft DOE-STD-1189 (i.e., ASCE/SEI 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, and ANSI ANS 2.26-2004, Categorization of Nuclear Facility Structures, Systems and Components for Seismic Design, as modified by draft DOE-STD-1189) to choose the criteria that will result in the most conservative seismic design categorization and engineering design. Following the process described in this document will result in a conservative seismic design categorization and design products. This approach is expected to resolve discrepancies between the existing and new requirements and reduce the risk that project designs and analyses will require revision when the draft DOE-STD-1189 is finalized

  10. TIBER II/ETR final design report: Volume 3, 5.0 Radiation safety and environment; 6.0 Physics and technology R and D needs

    International Nuclear Information System (INIS)

    Lee, J.D.

    1987-09-01

    This paper discusses the design of the TIBER II Tokamak. This particular volume discusses: safety and environmental requirements and design targets; accident analyses; personnel safety and maintenance exposure; effluent control; waste management and decommissioning; safety considerations in building design; and safety and environmental conclusions and recommendations

  11. Integrating Safety, Operations, Security, and Safeguards (ISOSS) into the design of small modular reactors : a handbook.

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, Bobby D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mendez, Carmen Margarita [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2013-10-01

    The existing regulatory environment for nuclear reactors impacts both the facility design and the cost of operations once the facility is built. Delaying the consideration of regulatory requirements until late in the facility design - or worse, until after construction has begun - can result in costly retrofitting as well as increased operational costs to fulfill safety, security, safeguards, and emergency readiness requirements. Considering the scale and scope, as well as the latest design trends in the next generation of nuclear facilities, there is an opportunity to evaluate the regulatory requirements and optimize the design process for Small Modular Reactors (SMRs), as compared to current Light Water Reactors (LWRs). To this end, Sandia has embarked on an initiative to evaluate the interactions of regulations and operations as an approach to optimizing the design of SMR facilities, supporting operational efficiencies, as well as regulatory requirements. The early stages of this initiative consider two focus areas. The first focus area, reported by LaChance, et al. (2007), identifies the regulatory requirements established for the current fleet of LWR facilities regarding Safety, Security, Operations, Safeguards, and Emergency Planning, and evaluates the technical bases for these requirements. The second focus area, developed in this report, documents the foundations for an innovative approach that supports a design framework for SMR facilities that incorporates the regulatory environment, as well as the continued operation of the facility, into the early design stages, eliminating the need for costly retrofitting and additional operating personnel to fulfill regulatory requirements. The work considers a technique known as Integrated Safety, Operations, Security and Safeguards (ISOSS) (Darby, et al., 2007). In coordination with the best practices of industrial operations, the goal of this effort is to develop a design framework that outlines how ISOSS

  12. Design of the Control System for Engineered Safety Features of KIJANG Research Reactor

    International Nuclear Information System (INIS)

    Kim, Hagtae; Kim, Jun-Yeon; Chae, Hee-Taek

    2015-01-01

    The purpose of this paper is to design an effective control system for the Engineered Safety Features (ESF) of KJRR such as the Safety Residual Heat Removal System (SRHRS) pumps and Siphon Break Valve (SBV) without an Engineered Safety Features-Component Control System (ESF-CCS). This control system is called a 'local motor starter', because this system controls motors in the SRHRS pumps and SBVs by receiving the signal from Reactor Protection System (RPS) and Alternate Protection System (APS) when the differential pressure or pool level reach the set points. In this paper, the design concepts and requirements of the local motor starter based on the design features of KJRR is proposed. An ESF is a safety system that mitigates consequences of the Anticipated Operational Occurrence (AOO) and Design Basis Accident (DBA). The results of this paper are able to be used for the development of control systems for research reactors similar to KJRR. The precondition for such application is to have a few ESFs and conduct simple logic. The proposed control system called a local motor starter is being designed, and a manufacture of the actual systems is expected in the foreseeable future

  13. 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.)

  14. 1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1977-08-01

    This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations

  15. 1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

    Energy Technology Data Exchange (ETDEWEB)

    Blomeke, J.O.

    1977-08-01

    This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations. (DLC)

  16. Tornado-resistance design for the nuclear safety structure of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Xia Zufeng.

    1987-01-01

    The primary design consideration of anti-tornado of the nuclear safety structure of Qinshan Nuclear Power Plant is briefly presented. It mainly includes estimating the probability of tornado arising in the site, ascertaining the design requirments of the anti-tornado structures and deciding the tornado load acted on the structures

  17. 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

  18. Design of a Construction Safety Training System using Contextual Design Methodology

    OpenAIRE

    Baldev, Darshan H.

    2006-01-01

    In the U.S., the majority of construction companies are small companies with 10 or fewer employees (BLS, 2004). The fatality rate in the construction industry is high, indicating a need for implementing safety training to a greater extent. This research addresses two main goals: to make recommendations and design a safety training system for small construction companies, and to use Contextual Design to design the training system. Contextual Design was developed by Holtzblatt (Beyer and Holtzb...

  19. An architecture pattern for safety critical automated driving applications: Design and analysis

    NARCIS (Netherlands)

    Luo, Y.; Saberi, A.K.; Bijlsma, T.; Lukkien, J.J.; Brand, M. van den

    2017-01-01

    Introduction of automated driving increases complexity of automotive systems. As a result, architecture design becomes a major concern for ensuring non-functional requirements such as safety, and modifiability. In the ISO 26262 standard, architecture patterns are recommended for system development.

  20. An architecture pattern for safety critical automated driving applications : design and analysis

    NARCIS (Netherlands)

    Luo, Y.; Khabbaz Saberi, A.; Bijlsma, T.; Lukkien, J.J.; van den Brand, M.G.J.

    2017-01-01

    Introduction of automated driving increases complexity of automotive systems. As a result, architecture design becomes a major concern for ensuring non-functional requirements such as safety, and modifiability. In the ISO 26262 standard, architecture patterns are recommended for system development.

  1. 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

  2. Small Column Ion Exchange Design and Safety Strategy

    International Nuclear Information System (INIS)

    Huff, T.; Rios-Armstrong, M.; Edwards, R.; Herman, D.

    2011-01-01

    Small Column Ion Exchange (SCIX) is a transformational technology originally developed by the Department of Energy (DOE) Environmental Management (EM-30) office and is now being deployed at the Savannah River Site (SRS) to significantly increase overall salt processing capacity and accelerate the Liquid Waste System life-cycle. The process combines strontium and actinide removal using Monosodium Titanate (MST), Rotary Microfiltration, and cesium removal using Crystalline Silicotitanate (CST, specifically UOP IONSIV(reg s ign)IE-911 ion exchanger) to create a low level waste stream to be disposed in grout and a high level waste stream to be vitrified. The process also includes preparation of the streams for disposal, e.g., grinding of the loaded CST material. These waste processing components are technically mature and flowsheet integration studies are being performed including glass formulations studies, application specific thermal modeling, and mixing studies. The deployment program includes design and fabrication of the Rotary Microfilter (RMF) assembly, ion-exchange columns (IXCs), and grinder module, utilizing an integrated system safety design approach. The design concept is to install the process inside an existing waste tank, Tank 41H. The process consists of a feed pump with a set of four RMFs, two IXCs, a media grinder, three Submersible Mixer Pumps (SMPs), and all supporting infrastructure including media receipt and preparation facilities. The design addresses MST mixing to achieve the required strontium and actinide removal and to prevent future retrieval problems. CST achieves very high cesium loadings (up to 1,100 curies per gallon (Ci/gal) bed volume). The design addresses the hazards associated with this material including heat management (in column and in-tank), as detailed in the thermal modeling. The CST must be size reduced for compatibility with downstream processes. The design addresses material transport into and out of the grinder and

  3. Test design requirements: Canister-scale heater test

    International Nuclear Information System (INIS)

    Schauer, M.I.; Craig, P.A.; Stickney, R.G.

    1986-03-01

    This document establishes the Test Design Requirements for the design of a canister scale heater test to be performed in the Exploratory Shaft test facility. The purpose of the test is to obtain thermomechanical rock mass response data for use in validation of the numerical models. The canister scale heater test is a full scale simulation of a high-level nuclear waste container in a prototypic emplacement borehole. Electric heaters are used to simulate the heat loads expected in an actual waste container. This document presents an overview of the test including objectives and justification for the test. A description of the test as it is presently envisioned is included. Discussions on Quality Assurance and Safety are also included in the document. 12 refs., 1 fig

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

    International Nuclear Information System (INIS)

    Meissner, Frank

    2008-01-01

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

  5. The IRSN experience feedback for the transport package design safety appraisals

    International Nuclear Information System (INIS)

    Sert, G.

    2007-01-01

    The activity of transportation of radioactive materials is in constant evolution; air transport of radio elements for medical use is growing rapidly as well as transport of instruments equipped with radioactive sources for inspections of buildings (controls of presence of lead in paintings) and in industry (non destructive examination of welding by gammagraphy, controls of density on building sites). Transports associated with the recycling of plutonium for the production of electricity by nuclear energy are now accomplished in routine. Globally, 900.000 packages are shipped each year in France; among them, approximately 100.000 packages belong to the category for which design approval is required. To maintain a high level of safety for this activity by limiting the probability of occurrence, the severity and consequences of the incidents and accidents, strict rules are implemented by users under the control of the Safety Authority According to the systematic approach of defence in depth, which is defined by the three principles of safety in design, of operational reliability and of effectiveness of emergency response, the robustness of the design of the package is of primary importance. It is based on regulatory requirements relating to the functions of safety (containment of radioactivity, protection against radiation and prevention of the risks of criticality) that must be ensured by the package in conditions of transport as well as in accident conditions. These rules and the way of applying them evolve with time. Indeed, on the one hand the regulation is reexamined periodically; on the other hand, the technical knowledge on the behaviour of the packages subject to the above mentioned conditions and the means of evaluation of this behaviour progress permanently

  6. 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

  7. Design and qualification of HPD based designs for safety systems

    International Nuclear Information System (INIS)

    Sharma, Mukesh Kr.; Chavan, Madhavi A.; Sawhney, Pratibha A.; Mohanty, Ashutos; John, Ajith K.; Ganesh, G.

    2014-01-01

    Field Programmable Gate Arrays (FPGA) and Complex Programmable Logic Devices (CPLD) are increasingly being used in C and I system of NPPs. The function of such an integrated circuit is not defined by the supplier of the physical component or micro-electronic technology but by the C and I designer. The hardware subsystems implemented in these devices typically use Hardware Description Language (HDL) like VHDL or Verilog to describe the functionality at the design entry level. These circuits are commonly known as 'HDL-Programmed Devices', (HPD). RCnD has developed a set of hardware boards to be used in next generation C and I systems. The boards have been designed based on present day technology and components. The intelligence of these boards has been implemented in HPDs (FPGA/CPLD) using VHDL. Since these boards are used in the safety and safety related systems, they have undergone a rigorous V and V process and qualification tests. This paper discusses the design attributes and qualification of these HPD based designs for nuclear class safety systems. (author)

  8. 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

  9. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    International Nuclear Information System (INIS)

    Tyagi, Himanshu; Soni, Jignesh; Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli; Gahlaut, Agrajit; Joshi, Jaydeep; Parmar, Deepak; Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun

    2016-01-01

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

  10. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    Energy Technology Data Exchange (ETDEWEB)

    Tyagi, Himanshu, E-mail: htyagi@iter-india.org [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Soni, Jignesh [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Gahlaut, Agrajit [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Joshi, Jaydeep; Parmar, Deepak [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India)

    2016-11-15

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

  11. 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.

  12. 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

  13. Safety design features of the IRIS

    International Nuclear Information System (INIS)

    2009-01-01

    The International Reactor Innovative and Secure (IRIS) is an advanced, integral, light water cooled reactor of medium generating capacity (335 MW(e)), that features an integral reactor vessel containing all the reactor primary system components, including steam generators, coolant pumps, pressurizer and heaters, and control rod drive mechanisms; in addition to the typical core, internals, control rods and neutron reflector. This integral configuration allows for the use of a small, high design pressure, spherical steel containment which results in a significant reduction in the size of the nuclear island. Other IRIS innovations include a simplified passive safety system concept and equipment features that derive from the 'safety-by-design' philosophy. This design approach allows for elimination of certain accident initiators at the design stage, or when outright elimination is not possible, decreases accident consequences and/or their probability of occurrence. Major design characteristics of the IRIS are given. As part of the IRIS pre-application licensing review by the U.S. Nuclear Regulatory Commission (NRC), the IRIS design team has developed a test plan that will provide the necessary data for safety analysis computer model verification, as well as for verifying the manufacturing feasibility, operability, and durability of new component designs

  14. 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

  15. Safety aspects on the Asea-Atom BWR 75 control room design

    International Nuclear Information System (INIS)

    Gemst, Paul van; Pedersen, Tor.

    1978-01-01

    The control room is an integrated part of the total plant layout and is located in a special building, known as the control building. The problems of designing a control room meeting all safety requirements and at the same time allowing for modifications to meet special customer specifications are described. (author)

  16. Design aspects of radiological safety in nuclear facilities

    International Nuclear Information System (INIS)

    Patkulkar, D.S.; Purohit, R.G.; Tripathi, R.M.

    2014-01-01

    In order to keep operational performance of a nuclear facility high and to keep occupational and public exposure ALARA, radiological safety provisions must be reviewed at the time of facility design. Deficiency in design culminates in deteriorated system performance and non adherence to safety standards and could sometimes result in radiological incident. Important radiological aspects relevant to safety were compiled based on operating experiences, design deficiencies brought out from past nuclear incidents, experience gained during maintenance, participation in design review of upcoming nuclear facilities and radiological emergency preparedness

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

    Science.gov (United States)

    Darpel, Scott; Beckman, Sean

    2016-01-01

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

  18. Regulatory requirements on the design and construction of nuclear power plant control and instrumentation systems in Finland

    International Nuclear Information System (INIS)

    Heikkila, M.A.

    1978-01-01

    The Department of Reactor Safety of the Institute of Radiation Protection, being the nuclear regulatory authority in Finland, has set up regulations which govern the design and construction of NPP systems and components. The regulations are partly compiled from existing codes and standards, published primarily in the United States and Federal Republic of Germany, and partly worked out at the Institute. The regulations are collected to a special set of YVL guides (guides for nuclear power plants), and one of these gives requirements on the design and construction of NPPCI systems and components. The scope of the requirements is based on the safety classification of the CI systems and components. Three safety classes have been singled out: the first for CI systems which take part in reactor protection, the second for other directly safety related, and the third for remaining CI systems important enough to deserve supervision. The safety class for CI components is inherited from the system they belong to. The safety classification of IC systems has direct bearing on the initial assumptions of plant accident analysis. The design principles of IC systems are inspected as part of the preliminary and final safety reports. Focus is directed on the principles of redundancy, separation, diversity, testability, etc. The requirements on IC components are directed to different stages of manufacture, installation and operation. The type tests shall be adequate and acceptably documented. The manufacture of components is followed, the test reports reviewed and the efficiency of manufacturers quality assurance program evaluated. Further requirements concern the installation phase and tests at the end of it, and finally guides include directions for maintenance and testing during the operations phase. (author)

  19. 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

  20. Design and implementation of an identification system in construction site safety for proactive accident prevention.

    Science.gov (United States)

    Yang, Huanjia; Chew, David A S; Wu, Weiwei; Zhou, Zhipeng; Li, Qiming

    2012-09-01

    Identifying accident precursors using real-time identity information has great potential to improve safety performance in construction industry, which is still suffering from day to day records of accident fatality and injury. Based on the requirements analysis for identifying precursor and the discussion of enabling technology solutions for acquiring and sharing real-time automatic identification information on construction site, this paper proposes an identification system design for proactive accident prevention to improve construction site safety. Firstly, a case study is conducted to analyze the automatic identification requirements for identifying accident precursors in construction site. Results show that it mainly consists of three aspects, namely access control, training and inspection information and operation authority. The system is then designed to fulfill these requirements based on ZigBee enabled wireless sensor network (WSN), radio frequency identification (RFID) technology and an integrated ZigBee RFID sensor network structure. At the same time, an information database is also designed and implemented, which includes 15 tables, 54 queries and several reports and forms. In the end, a demonstration system based on the proposed system design is developed as a proof of concept prototype. The contributions of this study include the requirement analysis and technical design of a real-time identity information tracking solution for proactive accident prevention on construction sites. The technical solution proposed in this paper has a significant importance in improving safety performance on construction sites. Moreover, this study can serve as a reference design for future system integrations where more functions, such as environment monitoring and location tracking, can be added. Copyright © 2011 Elsevier Ltd. All rights reserved.

  1. Relevant safety issues in designing the HTR-10 reactor

    International Nuclear Information System (INIS)

    Sun Yuliang; Xu Yuanghui

    2001-01-01

    The HTR-10 is a 10 MWth pebble bed high temperature gas cooled reactor being constructed as a research facility at the Institute of Nuclear Energy Technology. This paper discusses design issues of the HTR-10 which are related to safety. It addresses the safety criteria used in the development and assessment of the design, the safety important systems, and the safety classification of components. It also summarises the results of safety analysis, including the approach used for the radioactive source term, as well as the approach to containment design. (author)

  2. 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

  3. Environmental, health, and safety by design

    International Nuclear Information System (INIS)

    Soklow, R.G.

    1999-01-01

    Solar Turbines Incorporated created a self-directed work team, the Safety and Environmental Awareness (SEA) Team that initiated a company wide effort to raise employee awareness to promote integrating responsible environmental, health, and safety practices into product design, manufacturing, and services. Environmental, health, and safety issues influence how all businesses operate around the world. Companies choose to operate in an environmentally responsible manner because it not only benefits employees and the communities where they live, it also benefits the business when superior performance results in a competitive advantage. Solar surveyed gas turbines users to identify their top environmental and safety concerns and issues. The authors asked about various environmental and safety aspects of their equipment. Results from the survey has helped engineering and design focus efforts so that future products and product improvements assist customers in meeting their regulatory obligations and social responsibilities. Air pollution has historically been one of the most important environmental issues facing customers, because pollutant emissions greatly influence equipment choices and operation flexibility. There are other environmental, health and safety issues: sustainable fire suppression choices, start systems, hazardous materials use and ability to recycle materials, package accessibility, noise and product take back issues

  4. Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

    International Nuclear Information System (INIS)

    Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

    2011-01-01

    This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

  5. 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

  6. Use of a probabilistic safety study in the design of the Italian reference PWR

    International Nuclear Information System (INIS)

    Richardson, D.C.; Russino, G.; Valentini, V.

    1985-01-01

    The intent of this paper is to provide a description of the experience gained in having performed a Probabilistic Safety Study (PSS) on the proposed Italian reference pressurized water reactor. The experience revealed that through careful application of probabilistic techniques, Probabilistic Risk Assessment (PRA) can be used as a tool to develop an optimum plant design in terms of safety and cost. Furthermore, the PSS can also be maintained as a living document and a tool to assess additional regulatory requirements that may be imposed during the construction and operational life of the plant. Through the use of flexible probabilistic techniques, the probabilistic safety model can provide a living safety assessment starting from the conceptual design and continuing through the construction, testing and operational phases. Moreover, the probabilistic safety model can be used during the operational phase of the plant as a method to evaluate the operational experience and identify potential problems before they occur. The experience, overall, provided additional insights into the various aspects of the plants design and operation that would not have been identified through the use of traditional safety evaluation techniques

  7. 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.

  8. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This publication supports the Safety Requirements on the Safety of Nuclear Power Plants: Design. This Safety Guide was prepared on the basis of a systematic review of all the relevant publications including the Safety Fundamentals, Safety of Nuclear Power Plants: Design, current and ongoing revisions of other Safety Guides, INSAG reports and other publications that have addressed the safety of nuclear power plants. This Safety Guide also provides guidance for Contracting Parties to the Convention on Nuclear Safety in meeting their obligations under Article 14 on Assessment and Verification of Safety. The Safety Requirements publication entitled Safety of Nuclear Power Plants: Design states that a comprehensive safety assessment and an independent verification of the safety assessment shall be carried out before the design is submitted to the regulatory body. This publication provides guidance on how this requirement should be met. This Safety Guide provides recommendations to designers for carrying out a safety assessment during the initial design process and design modifications, as well as to the operating organization in carrying out independent verification of the safety assessment of new nuclear power plants with a new or already existing design. The recommendations for performing a safety assessment are suitable also as guidance for the safety review of an existing plant. The objective of reviewing existing plants against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The methods and the recommendations of this Safety Guide can also be used by regulatory bodies for the conduct of the regulatory review and assessment. Although most recommendations of this Safety Guide are general and applicable to all types of nuclear reactors, some specific recommendations and examples apply mostly to water cooled reactors. Terms such as 'safety assessment', 'safety analysis' and 'independent

  9. Review on conformance of JMTR reactor facility to safety design examination guides for water-cooled reactors for test and research

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Naka, Michihiro; Sakuta, Yoshiyuki; Hori, Naohiko; Matsui, Yoshinori; Miyazawa, Masataka

    2009-03-01

    The safety design examination guides for water-cooled reactors for test and research are formulated as fundamental judgements on the basic design validity for licensing from a viewpoint of the safety. Taking the refurbishment opportunity of the JMTR, the conformance of the JMTR reactor facility to current safety design examination guides was reviewed with licensing documents, annexes and related documents. As a result, it was found that licensing documents fully satisfied the requirements of the current guides. Moreover, it was found that the JMTR reactor facility itself also satisfied the guides requirements as well as the safety performance, since the facility with safety function such as structure, systems, devices had been installed based on the licensing documents under the permission by the regulation authority. Important devices for safety have been produced under authorization of regulating authority. Therefore, it was confirmed that the licensing was conformed to guides, and that the JMTR has enough performance. (author)

  10. 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.

  11. Design of control and safety rod and its drive mechanism of PFBR

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Govindarajan, S.; Chetal, S.C.

    1997-01-01

    Control and Safety Rod (CSR) is one of the two types of absorber rods in shutdown systems of PFBR. Control and Safety Rod Drive Mechanism (CSRDM) actuates CSR to have vertical translatory motion in reactor core. The dual responsibilities entrusted on CSR to control reactor power during normal operating condition and to shutdown the reactor by scram action during abnormal condition, necessitate highly reliable design, analysis, testing and surveillance of CSR and CSRDM. The paper discusses on the salient features of CSR and CSRDM and design and analysis of individual sub-assemblies, viz., gripper, scram-release electromagnet, hydraulic dash pot, seals. Also it discusses on the developmental activities proposed and surveillance test requirements. (author)

  12. Instrumentation and control systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. It supplements Safety Standards Series No. NS-R-1: Safety of Nuclear Power Plants: Design (the Requirements for Design), which establishes the design requirements for ensuring the safety of nuclear power plants. This Safety Guide describes how the requirements should be met for instrumentation and control (I and C) systems important to safety. This publication is a revision and combination of two previous Safety Guides: Safety Series Nos 50-SG-D3 and 50-SG-D8, which are superseded by this new Safety Guide. The revision takes account of developments in I and C systems important to safety since the earlier Safety Guides were published in 1980 and 1984, respectively. The objective of this Safety Guide is to provide guidance on the design of I and C systems important to safety in nuclear power plants, including all I and C components, from the sensors allocated to the mechanical systems to the actuated equipment, operator interfaces and auxiliary equipment. This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety. It expands on paragraphs of Ref in the area of I and C systems important to safety. This publication is intended for use primarily by designers of nuclear power plants and also by owners and/or operators and regulators of nuclear power plants. This Safety Guide provides general guidance on I and C systems important to safety which is broadly applicable to many nuclear power plants. More detailed requirements and limitations for safe operation specific to a particular plant type should be established as part of the design process. The present guidance is focused on the design principles for systems important to safety that warrant particular attention, and should be applied to both the design of new I and C systems and the modernization of existing systems. Guidance is provided on how design

  13. 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

  14. Utility requirements for safety in the passive advanced light-water reactor

    International Nuclear Information System (INIS)

    Marston, T.U.; Layman, W.H.; Bockhold, G. Jr.

    1993-01-01

    The objective of the passive plant design is to use passive systems to replace all the active engineered safety systems presently used in light-water reactors. The benefits derived from such an approach to safety design are multiple. First, it is expected that a passive design approach will significantly simplify the overall plant design, including a reduction in the number of components, and reduce the operation and maintenance burden. Second, it is expected that the overall safety and reliability of the passive systems will be improved over active systems, which will result in extremely low risk to public health and safety. Third, challenges to the operating staff will be minimized during transient and emergency conditions, which will reduce the uncertainty associated with human behavior. Finally, it is expected that reliance on passive safety features will lead to a better understanding by the general public and recognition that a major improvement in public safety has been achieved

  15. Safety and design limits

    International Nuclear Information System (INIS)

    Shishkov, L. K.; Gorbaev, V. A.; Tsyganov, S. V.

    2007-01-01

    The paper touches upon the issues of NPP safety ensuring at the stage of fuel load design and operation by applying special limitations for a series of parameters, that is, design limits. Two following approaches are compared: the one used by west specialists for the PWR reactor and the Russian approach employed for the WWER reactor. The closeness of approaches is established, differences that are mainly peculiarities of terms are noted (Authors)

  16. The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

    International Nuclear Information System (INIS)

    Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

    2006-06-01

    Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

  17. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

  18. 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.

  19. 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

  20. The current CEA/DRN safety approach for the design and the assessment of future nuclear installations

    International Nuclear Information System (INIS)

    Fiorini, G.L.; Pinto, P.L.; Costa, M.

    1999-01-01

    The purpose of the document is to present the basis of the safety approach currently implemented by the CEA/DRN, both for the design and the assessment of innovative systems and future nuclear installations. This approach is the result of the experience maturated, within the context of the CEA/DRN Innovative Programme through practical applications over several future concepts, both for fission and fusion reactors, as well as for waste disposal. The background of this experience is structured coherently with the European Safety Authorities recommendations and the European Utilities Requirements (EUR). The Defence In Depth principle and its application, by means, among others, of the barrier concept, remains the basis of the safety design process of future nuclear installations. Its adequacy is checked through the safety assessment. The methodology for Lines Of Defence (LOD) implementation as well as the one for the LOD architecture assessment is shown and motivated. The document shows that the clear and unambiguous definition of the safety approach provides an essential base for the organisation of the design tasks, being sure that the safety aspects are correctly taken into account and implemented, and for an adequate safety assessment of the final design, both from qualitative point of view as well as for the quantitative safety analysis. (author)

  1. Designing and Developing an Effective Safety Program for a Student Project Team

    Directory of Open Access Journals (Sweden)

    John Catton

    2018-05-01

    Full Text Available In the workplace, safety must be the first priority of all employers and employees alike. In order to maintain the safety and well-being of their employees, employers must demonstrate due diligence and provide the appropriate safety training to familiarize employees with the hazards within the workplace. Although, a student “project team” is not a business, the work done by students for their respective teams is synonymous with the work done in a place of business and thus requires that similar safety precautions and training be administered to students by their team leads and faculty advisors. They take on the role of supervisors within the team dynamic. Student teams often utilize the guidelines and policies that their universities or colleges have developed in order to build a set of standard operating procedures and safety training modules. These guidelines aid in providing a base for training for the team, however, they are no substitute for training specific to the safety risks associated with the work the team is doing. In order to comply with these requirements, a full analysis of the workplace is required to be completed. A variety of safety analysis techniques need to be applied to define the hazards within the workplace and institute appropriate measures to mitigate them. In this work, a process is developed for establishing a safety training program for a student project team, utilizing systems safety management techniques and the aspect of gamification to produce incentives for students to continue developing their skills. Although, systems safety management is typically applied to the design of active safety components or systems, the techniques for identifying and mitigating hazards can be applied in the same fashion to the workplace. They allow one to analyze their workplace and determine the hazards their employees might encounter, assign appropriate hazard ratings and segregate each respective hazard by their risks. In so

  2. Robot-Assisted Fracture Surgery: Surgical Requirements and System Design.

    Science.gov (United States)

    Georgilas, Ioannis; Dagnino, Giulio; Tarassoli, Payam; Atkins, Roger; Dogramadzi, Sanja

    2018-03-09

    The design of medical devices is a complex and crucial process to ensure patient safety. It has been shown that improperly designed devices lead to errors and associated accidents and costs. A key element for a successful design is incorporating the views of the primary and secondary stakeholders early in the development process. They provide insights into current practice and point out specific issues with the current processes and equipment in use. This work presents how information from a user-study conducted in the early stages of the RAFS (Robot Assisted Fracture Surgery) project informed the subsequent development and testing of the system. The user needs were captured using qualitative methods and converted to operational, functional, and non-functional requirements based on the methods derived from product design and development. This work presents how the requirements inform a new workflow for intra-articular joint fracture reduction using a robotic system. It is also shown how the various elements of the system are developed to explicitly address one or more of the requirements identified, and how intermediate verification tests are conducted to ensure conformity. Finally, a validation test in the form of a cadaveric trial confirms the ability of the designed system to satisfy the aims set by the original research question and the needs of the users.

  3. Developing design premises for a KBS-3V repository based on results from the safety assessment - 16027

    International Nuclear Information System (INIS)

    Andersson, Johan; Hedin, Allan

    2009-01-01

    As a part of the planned license application for a final repository for spent nuclear fuel the Swedish Nuclear Fuel and Waste Management Co. (SKB), has developed design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel. The purpose is to provide requirements from a long term safety aspect, to form the basis for the development of the reference design of the repository and to justify that design. Design premises typically concern specification on what mechanical loads the barriers must withstand, restrictions on the composition of barrier materials or acceptance criteria for the various underground excavations. These design constraints, if all fulfilled by the actual design, should form a good basis for demonstrating repository safety. The justification for these design premises is derived from SKB's most recent safety assessment SR-Can complemented by a few additional analyses. Some of the design premises may be modified in future stages of SKB's program, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. (authors)

  4. Safety research needs for Russian-designed reactors

    International Nuclear Information System (INIS)

    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. This Support Group was endorsed by the CSNI. The Support Group, which is composed of senior experts on safety research from several OECD countries and from Russia, prepared this Report. The Group reviewed the safety research performed to support Russian-designed reactors and set down its views on future needs. The review concentrates on the following main topics: Thermal-Hydraulics/Plant Transients for VVERs; Integrity of Equipment and Structures for VVERs; Severe Accidents for VVERs; Operational Safety Issues; Thermal-Hydraulics/Plant Transients for RBMKs; Integrity of Equipment and Structures for RBMKs; Severe Accidents for RBMKs. (K.A.)

  5. Nuclear Electric flask design and safety case development during the last ten years

    International Nuclear Information System (INIS)

    Dougall, I.; Jones, D.K.

    1994-01-01

    The scope of Nuclear Electric's (NE) requirements for irradiated fuel transport and the relevant safety and regulatory compliance standards are outlined. To illustrate NE's approach to demonstrating package design compliance the basis of the safety case for unbottled Magnox fuel transport in the Mk M2 Magnox flask is described. The considerations which lay behind the development of the Mk A2 AGR Flask are indicated and a description is given of the main design features of this flask. Finally there is a summary of progress in obtaining Type B(M) approvals, based on the 1985 IAEA Regulations, for the requisite range of flask contents. (author)

  6. 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

  7. 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...

  8. 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

  9. 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)

  10. PSA in design of passive/active safety reactors

    International Nuclear Information System (INIS)

    Sato, T.; Tanabe, A.; Kondo, S.

    1995-01-01

    PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

  11. Safety design integrated in the building delivery system

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2013-01-01

    . The purpose of this article is to demonstrate how safety and health can be integrated in the design phases integrated in the management delivery systems within construction, The method for the research was to go through the building delivery system step by step and create a normative description of what, when......In construction, it is important to view safety and health as an integrated part of the way that “designers” are working. The designers cowers architects, constructors, engineers and others who carry out their consulting services in the design phase of a construction project. The philosophy...... and how to fully integrate safety in each part of the process. The result is a concept and guideline including control forms for how to integrate safety design in the Building Delivery System plus what to do and when. The concept has been tested in an educational context. The practical value...

  12. SGHWR safety design and evaluation

    International Nuclear Information System (INIS)

    Smith, D.R.; Merrett, D.J.; Ward, D.A.

    1977-01-01

    The paper discusses the characteristic features of the S.G.H.W.R. and identifies the single channel concept as of considerable importance. The unique feature of the design is the provision of individual spray cooling E.C.C.S. to each channel. This spray cooling occupies a prominent position in the main line safety arguments. The reliance on this form of spray cooling leads to provision of a comprehensive E.C.C.S. system of high reliability. Duplicate systems with diverse power and water sources cover the complete pressure range to give very high confidence that spray cooling is available in all major L.O.C.A.s. On the other hand hydraulic analysis of the blowdown phase demonstrates that significant convective flow is available as an alternative/supplementary cooling regime for most faults. The reactor shutdown mechanisms have also been duplicated and will be designed to high reliabilities to give surety of reactor trip in all credible faults. The comparative performance of the two systems is considered. Extent of diversity and redundancy in trip parameters is also discussed. A feature of channel concept is that the pipe sizes can be made relatively small thus restricting rates of blowdown, and the paper discusses effects of this upon long term cooling and flooding arguments. The quantities of pipework in the primary circuit introduce considerations of integrity and the paper goes on to list the measures introduced to improve segregation and protection of individual sections of the plant so that the extent of possible L.O.C.A.s is minimised. The achievement of high standards of reliability by use of in-service inspection is covered, with particular reference to the steam drums. The impact of these inspection requirements upon the very low man-rem exposures required by U.K. utilities is also included. Finally, it is noted that the provision of containment in common with other L.W.R. practice also provides a valuable engineered safety feature. The principles of

  13. The reactor safety study of experimental multi-purpose VHTR design

    International Nuclear Information System (INIS)

    Yasuno, T.; Mitake, S.; Ezaki, M.; Suzuki, K.

    1981-01-01

    Over the past years, the design works of the Experimental Very High Temperature Reactor (VHTR) plant have been conducted at Japan Atomic Energy Research Institute. The conceptual design has been completed and the more detailed design works and the safety analysis of the experimental VHTR plant are continued. The purposes of design studies are to show the feasibility of the experimental VHTR program, to specify the characteristics and functions of the plant components, to point out the R and D items necessary for the experimental VHTR plant construction, and to analyze the feature of the plant safety. In this paper the summary of system design and safety features of the experimental reactor are indicated. Main issues are the safety philosophy for the design basis accident, the accidents assumed and the engineered safety systems adopted in the design works

  14. Design requirements, operation and maintenance of gas-cooled reactors

    International Nuclear Information System (INIS)

    1989-06-01

    At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

  15. LABORATORY DESIGN CONSIDERATIONS FOR SAFETY.

    Science.gov (United States)

    National Safety Council, Chicago, IL. Campus Safety Association.

    THIS SET OF CONSIDERATIONS HAS BEEN PREPARED TO PROVIDE PERSONS WORKING ON THE DESIGN OF NEW OR REMODELED LABORATORY FACILITIES WITH A SUITABLE REFERENCE GUIDE TO DESIGN SAFETY. THERE IS NO DISTINCTION BETWEEN TYPES OF LABORATORY AND THE EMPHASIS IS ON GIVING GUIDES AND ALTERNATIVES RATHER THAN DETAILED SPECIFICATIONS. AREAS COVERED INCLUDE--(1)…

  16. Design trade-offs in view of safety considerations

    International Nuclear Information System (INIS)

    Saji, G.; Kishida, K.; Inoue, T.

    1978-01-01

    In view of resolving conflicting demands of cost, safety, flexibility of operation and design margins, safety design of various plant systems is discussed referring to their weight on construction costs. An influence of hypothetical core disruptive accident (HCDA) and loss of piping integrity (LOPI) on plant design and thus on construction materials is discussed, in optimising future commercial FBR plants. (author)

  17. The enhancement of Ignalina NPP in design and operational safety

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1999-01-01

    Enhancement of Ignalina NPP design include: core design improvements; fuel channel integrity (multiple pressure tube rupture); improvements of shutdown systems; improvements of instrumentation and control devices; containment strength and tightness; design basis accident analysis; improvements of safety and support systems; seismic safety enhancement; Year 2000 project; cracks in pipes. Enhancement of operational safety includes: quality assurance; configuration management; safety management and safety culture; emergency operating procedures; training and full scope simulator; in-service inspection; fire protection and ageing monitoring and management

  18. 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.

  19. Status of conceptual safety design study of Japanese sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Kurisaka, Kenichi; Niwa, Hajime; Shimakawa, Yoshio

    2005-01-01

    In this paper, the current conceptual safety design and related evaluation of Japanese Sodium-cooled Fast Reactor which is studied in the framework of the Feasibility Study (FS) on commercialized Fast Reactor Cycle Systems in Japan are described. The purpose of the safety design is to establish a feasible safety concept of FBR which aims at a sustainable energy source of the next generations. The safety targets and the safety design principle are set aiming at realizing worldwide acceptability of the safety level. The basic safety design concept, which can meet the safety targets, was formulated taking along with the defense-in-depth philosophy as the basic safety design principle. In order to cope with wide range of energy and resource demands, there are some various designs both of oxide and metal fuel for JSFR. Some analytical results of typical design basis events, design extension conditions and core damage frequency estimation show the feasibility of the safety design concept for them. (author)

  20. Operating experience and systems analysis at Trillo NPP: A program intended for systematic review of plant safety systems to assess design basis requirements compliance

    International Nuclear Information System (INIS)

    Vega, R. de la

    1996-01-01

    The program was defined to apply to all plant safety systems and/or systems included in plant Technical Specifications. The goal of the program was to ensure, by systematic design, construction, and commissioning review, the adequacy of safety systems, structures and components to fulfill their safety functions. Also, as a result of the program, it was established that a complete, unambiguous, systematic, design basis definition shall take place. And finally, a complete documental review of the plant design shall result from the program execution

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

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

    Japan Atomic Energy Research Institute has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of a depressurization accident. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. (author)

  2. Design an optimum safety policy for personnel safety management - A system dynamic approach

    International Nuclear Information System (INIS)

    Balaji, P.

    2014-01-01

    Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making

  3. Design an optimum safety policy for personnel safety management - A system dynamic approach

    Energy Technology Data Exchange (ETDEWEB)

    Balaji, P. [The Glocal University, Mirzapur Pole, Delhi- Yamuntori Highway, Saharanpur 2470001 (India)

    2014-10-06

    Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.

  4. Design an optimum safety policy for personnel safety management - A system dynamic approach

    Science.gov (United States)

    Balaji, P.

    2014-10-01

    Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.

  5. 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)

  6. Review of the Safety Design Approaches in Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum

    2009-12-01

    The principle of the Defense in depth is essential in securing the safety of nuclear power plants, that is, to prevent cores-damaging severs accidents and to minimize the radiological consequences of the accidents 'as low as possible' (ALARA). One of the major design features of sodium fast reactors (SFRs) is that it has a large amount of sodium in the reactor vessel, providing a large heat capacity, such that it is feasible to contain the consequences of sever core damaging accidents in the vessel and primary system boundary. Containment of a severe accident in the primary system boundary, that is called in-vessel retention(IVR), is not a licensing requirement but set up as a design goal in most of the SFR design in the context of risk minimization. The objective of this report is to broadly review and compare the approaches and efforts made in the some of the major SFR designs of the US, Europe and Japan to prevent severe accidents and mitigate their consequences should they occur. Specifically, the subjects described in this report include design criteria or requirements, accident categorization and acceptance criteria, design features to prevent and contain severs accidents

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

    International Nuclear Information System (INIS)

    2013-01-01

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

  8. SAFETY IN THE DESIGN OF SCIENCE LABORATORIES AND BUILDING CODES.

    Science.gov (United States)

    HOROWITZ, HAROLD

    THE DESIGN OF COLLEGE AND UNIVERSITY BUILDINGS USED FOR SCIENTIFIC RESEARCH AND EDUCATION IS DISCUSSED IN TERMS OF LABORATORY SAFETY AND BUILDING CODES AND REGULATIONS. MAJOR TOPIC AREAS ARE--(1) SAFETY RELATED DESIGN FEATURES OF SCIENCE LABORATORIES, (2) LABORATORY SAFETY AND BUILDING CODES, AND (3) EVIDENCE OF UNSAFE DESIGN. EXAMPLES EMPHASIZE…

  9. The effect of regulatory requirements on the control and instrumentation system designer

    International Nuclear Information System (INIS)

    Golder, J.A.

    1978-01-01

    The difficulties encountered by the designer of control and protection systems for nuclear plant in attempting to satisfy the large number of imprecise regulations and recommendations which exist are described. The absence of fundamental quantitative safety requirements of international acceptability is deplored and the adoption of a major incident criteria expressed in quantitative terms as the basis for the derivation of target design criteria for protection systems and plant components is suggested. (author)

  10. 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)

  11. 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).

  12. Operational safety and radioprotection considerations when designing the ILW-LL disposal zone

    International Nuclear Information System (INIS)

    Voinis, S.; Roulet, A.; Claudel, D.; Lesavre, A.

    2008-01-01

    As for any other nuclear industrial facility, in a radioactive waste repository the various waste disposal operational activities from construction to closure can present a risk to human (workers and public) and the environment. In accordance with the December 30, 1991 French Waste Act, Andra has conducted feasibility studies regarding the disposal of HLW and ILW-LL waste in a clay host formation. The 'Dossier 2005 - Clay' includes a description of the operational safety analysis that was conducted for ILW-LL waste disposal in underground horizontal drifts. The objective of this paper is to present that safety analysis and its impact on the design at the feasibility stage. The safety analysis covered the operations from the reception of the waste transport casks to the disposal of the waste disposal package in its final emplacement location inside the disposal cell. Since the surface facilities' operations are similar to those of other nuclear ones, this paper focuses on the specificity of the deep repository, i.e. the operational safety and radioprotection aspects applied to the deep disposal drift. Andra has selected an ILW-LL design based on large horizontal drifts (diameters of 10 to 12 m, and lengths of 250 m). The primary waste packages are put inside a specific concrete overpack before their disposal. These overpacks are remotely stacked inside the horizontal drifts. The operational safety analysis aims to ensure that risks are kept under control through provisions in the design of the repository and by operating the facility in compliance with operational requirements and the safety functions. The requirements and the safety functions, developed at this stage of the feasibility studies, will be explained. The operational safety analysis is structured around physical components and real activities (construction, operation, closure) through a dedicated risk analysis. Due to the large variety of different ILW-LL waste, in order to identify the potential

  13. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

    2000-07-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

  14. 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.

  15. 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

  16. Effective safety training program design

    International Nuclear Information System (INIS)

    Chilton, D.A.; Lombardo, G.J.; Pater, R.F.

    1991-01-01

    Changes in the oil industry require new strategies to reduce costs and retain valuable employees. Training is a potentially powerful tool for changing the culture of an organization, resulting in improved safety awareness, lower-risk behaviors and ultimately, statistical improvements. Too often, safety training falters, especially when applied to pervasive, long-standing problems. Stepping, Handling and Lifting injuries (SHL) more commonly known as back injuries and slips, trips and falls have plagued mankind throughout the ages. They are also a major problem throughout the petroleum industry. Although not as widely publicized as other immediately-fatal accidents, injuries from stepping, materials handling, and lifting are among the leading causes of employee suffering, lost time and diminished productivity throughout the industry. Traditional approaches have not turned the tide of these widespread injuries. a systematic safety training program, developed by Anadrill Schlumberger with the input of new training technology, has the potential to simultaneously reduce costs, preserve employee safety, and increase morale. This paper: reviews the components of an example safety training program, and illustrates how a systematic approach to safety training can make a positive impact on Stepping, Handling and Lifting injuries

  17. Evaluation on safety issues of SMART

    International Nuclear Information System (INIS)

    Kim, W. S.; Seol, K. W.; Yoon, Y. K.; Lee, J. H.

    2001-01-01

    Safety issues on the SMART were evaluated in the light of the compliance with the Ministerial Ordinance of Technical Requirements applying to Nuclear Installations, which was recently revised. Evaluation concludes that regulatory requirements associated with following items have to be developed as the licensing criteria for the SMART: (1) proving the safety of design or materials different form existing reactors; (2) coping with beyond design basis accidents; (3) rulemaking on the safety of reactor safeguard vessel ; (4) ensuring integrity of steam generator tubes; and (5) classifying equipment based on their safety significance. Appropriate actions including implementation of new requirements under development should be taken for safety issues such as diversity of reactivity control and in-service inspection of steam generator tubes that are not complied with the current Technical Requirements. Safety level of the SMART design will be evaluated further by the more detailed assessment according to the Technical Requirements, and additional safety issues will be identified and resolved, if it necessary

  18. Key issues on safety design basis selection and safety assessment

    International Nuclear Information System (INIS)

    An, S.; Togo, Y.

    1976-01-01

    In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

  19. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

  20. Generic radiation safety design for SSRL synchrotron radiation beamlines

    Energy Technology Data Exchange (ETDEWEB)

    Liu, James C. [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States)]. E-mail: james@slac.stanford.edu; Fasso, Alberto [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Khater, Hesham [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Prinz, Alyssa [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Rokni, Sayed [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States)

    2006-12-15

    To allow for a conservative, simple, uniform, consistent, efficient radiation safety design for all SSRL beamlines, a generic approach has been developed, considering both synchrotron radiation (SR) and gas bremsstrahlung (GB) hazards. To develop the methodology and rules needed for generic beamline design, analytic models, the STAC8 code, and the FLUKA Monte Carlo code were used to pre-calculate sets of curves and tables that can be looked up for each beamline safety design. Conservative beam parameters and standard targets and geometries were used in the calculations. This paper presents the SPEAR3 beamline parameters that were considered in the design, the safety design considerations, and the main pre-calculated results that are needed for generic shielding design. In the end, the rules and practices for generic SSRL beamline design are summarized.

  1. Fire safety of wood construction

    Science.gov (United States)

    Robert H. White; Mark A. Dietenberger

    2010-01-01

    Fire safety is an important concern in all types of construction. The high level of national concern for fire safety is reflected in limitations and design requirements in building codes. These code requirements and related fire performance data are discussed in the context of fire safety design and evaluation in the initial section of this chapter. Because basic data...

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

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

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

  3. 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

  4. Technical basis for the ITER final design report, cost review and safety analysis (FDR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement. Refs, figs, tabs

  5. Technical basis for the ITER final design report, cost review and safety analysis (FDR)

    International Nuclear Information System (INIS)

    1998-01-01

    The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement

  6. Safety requirement of the nuclear power plants, after TMI-2 accident and their possible implementation on Bushehr NPP

    International Nuclear Information System (INIS)

    Mirhabibi, N.; Tochai, M.T.M.; Ashrafi, A.; Farnoudi, E.

    1985-01-01

    Based on the lessons learned from the TMI-2 accident and other research and developments, many improvements have been required for the design, manufacturing and operation of nuclear power plants in recent years. These requirements have already been implemented to the plants in operation and considered as new safety requirements for new plants. In the present paper these requirements and their possible implementation on Bushehr NPP are discussed. (Author)

  7. Consequences for designer and manufacturer of mechanical components due to future requirements in Europe

    International Nuclear Information System (INIS)

    Hans-Joachim, Frank

    2001-01-01

    In the frame of European harmonization, a lot of changes on requirements for designer and manufacturer of mechanical components have been performed. Differed organizations are involved in preparing future requirements for nuclear application. On one side the French German cooperation on the development of EPR. At the origin of this project was the common decision in 1989 of Framatome and Siemens to cooperate through NPI, to design the Nuclear Island, which meets the future needs of utilities. EDF and a group of the main German Utilities joined this cooperation in 1991 and since then they have been totally involved to the progress of the work. In addition, all the process was backed up to the end by the strong cooperation between the French and the German. Safety Authorities, which have a long lasting cooperation to define common requirements, which have to be applied to future Nuclear Power Plants. Furthermore an organization has been set up to elaborate common codes related to the EPR design, at the level of the French design and construction rules (RCC) or the German KTA safety standards, the so-called EPR technical codes (ETC). On the other side, the European utilities co-operate on a much broader basis for the establishment of European Utilities Requirements (EUR). These requirements are prepared by a group of European utilities that represent the major European electricity generating companies that are determined to keep the nuclear option open. The technical requirements specified in the EUR document define the boundaries in which future plants need to be designed in order to be acceptable for the needs of the utilities and in order to fulfill the basic requirements of competitive power generation costs and licensability in all countries represented in the EUR group. All the new requirements have to be applied by designer and manufacturer. Siemens /SNP act as a designer of a lot of various vessels and tanks, heat exchangers and other items of process

  8. 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...

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

    International Nuclear Information System (INIS)

    Nakai, Ryodai

    2013-01-01

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

  10. Guide to the safety design examination about light water reactor facilities for power generation

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This guide was compiled to evaluate the validity of the design policy when the safety design is examined at the time of the application for approval of the installation of nuclear reactors. About 7 years has elapsed since the existing guide was established, and the more appropriate guide to evaluate the safety should be made on the basis of the knowledge and experience accumulated thereafter. The range of application of this guide is limited to the above described evaluation, and it is not intended as the general standard for the design of nuclear reactors. First, the definition of the words used in this guide is given. Then, the guide to the safety examination is described about the general matters of reactor facilities, nuclear reactors and the measuring and controlling system, reactor-stopping system, reactivity-controlling system and safety protection system, reactor-cooling system, reactor containment vessels, fuel handling and waste treatment system. Several matters which require attention in the application of this guide or the clarification of the significance and interpretation of the guide itself were found, therefore the explanation about them was added at the end of this guide. (Kako, I.)

  11. Safety considerations in the design of PFBR

    International Nuclear Information System (INIS)

    Vaidyanathan, G.; Om Pal Singh; Govindarajan, S.; Chellapandi, P.; Chetal, S.C.; Shankar Singh, R.; Bhoje, S.B.

    1996-01-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe reactor under design in India. The overall safety approach adopted is based on the defence-in-depth principle. Design features have been incorporated to minimize occurrence of unsafe conditions. A plant protection system comprising reliable core monitoring to detect the off-normal condition, a reliable shutdown system to ensure safe shutdown and a passive decay heat removal system are provided. Containment is provided to prevent any release of radioactivity to the environment in case of failure of the protective devices. This paper provides a brief outline of the safety considerations in the design of PFBR. (author). 5 refs, 1 tab

  12. 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...

  13. European Workshop Industrical Computer Science Systems approach to design for safety

    Science.gov (United States)

    Zalewski, Janusz

    1992-01-01

    This paper presents guidelines on designing systems for safety, developed by the Technical Committee 7 on Reliability and Safety of the European Workshop on Industrial Computer Systems. The focus is on complementing the traditional development process by adding the following four steps: (1) overall safety analysis; (2) analysis of the functional specifications; (3) designing for safety; (4) validation of design. Quantitative assessment of safety is possible by means of a modular questionnaire covering various aspects of the major stages of system development.

  14. Reactor design and safety approach for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Davies, S.M.; Yamaki, Hideo; Goodman, L.

    1984-06-01

    A tank type plant has been designed that offers compactness, high reliability under seismic and thermal transients, and a safety design approach that provides a balance between public safety and plant availability. This report provides a description of the design philosophy and safety features of the reactor

  15. 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.)

  16. Outline of the requirements of application of computer based instrumentation and control systems in the systems important to safety on Bohunice NPPs

    International Nuclear Information System (INIS)

    Bacurik, J.

    1997-01-01

    The most important regulatory requirements and issues are described related to the review, evaluation and assessment of computer-based safety-related IandC systems, with emphasis on safety instrumentation and control. These aspects include safety classification and categorization of IandC, ranking of applicable codes and standards, design evaluation on the system level, and software assessment. (author)

  17. Nuclear reactor conceptual design: methodology for cost-effective internalisation of nuclear safety

    International Nuclear Information System (INIS)

    Gimenez, M.; Grinblat, P.; Schlamp, M.

    2002-01-01

    A novel and promising methodology to perform nuclear reactor design is presented in this work. It achieves to balance efficiently safety and economics at the conceptual engineering stage. The key to this integral approach is to take into account safety aspects in a design optimisation process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behaviour during accidents and from its probabilistic safety assessment -safety performance indicators-, are synthesised on Safety Design Maps. These maps allow one to compare these indicators with limit values, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimisation process, by means of additional rules to the neutronic, thermal-hydraulic and mechanical calculations. This methodology turns out to be promising to balance and optimise reactor and safety system design in an early engineering stage, in order to internalise cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels. Furthermore, through this methodology, a simplified design can be obtained, compared to the resultant complexity when these concepts are introduced in a later engineering stage. (author)

  18. Lightning protection design of nuclear power plants. KTA safety code, version 6/99

    International Nuclear Information System (INIS)

    1999-06-01

    This KTA safety code does not cover calculation methods for determination of lightning-induced voltage inputs to control room systems within the reactor building, as the literature presents a variety of applicable methods, which however cannot be directly applied to any power plant, due to the great diversity of geometries of the electrical systems and control room systems in nuclear power plants. Compliance with the design requirements of this safety code for shielding of buildings, and installation and shielding of cables, can be considered to offer the appropriate protection. (orig./CB) [de

  19. 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

  20. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

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

  1. Distinctive safety aspects of the CANDU-PHW reactor design

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    Two lectures are presented in this report. They were prepared in response to a request from IAEA to provide information on the 'Special characteristics of the safety analysis of heavy water reactors' to delegates from member states attending the Interregional Training Course on Safety Analysis Review, held at Karlsruhe, November 19 to December 20, 1979. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (auth)

  2. Balancing safety and economics

    International Nuclear Information System (INIS)

    Kroeger, W.; Fischer, P.U.

    2000-01-01

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

  3. 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)

  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. 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.

  8. Incorporation of Safety into Design Process : A Systems Engineering Perspective

    NARCIS (Netherlands)

    Rajabalinejad, M.

    2018-01-01

    This paper suggests integrating the best safety practices with the design process. This integration enriches the exploration experience for designers and adds extra values and competitor advantages for customers. The paper introduces the safety cube for combining common blocks for design, hazard

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  10. Development of Safety Analysis Technology for Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

  11. Study on Fusion Safety Infrastructure using ISAM

    International Nuclear Information System (INIS)

    Oh, Kyemin; Kang, Myungsuk; Heo, Gyunyoung; Kim, Hyoungchan

    2013-01-01

    The regulation of nuclear facilities have checked and managed safety throughout the entire process from design, construction, operation and decommissioning. Also, the same meaning as the regulatory requirements and design requirements, it will be important indicators for detailed design of K-DEMO. K-DEMO has many uncertainties because it is in conceptual design phase. Also, there is no reference material because demonstration scale fusion power plants were not operated yet in overseas. So, hazard that threaten the integrity of K-DEMO have to be defined preferentially to define regulatory or design requirements. This study proposed method that educe regulatory or design requirements and introduce web-based cloud infrastructure to perform renewal and sharing of information related with safety that is required in the study rapidly as a part of the R and D program funded by National Fusion Research Institute of Korea (NFRI). We have been performing QSR and PIRT in accordance with development of fusion DEMO plant, and preparing OPT, PSA and DPA for regulation requirements. This study introduces our recent research activities about ISAM for fusion and CCI built for expert and extant safety related information. Unlike fission, nuclear fusion's safety goal is non-evacuation of the public during an accident. To satisfy this goal not only various safety issues should be analyzed, but safety objectives, regulatory requirements, and design variables should also be established in detailed design phase. The web-based cloud infrastructure proposed in this paper will be able to offer input data of future studies and, it is expected to contribute on general and technical safety principles for national fusion power plant technology plan

  12. Cognitive Connected Vehicle Information System Design Requirement for Safety: Role of Bayesian Artificial Intelligence

    Directory of Open Access Journals (Sweden)

    Ata Khan

    2013-04-01

    Full Text Available Intelligent transportation systems (ITS are gaining acceptance around the world and the connected vehicle component of ITS is recognized as a high priority research and development area in many technologically advanced countries. Connected vehicles are expected to have the capability of safe, efficient and eco-driving operations whether these are under human control or in the adaptive machine control mode of operations. The race is on to design the capability to operate in connected traffic environment. The operational requirements can be met with cognitive vehicle design features made possible by advances in artificial intelligence-supported methodology, improved understanding of human factors, and advances in communication technology. This paper describes cognitive features and their information system requirements. The architecture of an information system is presented that supports the features of the cognitive connected vehicle. For better focus, information processing capabilities are specified and the role of Bayesian artificial intelligence is defined for data fusion. Example applications illustrate the role of information systems in integrating intelligent technology, Bayesian artificial intelligence, and abstracted human factors. Concluding remarks highlight the role of the information system and Bayesian artificial intelligence in the design of a new generation of cognitive connected vehicle.

  13. Cascade Distillation System Design for Safety and Mission Assurance

    Science.gov (United States)

    Sarguisingh, Miriam; Callahan, Michael R.; Okon, Shira

    2015-01-01

    Per the NASA Human Health, Life Support and Habitation System Technology Area 06 report "crewed missions venturing beyond Low-Earth Orbit (LEO) will require technologies with improved reliability, reduced mass, self-sufficiency, and minimal logistical needs as an emergency or quick-return option will not be feasible".1 To meet this need, the development team of the second generation Cascade Distillation System (CDS 2.0) chose a development approach that explicitly incorporate consideration of safety, mission assurance, and autonomy. The CDS 2.0 preliminary design focused on establishing a functional baseline that meets the CDS core capabilities and performance. The critical design phase is now focused on incorporating features through a deliberative process of establishing the systems failure modes and effects, identifying mitigation strategies, and evaluating the merit of the proposed actions through analysis and test. This paper details results of this effort on the CDS 2.0 design.

  14. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    International Nuclear Information System (INIS)

    Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

    2014-01-01

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  15. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

    2014-10-15

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  16. 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.

  17. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 1 - guideword applicability and method description.

    Science.gov (United States)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-12-15

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design.

  18. 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.

  19. 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

  20. 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

  1. Reliability Improved Design for a Safety System Channel

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Eung Se; Kim, Yun Goo [KHNP, Daejeon (Korea, Republic of)

    2016-05-15

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced.

  2. Reliability Improved Design for a Safety System Channel

    International Nuclear Information System (INIS)

    Oh, Eung Se; Kim, Yun Goo

    2016-01-01

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced

  3. Safety principles and design management of Chashma Nuclear Power Plant

    International Nuclear Information System (INIS)

    Geng Qirui; Cheng Pingdong

    1997-01-01

    The basic safety consideration and detailed design principles in the design of Chashma Nuclear Power Plant is elaborated. The management within the frame setting up by 'safety culture' and 'quality culture'

  4. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  5. 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

  6. Safety considerations in the design of the fusion engineering device

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1983-01-01

    Safety considerations play a significant role in the design of a near-term Fusion Engineering Device (FED). For the safety of the general public and the plant workers, the radiation environment caused by the reacting plasma and the potential release of tritium fuel are the dominant considerations. The U.S. Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris. The highly activated device components make it necessary to design many of the maintenance activities in the reactor building for totally remote operation. The hot cell facility has evolved as a totally remote maintenance facility due to the high radiation levels of the device components. Safety considerations have had substantial impacts on the design of FED. Several examples of safety-related design impacts are discussed in the paper. Feasible solutions have been identified for all outstanding safety-related items, and additional optimization of these solutions is anticipated in future design studies

  7. The comparison between layout and design of APR1400 MCR and EPRI requirement

    International Nuclear Information System (INIS)

    Suharyo Widagdo; Darlis; Sigit Santoso; T J Suryono

    2013-01-01

    A good man-machine interface system is needed by a nuclear power reactor because of its complexity and high safety factor. The man-machine interface system is centralized in the main control room (MCR). All of monitoring and controlling activity to keep the installation operation safety can be done from this place. In order to increase its ability in monitoring and controlling a hard effort to raise the interface system to the highest possible level is done not only by using the latest computer technology and instrumentation system but also using Human Engineering, as well. The paper aims to compare the lay out and design the APR1400 MCR and the requirement from EPRI about main control room lay out and design as response of TMI-2 accident. We can see from the discussion that the design of the MCR of APR1400 have been applied the request of EPRI. (author)

  8. 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.

  9. Design Information from the PSA for Digital Safety-Critical Systems

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Jang, Seung Cheol

    2005-01-01

    Many safety-critical applications such as nuclear field application usually adopt a similar design strategy for digital safety-critical systems. Their differences from the normal design for the non-safety-critical applications could be summarized as: multiple-redundancy, highly reliable components, strengthened monitoring mechanism, verified software, and automated test procedure. These items are focusing on maintaining the capability to perform the given safety function when it is requested. For the past several decades, probabilistic safety assessment (PSA) techniques are used in the nuclear industry to assess the relative effects of contributing events on plant risk and system reliability. They provide a unifying means of assessing physical faults, recovery processes, contributing effects, human actions, and other events that have a high degree of uncertainty. The applications of PSA provide not only the analysis results of already installed system but also the useful information for the system under design. The information could be derived from the PSA experience of the various safety-critical systems. Thanks to the design flexibility, the digital system is one of the most suitable candidates for risk-informed design (RID). In this article, we will describe the feedbacks for system design and try to develop a procedure for RID. Even though the procedure is not sophisticated enough now, it could be the start point of the further investigation for developing more complete and practical methodology

  10. Analysis and design on airport safety information management system

    Directory of Open Access Journals (Sweden)

    Yan Lin

    2017-01-01

    Full Text Available Airport safety information management system is the foundation of implementing safety operation, risk control, safety performance monitor, and safety management decision for the airport. The paper puts forward the architecture of airport safety information management system based on B/S model, focuses on safety information processing flow, designs the functional modules and proposes the supporting conditions for system operation. The system construction is helpful to perfecting the long effect mechanism driven by safety information, continually increasing airport safety management level and control proficiency.

  11. 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...

  12. National Ignition Facility subsystem design requirements target area auxiliary subsystem SSDR 1.8.6

    International Nuclear Information System (INIS)

    Reitz, T.

    1996-01-01

    This Subsystem Design Requirement (SSDR) establishes the performance, design, development, and test requirements for the Target Area Auxiliary Subsystems (WBS 1.8.6), which is part of the NIF Target Experimental System (WBS 1.8). This document responds directly to the requirements detailed in NIF Target Experimental System SDR 003 document. Key elements of the Target Area Auxiliary Subsystems include: WBS 1.8.6.1 Local Utility Services; WBS 1.8.6.2 Cable Trays; WBS 1.8.6.3 Personnel, Safety, and Occupational Access; WBS 1.8.6.4 Assembly, Installation, and Maintenance Equipment; WBS 1.8.6.4.1 Target Chamber Service System; WBS 1.8.6.4.2 Target Bay Service Systems

  13. Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling

    International Nuclear Information System (INIS)

    JANICEK, G.P.

    2000-01-01

    Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance

  14. Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling

    Energy Technology Data Exchange (ETDEWEB)

    JANICEK, G.P.

    2000-06-08

    Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance.

  15. Design and safety studies on an EFIT core with CERMET fuel

    International Nuclear Information System (INIS)

    Chen, Xue-Nong; Rineiski, Andrei; Liu, Ping; Maschek, Werner; Matzerath Boccaccini, Claudia; Gabrielli, Fabrizio; Sobolev, Vitaly

    2008-01-01

    Within the EUROTRANS Programme a European Facility for Industrial Transmutation (EFIT) is under development. This paper deals with the design and safety analyses of an EFIT core with Mo-matrix based CERMET fuel. A three zone core design was developed, which satisfies the EFIT general and specific requirements. The fuel/matrix ratio in each zone is determined for a suitable subcritical level at a k eff of about 0.97 and a total form factor around 1.5. The Pu/MA ratio also determines the transmutation rate and the burn-up characteristics, ranging between 46/54 at% to 40/60 at% for optimizing the reactivity swing and the MA transmutation efficiency. Based on the preliminary core design, safety calculations are performed with SIMMER-III for three types of transient: the unprotected loss of flow (ULOF), the unprotected transient of over power (UTOP) and the unprotected blockage accident (UBA). It can be shown that in the CERMET core the fuel and clad design limits are not violated under the conditions of ULOF and UTOP. In the UBA case, pin failures will happen and lead to a local voiding and reactivity insertion, but a fuel sweep-out process leads to a power reduction and restricts the core degradation. (authors)

  16. The safety designs for the TITAN reversed-field pinch reactor study

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.; Hoot, C.G.; Schultz, K.R.; Grotz, S.P.; Blanchard, J.; Sharafat, S.; Najmabadi, F.

    1989-01-01

    TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2 . The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections, and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated at a level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated at a level 2 of safety assurance. (orig.)

  17. The safety designs for the TITAN reversed-field pinch reactor study

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.; Hoot, C.G.; Schultz, K.R.; Grotz, S.P.; Blanchard, J.P.; Sharafat, S.; Najmabadi, F.

    1988-01-01

    TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2 . The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated as level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated as level 2 of safety assurance. 7 refs., 2 figs

  18. Design for safety: A cognitive engineering approach to the control and management of nuclear power plants

    International Nuclear Information System (INIS)

    Boy, Guy A.; Schmitt, Kara A.

    2013-01-01

    Highlights: ► Complexity must be understood and handled well in order to design for safety. ► Complexity can be reduced during design by using the AUTOS pyramid model. ► Procedures are human automation, much as software is machine automation. ► Identifying emergent behaviors reduces procedure accumulation. ► Human-in-the-loop-simulations help to understand emergent behaviors. -- Abstract: This paper presents an analytical approach to design for safety that is based on 30 years of experience in the field of Human-centered design. This field is often qualified as governing safety–critical systems where risk management is a crucial issue. We need to better understand what the main facets of safety are that should be taken into account during the design and development processes. There are many factors that contribute to design for safety. We propose some of these factors and an articulation of them from requirement gathering and synthesis to formative evaluations to summative evaluations. Among these factors, we analyze complexity, flexibility, stability, redundancy, support, training, experience and testing. However, we cannot design a safe and reliable product in one shot; design is incremental. A product and its various uses become progressively mature. When we deal with new products, issues come from the fact that practice features emerge from the use of the product and are difficult, even impossible, to predict ahead of time. The automation within is an important portion of this maturity, and must be understood well. This is why design for safety is not possible without anticipatory simulations and a period of tests in the real world, such as operational testing in nuclear power plants. In addition, designing for safety is not finished when the product is delivered; experience feedback, or human-in-the-loop simulation (HITLS) is an important part of the overall global design process. The AUTOS pyramid approach can assist in simplifying the

  19. 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)

  20. Safety design study of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.

    1992-01-01

    This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

  1. Research reactor safety - an overview of crucial aspects

    International Nuclear Information System (INIS)

    Laverie, M.

    1998-01-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  2. Safety methodology implementation in the conceptual design phase of a fusion reactor

    International Nuclear Information System (INIS)

    Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

    2007-01-01

    The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

  3. Westinghouse Advances in Passive Plant Safety

    International Nuclear Information System (INIS)

    Bruschi, H. J.; Manager, General; Gerstenhaber, E.

    1993-01-01

    On June 26, 1992, Westinghouse submitted the Ap600 Standard Safety Analysis Report and comprehensive PIRA results to the U. S. NRC for review as part of the Ap600 design certification program. This major milestone was met on time on a schedule set more than 3 years before submittal and is the result of the cooperative efforts of the U. S. Department of Energy (DOE), the Electric Power Requirements Program, and the Westinghouse Ap600 design team. These efforts were initiated in 1985 to develop a 600 MW advanced light water reactor plant design based on specific technical requirements established to provide the safety, simplicity, reliability, and economics necessary for the next generation of nuclear power plants. The Ap600 design achieves the ALRR safety requirements through ample design margins, simplified safety systems based on natural driving forces, and on a human-engineered man-machine interface system. Extensive Probabilistic Risk evolution, have recently shown that even if none of the active defense-in-depth safety systems are available, the passive systems alone meet safety goals. Furthermore, many tests in an extensive test program have begun or have been completed. Early tests show that passive safety perform well and meet design expectations

  4. AP1000 Containment Design and Safety Assessment

    International Nuclear Information System (INIS)

    Wright, Richard F.; Ofstun, Richard P.; Bachere, Sebastien

    2002-01-01

    The AP1000 is an up-rated version of the AP600 passive plant design that recently received final design certification from the US NRC. Like AP600, the AP1000 is a two-loop, pressurized water reactor featuring passive core cooling and passive containment safety systems. One key safety feature of the AP1000 is the passive containment cooling system which maintains containment integrity in the event of a design basis accident. This system utilizes a high strength, steel containment vessel inside a concrete shield building. In the event of a pipe break inside containment, a high pressure signal actuates valves which allow water to drain from a storage tank atop the shield building. Water is applied to the top of the containment shell, and evaporates, thereby removing heat. An air flow path is formed between the shield building and the containment to aid in the evaporation and is exhausted through a chimney at the top of the shield building. Extensive testing and analysis of this system was performed as part of the AP600 design certification process. The AP1000 containment has been designed to provide increased safety margin despite the increased reactor power. The containment volume was increased to accommodate the larger steam generators, and to provide increased margin for containment pressure response to design basis events. The containment design pressure was increased from AP600 by increasing the shell thickness and by utilizing high strength steel. The passive containment cooling system water capacity has been increased and the water application rate has been scaled to the higher decay heat level. The net result is higher margins to the containment design pressure limit than were calculated for AP600 for all design basis events. (authors)

  5. Safety critical software design approach developed for Canadian nuclear power plants

    International Nuclear Information System (INIS)

    Ichiyen, M.M.; Joannou, P.K.

    1995-01-01

    Recently two methodologies were developed that comply with a high safety critical standard: the Rational Design Process, which can be characterized as a methodology based on state machines where the required behaviour of the software is defined using mathematical functions written in a notation which has a well defined syntax and semantics, and the Integrated Approach, which uses a graphical functional notation to specify the functional software requirements. The first implementations based on the two methodologies are discussed. Results from all phases of testing show a remarkably low number of errors, demonstrating that the new methodologies have indeed led to a higher demonstrable level of software reliability. (orig./HP) [de

  6. Safety design and evaluation policy for future FBRs in Japan

    International Nuclear Information System (INIS)

    Aizawa, Kiyoto

    1991-01-01

    The safety policy for fast breeder reactors (FBRs) has gradually matured in accordance with the development of FBRs. The safety assessment of the Japanese prototype FBR, Monju during the licensing process accelerated the maturity and the integration of knowledge and databases. Results are expected to be reflected in the establishment of the safety design and evaluation policy for FBRs. Although the methodologies and safety policies developed for LWRs are applicable in principle to future FBRs, it is neither rational nor realistic to treat safety only with these policies. It is recommended that one should develop the methodologies and safety policies starting from understanding of the inherent safety characteristics of FBR's through safety research, plant operating experience and design work. In the last few years, some technical committees were organized in Japan and have discussed key safety issues which are specific to FBRs in order to provide preparatory reports and to establish safety standards and guidelines for future commercial FBRs. (author)

  7. Development of quantitative goals for inherent safety feature design and licensing

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.

    1987-01-01

    There is now considerable interest in the development of advanced fast reactors whose major focus is inherent safety. The achievement of inherent safety can be viewed from several aspects. In the Integral Fast Reactor Concept the approach is to utilize the intrinsic characteristics of pool-type liquid metal fast breeder reactors (LMFBRs) and the properties of metal fuels to integrate a high degree of inherent safety into the design. The PRISM and SAFR concepts focus on other inherent safety features. The reactors discussed above represent a radical departure from existing LWR designs as well as previous LMFBR designs (e.g., CRBRP) which are based, for the most part, on the General Design Criteria found in 10CFR50 Appendix. In view of these parallel developments (advanced reactors exploiting inherent safety and the use of quantitative goals to augment licensing), there appears to be a need to perform research on the development of methods for designing, assessing, and licensing inherent safety features in advanced reactors. The objectives of such research are outlined

  8. Designing continuous safety improvement within chemical industrial areas

    NARCIS (Netherlands)

    Reniers, G.L.L.; Ale, B. J.M.; Dullaert, W.; Soudan, K.

    This article provides support in organizing and implementing novel concepts for enhancing safety on a cluster level of chemical plants. The paper elaborates the requirements for integrating Safety Management Systems of chemical plants situated within a so-called chemical cluster. Recommendations of

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

    International Nuclear Information System (INIS)

    Ostenso, A.; May, R.

    1996-12-01

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

  11. Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process.

  12. Basic requirements for a preliminary conceptual design of the Korea advanced pyroprocess facility (KAPF)

    International Nuclear Information System (INIS)

    Lee, Ho Hee; Ko, Won Il; Chang, Hong Lae; Song, Dae Yong; Kwon, Eun Ha; Lee, Jung Won

    2008-12-01

    Korea Atomic Energy Research Institute (KAERI) has been developing technologies for pyroprocessing for spent PWR fuels. This study is part of a long term R and D program in Korea to develop an advanced recycle system that has the potential to meet and exceed the proliferation resistance, waste minimization, resource minimization, safety and economic goals of approved Korean Government energy policy, as well as the Generation IV International Forum (GIF) program. To support this R and D program, KAERI requires that an independent estimate be made of the conceptual design and cost for construction and operation of a 'Korea Advanced Pyroprocessing Facility', This document describes the basic requirements for preliminary conceptual design of the Korea Advanced Pyroprocess Facility (KAPF). The presented requirements will be modified to be more effective and feasible on an engineering basis during the subsequent design process

  13. Design requirements for new nuclear reactor facilities in Canada (focus on important improvements from RD-337 version 1 of 2008)

    International Nuclear Information System (INIS)

    Shim, S.; Harwood, C.; Ohn, M-Y; Liu, Y.C.; Young, T.

    2014-01-01

    The Canadian Nuclear Safety Commission (CNSC) has established the regulatory framework that includes the documentation of the requirements and guidance for each of CNSC's 14 Safety and Control Areas, one important area being the design of nuclear power plants (NPPs). For the design area, the CNSC published RD-337 version 1 Design of New Nuclear Power Plants in 2008. As such regulatory documents are reviewed on a regular basis, this document was recently updated as RD-337 version 2, and its guidance document GD-337, Guidance for the Design of New Nuclear Power Plants was developed to provide guidance on how to meet the requirements. REGDOC-2.5.2 Design of Reactor Facilities: Nuclear Power Plants that combines RD-337 version 2 and GD-337 version 1 was presented to the Commission on March 27, 2014 after two rounds of consultation with stakeholders, and was subsequently published in May 28, 2014. Although REGDOC-2.5.2 maintains the structure and the contents nearly the same as RD-337 version 1, it introduces several important improvements to: Include GD-337 guidance for further clarity to applicants, licensees and vendors on how to meet the requirements. This guidance provides the review criteria considered in CNSC staff's review in a transparent way; Ensure alignment with international standards including recent IAEA SSR 2/1, Safety of Nuclear Power Plants: Design; Implement CNSC Fukushima Task Force Report findings that pertain to the design of reactor facilities for severe accidents; and, Make necessary improvements such as addition of requirements for cyber security. This paper describes the overall regulatory framework related to CNSC's design requirements and guidance for NPPs, and focus on the important improvements included in REGDOC-2.5.2 and their reasoning. (author)

  14. Design requirements for new nuclear reactor facilities in Canada (focus on important improvements from RD-337 version 1 of 2008)

    Energy Technology Data Exchange (ETDEWEB)

    Shim, S.; Harwood, C.; Ohn, M-Y; Liu, Y.C.; Young, T. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2014-07-01

    The Canadian Nuclear Safety Commission (CNSC) has established the regulatory framework that includes the documentation of the requirements and guidance for each of CNSC's 14 Safety and Control Areas, one important area being the design of nuclear power plants (NPPs). For the design area, the CNSC published RD-337 version 1 Design of New Nuclear Power Plants in 2008. As such regulatory documents are reviewed on a regular basis, this document was recently updated as RD-337 version 2, and its guidance document GD-337, Guidance for the Design of New Nuclear Power Plants was developed to provide guidance on how to meet the requirements. REGDOC-2.5.2 Design of Reactor Facilities: Nuclear Power Plants that combines RD-337 version 2 and GD-337 version 1 was presented to the Commission on March 27, 2014 after two rounds of consultation with stakeholders, and was subsequently published in May 28, 2014. Although REGDOC-2.5.2 maintains the structure and the contents nearly the same as RD-337 version 1, it introduces several important improvements to: Include GD-337 guidance for further clarity to applicants, licensees and vendors on how to meet the requirements. This guidance provides the review criteria considered in CNSC staff's review in a transparent way; Ensure alignment with international standards including recent IAEA SSR 2/1, Safety of Nuclear Power Plants: Design; Implement CNSC Fukushima Task Force Report findings that pertain to the design of reactor facilities for severe accidents; and, Make necessary improvements such as addition of requirements for cyber security. This paper describes the overall regulatory framework related to CNSC's design requirements and guidance for NPPs, and focus on the important improvements included in REGDOC-2.5.2 and their reasoning. (author)

  15. Safety design integrated in the Building Delivery System

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2012-01-01

    phases of the building delivery system by using the principle of the lean construction modelling. The method for the research was to go through the lean construction building delivery system step by step and create a normative description of what to do, when to do and how to do to fully integration...... of safety in each process. The group of participants who created the description had a high experience in a combination of research, safety and health in general and especial in construction and knowledge of the lean construction processes both from the clients perspective as well as from the designers...... and the consultants. The result is a concept and guideline including control schemes for how to integrate safety design in the lean construction building delivery system including what to do and when. The concept has been tested in an educational context and found useful by the designers. The practical value...

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

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

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2012-01-01

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

  18. Safety design philosophy of the ABWR for the next generation LWRs

    International Nuclear Information System (INIS)

    Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

    2009-01-01

    The paper presents safety design philosophy of the advanced boiling water reactor (ABWR) to be reflected in developing the next generation light water reactors (LWRs). The basic policy of the ABWR safety design was to improve safety and reduce cost simultaneously by reflecting lessons learned of precursors, incidents and accidents that were beyond the design basis such as the Three Mile Island Unit 2 (TMI 2) accident. The ABWR is a fully active safety plant. The ABWR enhanced redundancy and diversity of active safety systems using probabilistic safety assessment (PSA) insights. It adopted a complete three division active emergency core cooling system (ECCS) and attained a very low core damage frequency (CDF) value of less than 10 -7 /ry for internal events. Only very small residual risks, if any, rather exist in external events such as an extremely large earthquake beyond the design basis. This is because external events can constitute a common cause that disables all the redundant active safety systems. Therefore, it is useless to add one more ECCS train and make a four division active ECCS for external events. Nowadays, however, fully passive safety LWRs are already established. Incorporating some of these passive safety systems we can also establish the next generation LWRs that are truly strong against external events. We can establish a plant that can survive a giant earthquake at least three days without AC power source, SA proof safety design that enables no containment failure and no evacuation to eliminate the residual risks. The same basic policy as the ABWR to improve safety and reduce cost simultaneously is again effective for the next generation LWRs. (author)

  19. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

    General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

  20. Research reactor safety - an overview of crucial aspects

    Energy Technology Data Exchange (ETDEWEB)

    Laverie, M. [Atomic Energy Commission, Saclay, F-91191 Gif sur Yvette (France)

    1998-07-01

    Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

  1. Design safety improvements of Kozloduy NPP to meet the modern safety requirements towards the old generation PWR

    International Nuclear Information System (INIS)

    Hinovski, M.P.; Sabinov, S.

    2001-01-01

    Activities related to safety improvement of Kozloduy NPP units, started at the end of 1970s included seismic resistance upgrading, fire safety improvement, reliable heat final absorber etc. During the last 10 years the approach was systematized and improved. Units 1 to 4 are of great interest; therefore here we will discuss these units only. As a result of studies and analyses performed at the end of the 1980s and the beginning of the 1990s, problems related to the safety were identified and complex of technical measures was developed and planned. A considerable part of these measures has already been implemented, and the rest will be performed during the next years. Activities were performed by stages, and at the moment the last stage is under way. It shall be finished by the year 2003. The number of the measures is quite large to describe them here in full scope -- during the first stage of the safety program (1991-1993) were developed and analyzed more than 4200 documents and more than 160 measures were executed. During the second and third stages more than 300 important improvements were realized. In the frame of the program, financed by EBRD, 10 new systems with great importance were implemented and 8 systems were significantly modified. The main measures are described below. (author)

  2. 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

  3. Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

    International Nuclear Information System (INIS)

    2000-01-01

    OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants

  4. Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-08-01

    OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.

  5. Design of an Active Automotive Safety System

    Directory of Open Access Journals (Sweden)

    Y. Wang

    2013-07-01

    Full Text Available With the development of the national economy, the people's standard of living got corresponding improvement, cars has been one of the indispensable traffic tools in many families. An active safety system is proposed, which can real-time detect the vehicle's running status and judge the security status of the vehicle. The system, which takes single-chip microcomputer as the controlling core and combines with millimeter-wave and ultrasonic distance measurement technology, can detect the distance from vehicle to vehicle and judge the security status of the vehicle. The hardware composition of the system and the data acquiring circuit are proposed, the mathematic model for different situation is established, and the controlling algorithm is completed. This system can accurately measure speed and distance between vehicles; the active safety control system can meet the relevant data measurement and transmission requirement; and can meet the functional requirement of the active safety control system

  6. Design characteristics of safety parameter display system for nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Yuangfang

    1992-02-01

    The design features of safety parameter display system (SPDS) developed by Tsinghua University is introduced. Some new features have been added into the system functions and they are: (1) hierarchical display structure; (2) human factor in the display format design; (3)automatic diagnosis of safety status of nuclear power plant; (4) extension of SPDS use scope; (5) flexible hardware structure. The new approaches in the design are: (1)adopting the international design standards; (2) selecting safety parameters strictly; (3) developing software under multitask operating system; (4) using a nuclear power plant simulator to verify the SPDS design

  7. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. ITER safety challenges and opportunities

    International Nuclear Information System (INIS)

    Piet, S.J.

    1991-01-01

    Results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ''ITER is capable of meeting anticipated regulatory dose limits,'' but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. We need much research and development (R ampersand D) and design analysis to establish that ITER meets regulatory requirements. We have a further opportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, we need to overcome three programmatic challenges and three technical challenges. The first programmatic challenge is to fund a comprehensive safety and environmental ITER R ampersand D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of the three key technical challenges is plasma engineering -- burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost. Some design requirements are now too strict; some are too lax. Fuel cycle design requirements are presently too strict, mandating inappropriate T separation from H and D. Heat sink requirements are presently too lax; they should be strengthened to ensure that maximum loss of coolant accident temperatures drop

  11. 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

  12. Technical Safety Requirements for the B695 Segment

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-09-11

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

  13. Technical Safety Requirements for the B695 Segment

    International Nuclear Information System (INIS)

    Laycak, D.

    2008-01-01

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

  14. Safety considerations in the design of an electron beam melter for a fissile metal

    International Nuclear Information System (INIS)

    Jenkins, T.W.

    1995-01-01

    The purpose of the EBM in the melting of highly enriched uranium metal is principally to recycle machining swarf but it may also be used for the manufacture of alloys and the recycling of alloy swarf. The current recycling of swarf involves the cleaning and compaction of swarf to pellets, these are then melted in a VIF with losses of > 20% of the metal as dross lift in the crucible. This has to be converted to oxide and then chemically recycled, a very expensive process. Use of the EBM would limit losses to less than 3% and make significant savings and also solve some safety problems. The EBM will be attached to a glove box and fume cupboard and the hazards addressed by the design include: (1) Criticality, (2) Radiation, (3) Release of Radioactive Contamination, (4) Explosion, (5) Fire, (6) Mechanical Handling, (7) Electrical, (8) Other Safety Considerations. These are addressed with details of the considerations including interlocks required to mitigate the hazards. The requirements of Safety documentation and Hazard and Operability studies are outlined together with Quality Assurance demands and training requirements. A number of safety considerations are based on previous accident scenarios in which Dangerous Occurrences took place involving equipment faults and operator errors on high vacuum equipment associated with glove boxes in a radioactive controlled area

  15. 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.

  16. Safety design concept and analysis for the upgrading JRR-3

    International Nuclear Information System (INIS)

    Onishi, N.; Isshiki, M.; Takahashi, H.; Takayanagi, M.

    1990-01-01

    The Research Reactor No.3 (JRR-3) is under reconstruction for upgrading. This paper describes the safety design concepts of the architectural and engineering design, anticipated operational transients and accident conditions which are the postulated initiating events for the safety evaluation, and the safety criteria of the upgraded JRR-3. The safety criteria are defined taking into account those of Light Water Reactors and the characteristics of the research reactor. Using the example of the safety analysis, this paper describes analytical results of a reactivity insertion by removal of in-core irradiation samples, a pipeline break at the primary coolant loop and flow blockage to a coolant channel, which are the severest postulated initiating events of the JRR-3

  17. Development of safety and regulatory requirements for Korean next generation reactor - Development of human factors design review guidelines (II)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Woon; Oh, In Suk; Lee, Hyun Chul; Cheon, Se Woo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    The objective of this study is to develop human factors engineering program review guidelines and alarm system review guidelines in order to resolve the two major technical issues: '25. Human Factors Engineering Program Review Model' and '26. Review Criteria for Human Factors Aspects of Advanced Controls and Instrumentation', which are related to the development of human factors safety regulation guides being performed by KINS. For the development of human factors program review guidelines, we made a Korean version of NUREG-0711 and added our comments by considering Korean regulatory situation and the characteristics of the KNGR design, and reviewing the reference documents of NURGE-0711. We also computerized the Korean version of NUREG-0711, additional comments, and selected portion of the reference documents for the developer of safety regulation guides at KINS to see the contents comparatively at a glance and use them easily. For the development of alarm system design review guidelines, we made a Korean version of NUREG/CR-6105, which was published by NRC in 1994 as a guideline document for the human factors review of alarm systems. Then we updated the guidelines by reviewing the literature related to alarm design that published after 1994. 12 refs., 11 figs., 2 tabs. (Author)

  18. Development of safety and regulatory requirements for Korean next generation reactor - Development of human factors design review guidelines (II)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Woon; Oh, In Suk; Lee, Hyun Chul; Cheon, Se Woo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    The objective of this study is to develop human factors engineering program review guidelines and alarm system review guidelines in order to resolve the two major technical issues: '25. Human Factors Engineering Program Review Model' and '26. Review Criteria for Human Factors Aspects of Advanced Controls and Instrumentation', which are related to the development of human factors safety regulation guides being performed by KINS. For the development of human factors program review guidelines, we made a Korean version of NUREG-0711 and added our comments by considering Korean regulatory situation and the characteristics of the KNGR design, and reviewing the reference documents of NURGE-0711. We also computerized the Korean version of NUREG-0711, additional comments, and selected portion of the reference documents for the developer of safety regulation guides at KINS to see the contents comparatively at a glance and use them easily. For the development of alarm system design review guidelines, we made a Korean version of NUREG/CR-6105, which was published by NRC in 1994 as a guideline document for the human factors review of alarm systems. Then we updated the guidelines by reviewing the literature related to alarm design that published after 1994. 12 refs., 11 figs., 2 tabs. (Author)

  19. Top-tier requirements for KNGR

    International Nuclear Information System (INIS)

    Sung-Jae, Ch.; Kwangho, L.; Dong Wook, J.

    1996-01-01

    In 1992, Korea Electric Power Corporation (KEPCO) has launched the next generation reactor project to develop the standard design of an advanced pressurized water reactor by 2000. This advanced reactor aims to have the sufficient capability to be a safe, environmentally sound and economical energy source for 2000's in Korea. In conjunction with the project development, the program phase I is studied and it is in the Korean Next Generation Reactor (KNGR) first phase project that the requirements of this specification called ''Top-tier'' have been established. These functional requirements are of the first importance for the design, construction and operation of a nuclear power plant. These requirements are divided into safety requirements, serious accidents control, design base requirements, definition of the system characteristics, performance, construction feasibility, economical objectives, site parameters and design processes. The ''Top-tier'' requirements are concentrated on the improvement of the safety and reliability. Safety is one of the first priorities. In particular, the requirements for the design of the next reactors generation must include the capacity to control serious accidents because when an accident occurs, the protection degree is crucial. The KNGR requirements include the existing nuclear power plants competitiveness as well as those of the coal thermal plants. Moreover, when safety is reinforced, the economic competitiveness can be assured. At the present time, a subsequent specification for the KNGR considering the bases of the domestic technology and experimenting the running. (O.M.)

  20. Application of Framework for Integrating Safety, Security and Safeguards (3Ss) into the Design Of Used Nuclear Fuel Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Badwan, Faris M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Demuth, Scott F [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-06

    Department of Energy’s Office of Nuclear Energy, Fuel Cycle Research and Development develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development focused on used nuclear fuel recycling and waste management to meet U.S. needs. Used nuclear fuel is currently stored onsite in either wet pools or in dry storage systems, with disposal envisioned in interim storage facility and, ultimately, in a deep-mined geologic repository. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Integrating safety, security, and safeguards (3Ss) fully in the early stages of the design process for a new nuclear facility has the potential to effectively minimize safety, proliferation, and security risks. The 3Ss integration framework could become the new national and international norm and the standard process for designing future nuclear facilities. The purpose of this report is to develop a framework for integrating the safety, security and safeguards concept into the design of Used Nuclear Fuel Storage Facility (UNFSF). The primary focus is on integration of safeguards and security into the UNFSF based on the existing Nuclear Regulatory Commission (NRC) approach to addressing the safety/security interface (10 CFR 73.58 and Regulatory Guide 5.73) for nuclear power plants. The methodology used for adaptation of the NRC safety/security interface will be used as the basis for development of the safeguards /security interface and later will be used as the basis for development of safety and safeguards interface. Then this will complete the integration cycle of safety, security, and safeguards. The overall methodology for integration of 3Ss will be proposed, but only the integration of safeguards and security will be applied to the design of the