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.

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.

The tendency of medical electrical equipment - IEC 60601-2-54: Particular requirements for the basic safety and essential performance of x-ray equipment for radiography and radioscopy

International Nuclear Information System (INIS)

Roh, Young Hoon; Kim, Jung Min

2015-01-01

Medical electrical equipment - Part 1: General requirement for basic safety and essential performance of MFDS was revised as 3th edition and Medical electrical equipment Part 2-54: Particular requirements for the basic safety and essential performance of X-ray equipment will be expected to be announced as notification. Therefore this technical report was written to introduce provision of the particular requirements, replacement, addition, amendment. The purpose of this particular requirements is to secure requirements for basic safety and essential performance of X-ray equipment for radiography and radioscopy. X-ray high voltage generator, mechanical protective device, protection against radiation is included in this particular requirements. Medical electrical equipment - Part 1, Part 1-2, Part 1-3 is applied to this particular requirements. If the requirements is announced as notification, It is expected to widen understanding for basic safety and essential performance of X-ray equipment for radiography and radioscopy and play a part to internationalize of medical equipment

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

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

A concurrent diagnosis of microbiological food safety output and food safety management system performance: Cases from meat processing industries

NARCIS (Netherlands)

Luning, P.A.; Jacxsens, L.; Rovira, J.; Oses Gomez, S.; Uyttendaele, M.; Marcelis, W.J.

2011-01-01

Stakeholder requirements force companies to analyse their food safety management system (FSMS) performance to improve food safety. Performance is commonly analysed by checking compliance against preset requirements via audits/inspections, or actual food safety (FS) output is analysed by

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

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.

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.

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.

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

International Nuclear Information System (INIS)

2009-01-01

Fundamental Safety Principles. Section 3 describes the graded approach to implementation of the requirements for safety assessment for different facilities and activities. Section 4 establishes the overall requirements for a safety assessment and specific requirements that relate to the assessment of features relevant to safety. Section 4 also establishes the requirements to address defence in depth and safety margins, to perform safety analysis, to document the safety assessment and to carry out an independent verification. Section 5 establishes the requirements for the management, use and maintenance of the safety assessment

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

International Nuclear Information System (INIS)

2009-01-01

Fundamental Safety Principles. Section 3 describes the graded approach to implementation of the requirements for safety assessment for different facilities and activities. Section 4 establishes the overall requirements for a safety assessment and specific requirements that relate to the assessment of features relevant to safety. Section 4 also establishes the requirements to address defence in depth and safety margins, to perform safety analysis, to document the safety assessment and to carry out an independent verification. Section 5 establishes the requirements for the management, use and maintenance of the safety assessment

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

International Nuclear Information System (INIS)

2010-01-01

Fundamental Safety Principles. Section 3 describes the graded approach to implementation of the requirements for safety assessment for different facilities and activities. Section 4 establishes the overall requirements for a safety assessment and specific requirements that relate to the assessment of features relevant to safety. Section 4 also establishes the requirements to address defence in depth and safety margins, to perform safety analysis, to document the safety assessment and to carry out an independent verification. Section 5 establishes the requirements for the management, use and maintenance of the safety assessment

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

International Nuclear Information System (INIS)

2009-01-01

from the Fundamental Safety Principles. Section 3 describes the graded approach to implementation of the requirements for safety assessment for different facilities and activities. Section 4 establishes the overall requirements for a safety assessment and specific requirements that relate to the assessment of features relevant to safety. Section 4 also establishes the requirements to address defence in depth and safety margins, to perform safety analysis, to document the safety assessment and to carry out an independent verification. Section 5 establishes the requirements for the management, use and maintenance of the safety assessment

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

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

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

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

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.

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.

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)

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

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...

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.

2011 Annual Criticality Safety Program Performance Summary

Energy Technology Data Exchange (ETDEWEB)

Andrea Hoffman

2011-12-01

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

Performance Analysis of Multi Stage Safety Injection Tank

International Nuclear Information System (INIS)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo

2015-01-01

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

Performance Analysis of Multi Stage Safety Injection Tank

Energy Technology Data Exchange (ETDEWEB)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

Experience with performance based training of nuclear criticality safety engineers

International Nuclear Information System (INIS)

Taylor, R.G.

1993-01-01

For non-reactor nuclear facilities, the U.S. Department of Energy (DOE) does not require that nuclear criticality safety engineers demonstrate qualification for their job. It is likely, however, that more formalism will be required in the future. Current DOE requirements for those positions which do have to demonstrate qualification indicate that qualification should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis is incompletely developed in some areas

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

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.

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

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.

Performance Testing Methodology for Safety-Critical Programmable Logic Controller

International Nuclear Information System (INIS)

Kim, Chang Ho; Oh, Do Young; Kim, Ji Hyeon; Kim, Sung Ho; Sohn, Se Do

2009-01-01

The Programmable Logic Controller (PLC) for use in Nuclear Power Plant safety-related applications is being developed and tested first time in Korea. This safety-related PLC is being developed with requirements of regulatory guideline and industry standards for safety system. To test that the quality of the developed PLC is sufficient to be used in safety critical system, document review and various product testings were performed over the development documents for S/W, H/W, and V/V. This paper provides the performance testing methodology and its effectiveness for PLC platform conducted by KOPEC

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.

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.

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.

Determination of Safety Performance Grade of NPP Using Integrated Safety Performance Assessment (ISPA) Program

International Nuclear Information System (INIS)

Chung, Dae Wook

2011-01-01

Since the beginning of 2000, the safety regulation of nuclear power plant (NPP) has been challenged to be conducted more reasonable, effective and efficient way using risk and performance information. In the United States, USNRC established Reactor Oversight Process (ROP) in 2000 for improving the effectiveness of safety regulation of operating NPPs. The main idea of ROP is to classify the NPPs into 5 categories based on the results of safety performance assessment and to conduct graded regulatory programs according to categorization, which might be interpreted as 'Graded Regulation'. However, the classification of safety performance categories is highly comprehensive and sensitive process so that safety performance assessment program should be prepared in integrated, objective and quantitative manner. Furthermore, the results of assessment should characterize and categorize the actual level of safety performance of specific NPP, integrating all the substantial elements for assessing the safety performance. In consideration of particular regulatory environment in Korea, the integrated safety performance assessment (ISPA) program is being under development for the use in the determination of safety performance grade (SPG) of a NPP. The ISPA program consists of 6 individual assessment programs (4 quantitative and 2 qualitative) which cover the overall safety performance of NPP. Some of the assessment programs which are already implemented are used directly or modified for incorporating risk aspects. The others which are not existing regulatory programs are newly developed. Eventually, all the assessment results from individual assessment programs are produced and integrated to determine the safety performance grade of a specific NPP

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.)

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.

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)

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.

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

Development of Safety Grade PLC (POSAFE-Q) and Performance Test Results

International Nuclear Information System (INIS)

Kim, Chang Hwoi; Park, Won Man; Choi, Jong Gyun; Lee, Dong Young; No, Young Hun; Song, Seung Hwan

2006-01-01

The safety grade PLC (POSAFE-Q) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as the RTOS and firmware are being developed according to the safety critical software life cycle. Especially, the formal method is applied to design the SRS (Software Requirement Spec.) and the SDS (Software Design Specification.) to be error-free. The POSAFE-Q has several modules such as processor module, input and output modules, communication modules, redundant processor module, redundant power modules, etc,. To verify the function and performance, several tests such as CT, IT and ST were performed. And also, the equipment qualification test for environment, EMI and EMC, and seismic ware performed. All tests are satisfied with the requirements and specification for safety grade PLC, and the criteria for safety system in nuclear power plants

Development of Safety Grade PLC (POSAFE-Q) and Performance Test Results

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Hwoi; Park, Won Man; Choi, Jong Gyun; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); No, Young Hun; Song, Seung Hwan [POSCON, Seoul (Korea, Republic of)

2006-07-01

The safety grade PLC (POSAFE-Q) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. The PLC satisfies Safety Class 1E, Quality Class 1, and Seismic Category I. The software such as the RTOS and firmware are being developed according to the safety critical software life cycle. Especially, the formal method is applied to design the SRS (Software Requirement Spec.) and the SDS (Software Design Specification.) to be error-free. The POSAFE-Q has several modules such as processor module, input and output modules, communication modules, redundant processor module, redundant power modules, etc,. To verify the function and performance, several tests such as CT, IT and ST were performed. And also, the equipment qualification test for environment, EMI and EMC, and seismic ware performed. All tests are satisfied with the requirements and specification for safety grade PLC, and the criteria for safety system in nuclear power plants.

Safety KPIs - Monitoring of safety performance

Directory of Open Access Journals (Sweden)

Andrej Lališ

2014-09-01

Full Text Available This paper aims to provide brief overview of aviation safety development focusing on modern trends represented by implementation of Safety Key Performance Indicators. Even though aviation is perceived as safe means of transport, it is still struggling with its complexity given by long-term growth and robustness which it has reached today. Thus nowadays safety issues are much more complex and harder to handle than ever before. We are more and more concerned about organizational factors and control mechanisms which have potential to further increase level of aviation safety. Within this paper we will not only introduce the concept of Key Performance Indicators in area of aviation safety as an efficient control mechanism, but also analyse available legislation and documentation. Finally we will propose complex set of indicators which could be applied to Czech Air Navigation Service Provider.

A road safety performance indicator for vehicle fleet compatibility.

Science.gov (United States)

Christoph, Michiel; Vis, Martijn Alexander; Rackliff, Lucy; Stipdonk, Henk

2013-11-01

This paper discusses the development and the application of a safety performance indicator which measures the intrinsic safety of a country's vehicle fleet related to fleet composition. The indicator takes into account both the 'relative severity' of individual collisions between different vehicle types, and the share of those vehicle types within a country's fleet. The relative severity is a measure for the personal damage that can be expected from a collision between two vehicles of any type, relative to that of a collision between passenger cars. It is shown how this number can be calculated using vehicle mass only. A sensitivity analysis is performed to study the dependence of the indicator on parameter values and basic assumptions made. The indicator is easy to apply and satisfies the requirements for appropriate safety performance indicators. It was developed in such a way that it specifically scores the intrinsic safety of a fleet due to its composition, without being influenced by other factors, like helmet wearing. For the sake of simplicity, and since the required data is available throughout Europe, the indicator was applied to the relative share of three of the main vehicle types: passenger cars, heavy goods vehicles and motorcycles. Using the vehicle fleet data from 13EU Member States and Norway, the indicator was used to rank the countries' safety performance. The UK was found to perform best in terms of its fleet composition (value is 1.07), while Greece has the worst performance with the highest indicator value (1.41). Copyright © 2013 Elsevier Ltd. All rights reserved.

Safety performance indicators program

International Nuclear Information System (INIS)

Vidal, Patricia G.

2004-01-01

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

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

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)

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

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

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

Experience with performance based training of nuclear criticality safety engineers

International Nuclear Information System (INIS)

Taylor, R.G.

1993-01-01

Historically, new entrants to the practice of nuclear criticality safety have learned their job primarily by on-the-job training (OJT) often by association with an experienced nuclear criticality safety engineer who probably also learned their job by OJT. Typically, the new entrant learned what he/she needed to know to solve a particular problem and accumulated experience as more problems were solved. It is likely that more formalism will be required in the future. Current US Department of Energy requirements for those positions which have to demonstrate qualification indicate that it should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis i's incompletely developed in some areas. Details of this analysis are provided in this report

Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Baji, C.; Vamos, G.; Toth, J.

2001-01-01

The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

Radiation legacy of nuclear tests at the Semipalatinsk test site in the light of requirements ensuring radiation safety performance

International Nuclear Information System (INIS)

Logachev, V.A.; Logacheva, L.A.

2005-01-01

Peculiarities of nuclear tests radiation legacy at the Semipalatinsk test site (STS) are shown in the light of performance of requirements ensuring radiation safety, decrease radiation contamination levels in environment and minimize exposure of radiation for population residing contaminated areas by radioactive fallout. The paper provides data on characterization of peculiarities of the STS operation legacy based on review of archival data of the former 3-d General Administration under USSR Ministry of Health. (author)

Potential safety features and safety analysis aspects for high performance light water reactor (HPLWR)

International Nuclear Information System (INIS)

Aksan, N.; Schulenberg, T.; Squarer, D.

2003-01-01

Research Activities are ongoing worldwide to develop advanced nuclear power plants with high thermal efficiency for the purpose to improve their economical competitiveness. Within the 5th Framework Programme of the European Commission, a project has been launched with the main objective to assess the technical and economical feasibility of a high efficiency LWR operating at super critical pressure conditions. Several European research institutions, industrial partners and the University of Tokyo participated and worked in this common research project. Within the aims of the development of the HPLWR is to use both passive and active safety systems for performing safety related functions in the event of transients or accidents. Consequently substantial effort has been invested in order to define the safety features of the plant in a European environment, as well as to incorporate passive safety features into the design. Throughout this process, the European Utility Requirements (EUR) and requirements known from Generation IV initiative were considered as a guideline in general terms in order to include further advanced ideas. The HPLWR general features were compared to both requirements, indicating a potential to meet these. Since, the supercritical HPLWR represents a challenge for best-estimate safety codes like RELAP5, CATHARE and TRAB due to the fact that these codes were developed for two-phase or single-phase coolant at pressures far below critical point, work on the preliminary assessment of the appropriateness of these codes have been performed for selected relevant phenomena, and application of the codes to the selected transients on the basis of defined 'reference design'. An overview on their successful upgrade to supercritical pressures and application to some plant safety analysis are provided in the paper. Further elaborations in relation to future needs are also discussed. (author)

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)

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

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

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

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

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

International Nuclear Information System (INIS)

Dey, M.

1993-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-09-01

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

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)

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.

Food safety performance indicators to benchmark food safety output of food safety management systems.

Science.gov (United States)

Jacxsens, L; Uyttendaele, M; Devlieghere, F; Rovira, J; Gomez, S Oses; Luning, P A

2010-07-31

There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses. Validation was conducted on the basis of an extensive microbiological assessment scheme (MAS). The assumption behind the food safety performance diagnosis is that food businesses which evaluate the performance of their food safety management system in a more structured way and according to very strict and specific criteria will have a better insight in their actual microbiological food safety performance, because food safety problems will be more systematically detected. The diagnosis can be a useful tool to have a first indication about the microbiological performance of a food safety management system present in a food business. Moreover, the diagnosis can be used in quantitative studies to get insight in the effect of interventions on sector or governmental level. Copyright 2010 Elsevier B.V. All rights reserved.

Is road safety management linked to road safety performance?

Science.gov (United States)

Papadimitriou, Eleonora; Yannis, George

2013-10-01

This research aims to explore the relationship between road safety management and road safety performance at country level. For that purpose, an appropriate theoretical framework is selected, namely the 'SUNflower' pyramid, which describes road safety management systems in terms of a five-level hierarchy: (i) structure and culture, (ii) programmes and measures, (iii) 'intermediate' outcomes'--safety performance indicators (SPIs), (iv) final outcomes--fatalities and injuries, and (v) social costs. For each layer of the pyramid, a composite indicator is implemented, on the basis of data for 30 European countries. Especially as regards road safety management indicators, these are estimated on the basis of Categorical Principal Component Analysis upon the responses of a dedicated road safety management questionnaire, jointly created and dispatched by the ETSC/PIN group and the 'DaCoTA' research project. Then, quasi-Poisson models and Beta regression models are developed for linking road safety management indicators and other indicators (i.e. background characteristics, SPIs) with road safety performance. In this context, different indicators of road safety performance are explored: mortality and fatality rates, percentage reduction in fatalities over a given period, a composite indicator of road safety final outcomes, and a composite indicator of 'intermediate' outcomes (SPIs). The results of the analyses suggest that road safety management can be described on the basis of three composite indicators: "vision and strategy", "budget, evaluation and reporting", and "measurement of road user attitudes and behaviours". Moreover, no direct statistical relationship could be established between road safety management indicators and final outcomes. However, a statistical relationship was found between road safety management and 'intermediate' outcomes, which were in turn found to affect 'final' outcomes, confirming the SUNflower approach on the consecutive effect of each layer

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.

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

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...

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

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)

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

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.

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

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.)

Performance Evaluation of SMART Passive Safety System for Small Break LOCA Using MARS Code

International Nuclear Information System (INIS)

Chun, Ji Han; Lee, Guy Hyung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

2013-01-01

SMART has significantly enhanced safety by reducing its core damage frequency to 1/10 that of a conventional nuclear power plant. KAERI is developing a passive safety injection system to replace the active safety injection pump in SMART. It consists of four trains, each of which includes gravity-driven core makeup tank (CMT) and safety injection tank (SIT). This system is required to meet the passive safety performance requirements, i.e., the capability to maintain a safe shutdown condition for a minimum of 72 hours without an AC power supply or operator action in the case of design basis accidents (DBAs). The CMT isolation valve is opened by the low pressurizer pressure signal, and the SIT isolation valve is opened at 2 MPa. Additionally, two stages of automatic depressurization systems are used for rapid depressurization. Preliminary safety analysis of SMART passive safety system in the event of a small-break loss-of-coolant accident (SBLOCA) was performed using MARS code. In this study, the safety analysis results of a guillotine break of safety injection line which was identified as the limiting SBLOCA in SMART are given. The preliminary safety analysis of a SBLOCA for the SMART passive safety system was performed using the MARS code. The analysis results of the most limiting SI line guillotine break showed that the collapsed liquid level inside the core support barrel was maintained sufficiently high above the top of core throughout the transient. This means that the passive safety injection flow from the CMT and SIT causes no core uncovery during the 72 hours following the break with no AC power supply or operator action, which in turn results in a consistent decrease in the fuel cladding temperature. Therefore, the SMART passive safety system can meet the passive safety performance requirement of maintaining the plant at a safe shutdown condition for a minimum of 72 hours without AC power or operator action for a representing accident of SBLOCA

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-01-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of performance, safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith S.G.H.W.R. that the necessary bias toward safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications and has to be paid for. By a detailed evaluation of S.G.H.W.R. fuel defects it is shown that very few defects can be shown to be related to design, rating or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that must be met to achieve the desirable reliability level. It is possible that large-scale experience with S.G.H.W.R. fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to S.G.H.W.R. fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. The philosophy and supporting experimental work programme are outlined and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to new 60 pin fuel element to be used in the commercial S.G.H.W.R.'s and how it compares in design and performance aspects with the 36 pin element that has been used to date in the Winfrith S.G.H.W.R

Health and safety plan for operations performed for the Environmental Restoration Program

Energy Technology Data Exchange (ETDEWEB)

Trippet, W.A. II (IT Corp., (United States)); Reneau, M.; Morton, S.L. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

1992-04-01

This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the EPR. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

Health and safety plan for operations performed for the Environmental Restoration Program

International Nuclear Information System (INIS)

Trippet, W.A. II; Reneau, M.; Morton, S.L.

1992-04-01

This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG ampersand G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the EPR. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP

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

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...

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.

Development of safety performance indicators for HANARO

International Nuclear Information System (INIS)

Wu, Jong-Sup; Jung, Hoan-Sung; Ahn, Guk-Hoon; Lee, Kye-Hong; Lim, In-Cheol

2007-01-01

The nuclear facilities need an extensive basis for ensuring their safety. An operating organization should conduct its operation and utilization important to the safety in accordance with approved procedures and regulations. The general aims of a management system for nuclear facilities are to improve the safety performance through a planning, control and supervision of safety related activities and to foster a strong safety culture. The effectiveness of a management system can be monitored and measured to confirm the ability of its processes to achieve the intended safety performance by an assessment of the operational performance. The Operational Safety Performance Indicators, also known as SPI, help an organization define and measure a progress with regard to safety activity goals. The elements of a SPI are quantifiable measurements that reflect the critical success factors of an organizational safety. Since 1995, efforts have been directed towards the elaboration of a framework for the establishment of an operational safety performance indicator program in nuclear power plants (NPP). IAEA-TECDOC-1141, 'Operational safety performance indicators for NPP' attempted to provide a frame work for an identification of performance indicators which have a relationship to the desired safety attributes, and therefore, to a safe plant operation. Three key attributes of a smooth operation, an operation with a low risk, and an operation with a positive safety attitude, were recommended, which are associated with a safe operation. Because these attributes cannot be directly measured, an indicator structure is expanded further until a level of easily quantifiable or directly measurable indicators is identified. The intention of this approach is to use quantitative information provided by the specific indicators and to analyze performance trends relative to established goals. The safety activities in HANARO have been continuously conducted to enhance its safe operation. HANARO

42 CFR 84.103 - Man tests; performance requirements.

Science.gov (United States)

2010-10-01

... 42 Public Health 1 2010-10-01 2010-10-01 false Man tests; performance requirements. 84.103 Section 84.103 Public Health PUBLIC HEALTH SERVICE, DEPARTMENT OF HEALTH AND HUMAN SERVICES OCCUPATIONAL SAFETY AND HEALTH RESEARCH AND RELATED ACTIVITIES APPROVAL OF RESPIRATORY PROTECTIVE DEVICES Self-Contained Breathing Apparatus § 84.103 Man tests;...

Critical review of safety performance metrics

NARCIS (Netherlands)

Karanikas, Nektarios

2016-01-01

Various tools for safety performance measurement have been introduced in order to fulfil the need for safety monitoring in organisations, which is tightly related to their overall performance and achievement of their business goals. Such tools include accident rates, benchmarking, safety culture and

Regulatory activities in the area of fuel safety and performance

International Nuclear Information System (INIS)

Viktorov, A.; Couture, M.

2005-01-01

Generic Action Item 94G02 'Impact of Fuel Bundle Condition on Reactor Safety' in many ways determined the present priorities in regulatory activities related to fuel performance. As one of the closure criteria it required that all licensees establish 'an effective formal and systematic process for integrating fuel design, fuel and channel inspection, laboratory examination, research, operating limits and safety analysis'. To date, such a process has been, to a large extent, put in place by all licensees. To assure that such processes remain operational and effective after the GAI closure, CNSC required, through S-99, to report annually on fuel performance and major activities in the fuel safety area. The scope of reported information has been defined to allow CNSC staff evaluation of key events and trends in fuel performance. To compliment reporting by the industry, CNSC staff has conducted targeted inspections of fuel compliance programs at all sites. Combined together, these activities provide the regulator with the confidence that CANDU fuel is robust and operates with safety margins. The scrutiny, to which fuel performance has been subjected lately, has allowed identification of certain programmatic weaknesses and gaps in the knowledge concerning the fuel behaviour under various conditions. It has become apparent that top-level strategies for assessment of fuel performance may have been inadequate and far from systematic; fuel inspection practices and capabilities have varied significantly from site to site; certain issues were identified but remained unaddressed for significant time; priorities in experimental or design support activities were not assigned consistently. The presentation gives examples of areas where, in the opinion of the CNSC staff, further work is required to support fuel design and safety envelopes. The implementation of new CANFLEX fuel designs is currently being considered by the industry and CNSC staff has been engaged in the review

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.

Analysis on Occupants’ Satisfaction for Safety Performance Assessment in Low Cost Housing

Directory of Open Access Journals (Sweden)

Husin Husrul Nizam

2014-01-01

Full Text Available The delivery performance of the low cost housing is questioned since the occupants are prone towards safety hazards in the housing complex, such as structural instability and falling building fragments. Without defining the occupants’ requirements for the development of low cost housing, the prevailing safety factors are hard to be determined. This paper explores the rationale of safety performance assessment in the low cost housing by considering the occupants’ participation to achieve a better safety provision during occupancy period. Questionnaire survey was distributed to 380 occupants of the low cost housing in Kuala Lumpur and Selangor, Malaysia. The result shows that 80.8% of the respondents had expressed their dissatisfaction with the safety performance of the lift. By referring to the mode of ranking level, the most significant aspect rated by the respondents is Building Safety Features, with 51.6% respondents. The attained aspects can be fundamental parameters which can be considered in the future development of low cost housing.

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.

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)

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.

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

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...

Food safety performance indicators to benchmark food safety output of food safety management systems

NARCIS (Netherlands)

Jacxsens, L.; Uyttendaele, M.; Devlieghere, F.; Rovira, J.; Oses Gomez, S.; Luning, P.A.

2010-01-01

There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses.

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.

Indicators to monitor NPP operational safety performance

International Nuclear Information System (INIS)

Gomez-Cobo, Ana

2002-01-01

Since December 1995 the IAEA activities on safety performance indicators focused on the elaboration of a framework for the establishment of an operational safety performance indicator programme. The development of this framework began with the consideration of the concept of NPP operational safety performance and the identification of operational safety attributes. For each operational safety attribute, overall indicators, envisioned as providing an overall evaluation of relevant aspects of safety performance, were established. Associated with each overall indicator is a level of strategic indicators intended to provide a bridge from overall to specific indicators. Finally each strategic indicator was supported by a set of specific indicators, which represent quantifiable measures of performance. The programme development was enhanced by pilot plant studies, conducted over a 15 month period from January 1998 to March 1999. The result of all this work is compiled in the IAEA-TECDOC-1141, to be published shortly. This paper presents a summary of this IAEA TECDOC. It describes the operational safety performance indicator framework proposed and discusses the results of and lessons learned from the pilot studies. Despite the efforts described, it is clear that additional research is still necessary in areas such as plant-specific adaptation of proposed frameworks in order to suit individual data collection systems and plant characteristics, indicator selection, indicator definition, goal setting, action thresholds, analysis of trends, indicator display systems, analysis of overall safety performance (i.e., aggregation or combination of indicators), safety culture indicators, qualitative indicators, and use of additional indicators to address issues such as industrial safety attitude and performance, staff welfare, and environmental compliance. This is the rationale for a new IAEA Coordinated Research Project on 'Development and application of indicators to monitor NPP

Benchmarking road safety performance: Identifying a meaningful reference (best-in-class).

Science.gov (United States)

Chen, Faan; Wu, Jiaorong; Chen, Xiaohong; Wang, Jianjun; Wang, Di

2016-01-01

For road safety improvement, comparing and benchmarking performance are widely advocated as the emerging and preferred approaches. However, there is currently no universally agreed upon approach for the process of road safety benchmarking, and performing the practice successfully is by no means easy. This is especially true for the two core activities of which: (1) developing a set of road safety performance indicators (SPIs) and combining them into a composite index; and (2) identifying a meaningful reference (best-in-class), one which has already obtained outstanding road safety practices. To this end, a scientific technique that can combine the multi-dimensional safety performance indicators (SPIs) into an overall index, and subsequently can identify the 'best-in-class' is urgently required. In this paper, the Entropy-embedded RSR (Rank-sum ratio), an innovative, scientific and systematic methodology is investigated with the aim of conducting the above two core tasks in an integrative and concise procedure, more specifically in a 'one-stop' way. Using a combination of results from other methods (e.g. the SUNflower approach) and other measures (e.g. Human Development Index) as a relevant reference, a given set of European countries are robustly ranked and grouped into several classes based on the composite Road Safety Index. Within each class the 'best-in-class' is then identified. By benchmarking road safety performance, the results serve to promote best practice, encourage the adoption of successful road safety strategies and measures and, more importantly, inspire the kind of political leadership needed to create a road transport system that maximizes safety. Copyright © 2015 Elsevier Ltd. All rights reserved.

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)

Safety performance indicators for the road network.

NARCIS (Netherlands)

Weijermars, W. Gitelman, V. Papadimitriou, E. Lima De & Azevedo, C.

2010-01-01

Within the 6th FP European project SafetyNet, a team has worked on the development of Safety Performance Indicators (SPIs) on seven road safety related areas. These SPIs reflect the operational conditions of the road traffic system that influence the system's safety performance. SPIs were developed

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

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

Pickman, D.O.; Inglis, G.H.

1977-05-01

The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith SGHWR that the necessary bias towards safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications. By a detailed evaluation of SGHWR fuel defects it is shown that very few defects can be shown to be related to design, rating, or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that most be met to achieve the desirable reliability level. It is possible that large scale experience on SGHWR fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to SGHWR fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. the philosophy and supporting experimental work programme are outlines and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to the new 60-pin fuel element to be used in the commercial SGHWRs and to its comparison in design and performance aspects with the 36-pin element that has been used to date in the Winfrith SGHWR. (author)

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

Nursing home safety: does financial performance matter?

Science.gov (United States)

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

2011-01-01

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

Prospective safety performance evaluation on construction sites.

Science.gov (United States)

Wu, Xianguo; Liu, Qian; Zhang, Limao; Skibniewski, Miroslaw J; Wang, Yanhong

2015-05-01

This paper presents a systematic Structural Equation Modeling (SEM) based approach for Prospective Safety Performance Evaluation (PSPE) on construction sites, with causal relationships and interactions between enablers and the goals of PSPE taken into account. According to a sample of 450 valid questionnaire surveys from 30 Chinese construction enterprises, a SEM model with 26 items included for PSPE in the context of Chinese construction industry is established and then verified through the goodness-of-fit test. Three typical types of construction enterprises, namely the state-owned enterprise, private enterprise and Sino-foreign joint venture, are selected as samples to measure the level of safety performance given the enterprise scale, ownership and business strategy are different. Results provide a full understanding of safety performance practice in the construction industry, and indicate that the level of overall safety performance situation on working sites is rated at least a level of III (Fair) or above. This phenomenon can be explained that the construction industry has gradually matured with the norms, and construction enterprises should improve the level of safety performance as not to be eliminated from the government-led construction industry. The differences existing in the safety performance practice regarding different construction enterprise categories are compared and analyzed according to evaluation results. This research provides insights into cause-effect relationships among safety performance factors and goals, which, in turn, can facilitate the improvement of high safety performance in the construction industry. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Development of safety performance indicators of regulatory interest (SAFPER) in Pakistan

International Nuclear Information System (INIS)

Khatoon, Abida

2002-01-01

enhanced only if meaningful goals and targets are established); - analysis of overall plant performance; - safety culture indicator (qualitative indicator). Some of the indicators, like Sudden outages, unavailability of Safety Systems, Collective Radiation Exposure, Station Effluents, are analysed under Operating Performance parameter of KANUPP Report for the period 1994-1998. As observed from the findings by DNSRP inspectors and corrective actions taken by KANUPP during regulatory inspections and the trend analysis of the safety performance indicators for Workers Radiation Safety it can be concluded that the Regulatory Body in Pakistan is performing its responsibility in an effective manner with the co-operation of the Management of the Utility. It is concluded that the general trend observed during this study is expected from an old plant like KANUPP. The encouraging aspect which matters for the Regulators is the trend observed for the radiation dose internal to external ratio for the period 1994 to 1998. This ratio has started decreasing now after doing some maintenance work. Another important parameter of safety concern is the number of forced outages which have been decreasing from 1994 to 1998. Kanupp is facing operational problems mainly due to its Ageing and some other maintenance activities. These conditions can cause safety concerns that, if not appropriately addressed, would require the licensee to shut down the plant. This deficiency will be rectified to a large extent when the effort to produce a revised full scope PSAR is complete, and a new license issued. (author)

Reviewing real-time performance of nuclear reactor safety systems

International Nuclear Information System (INIS)

Preckshot, G.G.

1993-08-01

The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems

Reviewing real-time performance of nuclear reactor safety systems

Energy Technology Data Exchange (ETDEWEB)

Preckshot, G.G. [Lawrence Livermore National Lab., CA (United States)

1993-08-01

The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

Evaluation of operating experience for early recognition of deteriorating safety performance

International Nuclear Information System (INIS)

Beckmerhagen, I.A.; Berg, H.P.

2004-01-01

One of the most difficult challenges facing nuclear power plants is to recognize the early signs of degrading safety performance before regulatory requirements are imposed or serious incidents or accidents occur. Today, the nuclear industry is striving for collecting more information on occurrences that could improve the operational safety performance. To achieve this, the reporting threshold has been lowered from incidents to anomalies with minor or no impact to safety. Industry experience (also outside nuclear industry) has shown that these are typical issues which should be considered when looking for such early warning signs. Therefore, it is important that nuclear power plant operators have the capability to trend, analyse and recognize early warning signs of deteriorating performance. It is necessary that plant operators are sensitive to these warning signs which may not be immediately evident. Reviewing operating experience is one of the main tasks for plant operators in their daily activities. Therefore, self assessment should be at the centre of any operational safety performance programme. One way of applying a self assessment program is through the following four basic elements: operational data, events, safety basis, and related experience. This approach will be described in the paper in more details. (authors)

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

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.

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.

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

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

Development of safety performance indicators in Japan

International Nuclear Information System (INIS)

Ohashi, H.; Tamao, S.; Tanaka, J.; Sawayama, T.

2001-01-01

For the purpose of safety regulations of operating nuclear power stations in Japan, the regulatory authorities utilize two types of regulations. One is the direct regulation, such as periodical inspection to inspect the function and performance of equipment important to safety, and the other is the audit type regulation such as preservation inspection to audit the compliance with the safety preservation rules. As performance indicators are expected to be an effective tool to evaluate the activities by audit type regulations, NUPEC is studying a comprehensive set of operational performance indicators to meet the effective evaluation method for the safety preservation activities in the audit type regulations under the frame of current safety regulation system. The study includes the establishment of comprehensive operational performance indicators applicable in Japan, the effective application of performance indicators to the current Japanese regulation, the clarification of the applicable scope of utilization, the possibility of applying the performance indicators. This report describes the present status of our performance indicator studies. After the completion of these studies the regulatory authorities will evaluate if and how the new set of comprehensive performance indicators could be introduced to Japanese regulatory scheme. (author)

Safety performance indicators. Topical issues paper no. 5

International Nuclear Information System (INIS)

Dahlgren, K.; Lederman, L.; Szikszai, T.; Palomo, J.

2001-01-01

Since its creation the nuclear industry has been struggling with the question of how safe is safe enough. Safety is a common goal to all involved in the design, operation and regulation of a nuclear installation. As a concept safety is not easy to define. However, there is a general understanding of what attributes a nuclear power plant should have in order to operate safely. The challenge lies in measuring the attributes. The new competitive open electricity market, in many countries throughout the world, is increasing the economic pressure on operators to lower operating costs without jeopardizing safety. Challenges are occurring at a rate that is unprecedented in the nuclear industry: competitiveness; downsizing; ageing; policy changes; reorganization; restructuring; mergers; globalization; and takeovers demand increasing attention to the management of safety. There are various means to measure safety performance, some of which are more qualitative in nature and others which through quantitative measures provide the means by which to evaluate performance trends with clear ties to safety. According to their use, indicators are generally considered in two groups: leading or proactive and lagging or reactive indicators. Leading indicators are most useful as a precursor to safety degradation for early management reaction. Lagging indicators are most commonly used to drive plant performance, to monitor and for benchmarking against similar plants. The actual values of the indicators are not intended to be direct measures of safety, although safety performance can be inferred from the results achieved. Each plant needs to determine which indicators best serve its needs. Selected indicators should not be static, but should be adapted to the conditions and performance of the plant, considering the cost-benefit of maintaining each individual indicator. It should be recognized that while indicators provide valuable information in the effective management of plant safety

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

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

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)

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

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

Road safety performance indicators for the interurban road network.

Science.gov (United States)

Yannis, George; Weijermars, Wendy; Gitelman, Victoria; Vis, Martijn; Chaziris, Antonis; Papadimitriou, Eleonora; Azevedo, Carlos Lima

2013-11-01

Various road safety performance indicators (SPIs) have been proposed for different road safety research areas, mainly as regards driver behaviour (e.g. seat belt use, alcohol, drugs, etc.) and vehicles (e.g. passive safety); however, no SPIs for the road network and design have been developed. The objective of this research is the development of an SPI for the road network, to be used as a benchmark for cross-region comparisons. The developed SPI essentially makes a comparison of the existing road network to the theoretically required one, defined as one which meets some minimum requirements with respect to road safety. This paper presents a theoretical concept for the determination of this SPI as well as a translation of this theory into a practical method. Also, the method is applied in a number of pilot countries namely the Netherlands, Portugal, Greece and Israel. The results show that the SPI could be efficiently calculated in all countries, despite some differences in the data sources. In general, the calculated overall SPI scores were realistic and ranged from 81 to 94%, with the exception of Greece where the SPI was relatively lower (67%). However, the SPI should be considered as a first attempt to determine the safety level of the road network. The proposed method has some limitations and could be further improved. The paper presents directions for further research to further develop the SPI. Copyright © 2012 Elsevier Ltd. All rights reserved.

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

Nuclear power performance and safety. V.3. Safety and international co-operation

International Nuclear Information System (INIS)

1988-01-01

The International Conference on Nuclear Power Performance and Safety, organized by the International Atomic Energy Agency, was held at the Austria Centre Vienna (ACV) in Vienna, Austria, from 28 September to 2 October 1987. The objective of the Conference was to promote an exchange of worldwide information on the current trends in the performance and safety of nuclear power and its fuel cycle, and to take a forward look at the expectations and objectives for the 1990s. This objective was accomplished through presentation and discussion of about 200 papers at the Conference. Almost 500 participants and observers from 40 countries and 12 organizations discussed three major questions which were posed as the focus of this Conference: (1) What are the current trends and major issues with regard to performance and safety of nuclear power, the nuclear fuel cycle and radioactive waste management? (2) What steps are being taken or need to be taken to resolve outstanding issues in order to improve the performance of nuclear power with assured safety? (3) What performance objectives and achievements can be anticipated for the 1990s? All presentations of this Conference were divided into six volumes. This is Volume 3 which is devoted to the problems of safety and international cooperation. All presentations of Volume 3 were divided into four sessions as follows: the need for safety in nuclear power programmes (4 papers); international cooperation in nuclear safety (6 papers); technical aspects in plant safety (7 papers); approaches to safety (3 papers). A separate abstract was prepared for each of these 20 papers. Refs, figs and tabs

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

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

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)

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)

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.

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

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.

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.

Safety analysis report for packaging (onsite) transuranic performance demonstration program sample packaging

International Nuclear Information System (INIS)

Mccoy, J.C.

1997-01-01

The Transuranic Performance Demonstration Program (TPDP) sample packaging is used to transport highway route controlled quantities of weapons grade (WG) plutonium samples from the Plutonium Finishing Plant (PFP) to the Waste Receiving and Processing (WRAP) facility and back. The purpose of these shipments is to test the nondestructive assay equipment in the WRAP facility as part of the Nondestructive Waste Assay PDP. The PDP is part of the U. S. Department of Energy (DOE) National TRU Program managed by the U. S. Department of Energy, Carlsbad Area Office, Carlsbad, New Mexico. Details of this program are found in CAO-94-1045, Performance Demonstration Program Plan for Nondestructive Assay for the TRU Waste Characterization Program (CAO 1994); INEL-96/0129, Design of Benign Matrix Drums for the Non-Destructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996a); and INEL-96/0245, Design of Phase 1 Radioactive Working Reference Materials for the Nondestructive Assay Performance Demonstration Program for the National TRU Program (INEL 1996b). Other program documentation is maintained by the national TRU program and each DOE site participating in the program. This safety analysis report for packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the TRU PDP sample packaging meets the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for an onsite Transportation Hazard Indicator (THI) 2 packaging. This SARP, however, does not include evaluation of any operations within the PFP or WRAP facilities, including handling, maintenance, storage, or operating requirements, except as they apply directly to transportation between the gate of PFP and the gate of the WRAP facility. All other activities are subject to the requirements of the facility safety analysis reports (FSAR) of the PFP or WRAP facility and requirements of the PDP

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)

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).

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.

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.

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

Combined leadless pacemaker and subcutaneous implantable defibrillator therapy: feasibility, safety, and performance

NARCIS (Netherlands)

Tjong, F. V. Y.; Brouwer, T. F.; Smeding, L.; Kooiman, K. M.; de Groot, J. R.; Ligon, D.; Sanghera, R.; Schalij, M. J.; Wilde, A. A. M.; Knops, R. E.

2016-01-01

The subcutaneous implantable cardioverter-defibrillator (S-ICD) and leadless pacemaker (LP) are evolving technologies that do not require intracardiac leads. However, interactions between these two devices are unexplored. We investigated the feasibility, safety, and performance of combined LP and

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

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

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

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

Safety research needs for Russian-designed reactors. Requirements situation

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

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

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...

Performance and safety analysis of WP-cave concept

International Nuclear Information System (INIS)

Skagius, K.; Svemar, C.

1989-08-01

The report presents a performance safety, and cost analysis of the WP-cave, WPC, concept. In the performance analysis, questions specific to the WPC have been addressed which have been identified to require more detailed studies. Based on the outcome of this analysis, a safety analysis has been made which comprises of the modeling and calculation of radionuclide transport from the repository to the biosphere and the resulting dose exposure to man. The result of the safety analysis indicates that the present design of a WPC repository may give unacceptably high doses. By improving the properties of the bentonite/sand barrier such that the hydraulic conductivity is reduced, or by changing the short-lived steel canisters to more long-lived canisters, e.g. copper canisters, it is judged possible to achieve a sufficiently low level of dose exposure rates to man. The cost for a WPC repository of the studied design is significantly higher than for a KBS-3 repository considering the Swedish conditions and the Swedish amount of spent fuel. The major costs are connected to the excavation and backfilling of the bentonite/sand barrier. The potential for cost savings is high but it is not judged possible to account for savings in such a way that the WPC concept shows lower cost than the KBS-3 concept. (34 figs., 33 tabs., 29 refs.)

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

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...

Experience of Tecnatom in Developing a Strong Leadership for Safety and Performance

International Nuclear Information System (INIS)

González, F.; Villadóniga, J. I.

2016-01-01

This paper presents experience and insights of Tecnatom in the support of internal and external clients to develop a strong Leadership for Safety. Several cases are presented briefly: (a) The leadership and culture change activities for a utility, a radwaste company, and for Tecnatom itself. One important characteristic of the work performed is the detailed consideration of the underlying organizational culture that underpins the safety culture. Measurable improvements have been achieved and some of the key insights are shared in this paper. (b) The development and implementation of a leadership model with 17 competencies, including safety explicitly. One benefit of this model is that allows to perform a quantitative assessment of leadership effectiveness, something vital to be able to ensure that leadership development actions are truly supporting safety. The model uses an approach to development oriented to strengths and the use of companion competencies to further develop leadership. Moreover it aims to produce significant improvements on safety but also on performance, since both are not competing goals when the proper leadership model is selected. The training material prepared was shortlisted in the 2014 Nuclear Training Awards. (c) The design and implementation of a training development program on Safety Culture, and required competencies of Leadership, for Top Managers of the nuclear industry, as part of the project NUSHARE of the European Commission’s 7th research framework program. The program is sensible to the reduced time availability of Top Managers and uses a combination of learning approaches (webinars, micro-elearnings, web meetings) that provide higher flexibility for the learner, but complemented with other proven methods (group dialog, journaling, mentoring, etc.) to ensure that the program is effective. All these experiences reveal that to improve the organizational Safety Culture we need to enhance Leadership for Safety and Performance

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

OpenAIRE

Salim Keffane (s)

2014-01-01

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

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.)

Performance of dynamic safety barriers-Structuring, modelling and visualization

OpenAIRE

Wikdahl, Olga

2014-01-01

The main objective of this master thesis is to discuss performance of dynamic safety barriers. A comprehensive literature review is performed in order to get understanding what dynamic safety barrier is. Three different concepts of dynamic safety barriers based on various meanings of dynamic were derived from the literature review: - dynamic safety barriers related to motion or physical force - dynamic safety barriers as updated barriers from dynamic risk analysis - dynamic safety ...

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

Electrochemical performance and safety features of high-safety lithium ion battery using novel branched additive for internal short protection

International Nuclear Information System (INIS)

Li Yuhan; Lee, Meng-Lun; Wang Fuming; Yang, Chang-Rung; Chu, Peter P.J.; Yau, Shueh-Lin; Pan, Jing-Pin

2012-01-01

Highlights: ► N-phenylmaleimide-containing branched oligomer has been employed as an additive in lithium cells. ► The branched oligomer additive enhances safety and cycling performance of Li ion battery. ► The highest temperature of branched oligomer-containing battery was only 85 °C in the nail penetration test. - Abstract: In this study, we have investigated N-phenylmaleimide/bismaleimide-containing branched oligomer (BO1) as additive in Li-ion batteries to increase the safety performance by reducing the probability of batteries suffering an internal short circuit. In the nail penetration test, a LiCoO 2 /MCMB full battery with N-phenylmaleimide/bismaleimide-containing branched oligomer (BO1) showed a significant improvement in thermal stability and was able to restrain the temperature of the battery at about 85 °C. Furthermore, we found that N-phenylmaleimide/bismaleimide-containing branched oligomer (BO1) contained battery revealed better cycling and electrochemical performance, compared with the battery with bismaleimide-containing branched oligomer (BO3) in the electrolyte. The improvement might result from the favorable ionic conductivity, Li ion mobility and lower resistance in the battery. This additive can meet the cycling performance and safety requirements for Li-ion batteries.

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)

Conducting organizational safety reviews - requirements, methods and experience

International Nuclear Information System (INIS)

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

2008-03-01

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

Conducting organizational safety reviews - requirements, methods and experience

Energy Technology Data Exchange (ETDEWEB)

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

2008-03-15

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

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

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

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

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

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.

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

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

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

Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

Directory of Open Access Journals (Sweden)

Zongzhi Li

2017-04-01

Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.

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

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.

Implications of safety requirements for the treatment of THMC processes in geological disposal systems for radioactive waste

Directory of Open Access Journals (Sweden)

Frédéric Bernier

2017-06-01

Full Text Available The mission of nuclear safety authorities in national radioactive waste disposal programmes is to ensure that people and the environment are protected against the hazards of ionising radiations emitted by the waste. It implies the establishment of safety requirements and the oversight of the activities of the waste management organisation in charge of implementing the programme. In Belgium, the safety requirements for geological disposal rest on the following principles: defence-in-depth, demonstrability and the radiation protection principles elaborated by the International Commission on Radiological Protection (ICRP. Applying these principles requires notably an appropriate identification and characterisation of the processes upon which the safety functions fulfilled by the disposal system rely and of the processes that may affect the system performance. Therefore, research and development (R&D on safety-relevant thermo-hydro-mechanical-chemical (THMC issues is important to build confidence in the safety assessment. This paper points out the key THMC processes that might influence radionuclide transport in a disposal system and its surrounding environment, considering the dynamic nature of these processes. Their nature and significance are expected to change according to prevailing internal and external conditions, which evolve from the repository construction phase to the whole heating–cooling cycle of decaying waste after closure. As these processes have a potential impact on safety, it is essential to identify and to understand them properly when developing a disposal concept to ensure compliance with relevant safety requirements. In particular, the investigation of THMC processes is needed to manage uncertainties. This includes the identification and characterisation of uncertainties as well as for the understanding of their safety-relevance. R&D may also be necessary to reduce uncertainties of which the magnitude does not allow

Barrier and system performances within a safety case: their functioning and evolution with time

International Nuclear Information System (INIS)

Hedin, A.; Voinis, S.; Fillion, E.; Keller, S.; Lalieux, Ph.; Nachmilner, L.; Nys, V.; Rodriguez, J.; Sevougian, D.; Wollrath, J.

2002-01-01

The following six questions were used as the basis for the discussions in a Working Group: - What is the role of each barrier as a function of time or in the different time frames? What is its contribution to the overall system performance or safety as a function of time? - Which are the main uncertainties on the performance of barriers in the timescales? To what extent should we enhance the robustness of barriers because of the uncertainties of some component behaviour with time? - What is the requested or required performance versus the expected realistic or conservative behaviour with time? How are these safety margins used as arguments in a safety case? - What is the issue associated with the geosphere stability for different geological systems? - How are barriers and system performances, as a function of time, evaluated (presented and communicated) in a safety case? - What kind of measures are used for siting, designing and optimising robust barriers corresponding to situations that can vary with time? Are human actions considered to be relevant? (authors)

Occupational safety of different industrial sectors in Khartoum State, Sudan. Part 1: Safety performance evaluation.

Science.gov (United States)

Zaki, Gehan R; El-Marakby, Fadia A; H Deign El-Nor, Yasser; Nofal, Faten H; Zakaria, Adel M

2012-12-01

Safety performance evaluation enables decision makers improve safety acts. In Sudan, accident records, statistics, and safety performance were not evaluated before maintenance of accident records became mandatory in 2005. This study aimed at evaluating and comparing safety performance by accident records among different cities and industrial sectors in Khartoum state, Sudan, during the period from 2005 to 2007. This was a retrospective study, the sample in which represented all industrial enterprises in Khartoum state employing 50 workers or more. All industrial accident records of the Ministry of Manpower and Health and those of different enterprises during the period from 2005 to 2007 were reviewed. The safety performance indicators used within this study were the frequency-severity index (FSI) and fatal and disabling accident frequency rates (DAFR). In Khartoum city, the FSI [0.10 (0.17)] was lower than that in Bahari [0.11 (0.21)] and Omdurman [0.84 (0.34)]. It was the maximum in the chemical sector [0.33 (0.64)] and minimum in the metallurgic sector [0.09 (0.19)]. The highest DAFR was observed in Omdurman [5.6 (3.5)] and in the chemical sector [2.5 (4.0)]. The fatal accident frequency rate in the mechanical and electrical engineering industry was the highest [0.0 (0.69)]. Male workers who were older, divorced, and had lower levels of education had the lowest safety performance indicators. The safety performance of the industrial enterprises in Khartoum city was the best. The safety performance in the chemical sector was the worst with regard to FSI and DAFR. The age, sex, and educational level of injured workers greatly affect safety performance.

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

The performance regulatory approach in quality assurance: Its application to safety in nuclear power plants

International Nuclear Information System (INIS)

Sajaroff, Pedro M.

2000-01-01

In early 1991, the IAEA assembled an Advisory Group on the Comprehensive Revision of the Code and the Safety on Quality Assurance of the NUSS Programme. The Group was made up by specialists from a number of countries and from ISO, FORATOM, the EC and the IAEA itself, and its objective was completed in June 1995. This paper is aimed at describing the conceptual contents of the final draft of the revision 2 of the 50-C-QA Code 'Quality Assurance for Safety in Nuclear Power Plants and other Nuclear Facilities' (hereinafter, the Code) which is essentially based on performance. Although the performance regulatory approach is not new in Argentina and in other countries, what is indeed novel is applying performance based QA. In such a way the Code will contribute to preventing both QA misinterpretations (i.e., a formalistic regulatory requirement) and the execution of non-effective work without attaining the needed quality level (what may be seen as a pathological deviation of QA). The Code contains ten basic requirements to be adopted when QA programmes are established and implemented in nuclear power plants. The goal is improving safety through an improvement in the methods applied for attaining quality. In line with the current developments in quality management techniques, priority is given to effectiveness of the QA programme. All the involved individuals (that is those in the managerial level, those performing the work and those assessing the work performed) must contribute to quality in a co-ordinated manner. The revised Safety Guides are being introduced, standing out those non existing before. Interrelation between quality assurance, safety culture and quality culture is to be noted. Besides QA for safety-related software mentioned as an issue to be considered by the IAEA. (author)

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.

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

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

Tools for the performance assessment and improvement of food safety management systems ; review

NARCIS (Netherlands)

Jacxsens, L.; Luning, P.A.; Marcelis, W.J.; Boekel, van M.A.J.S.; Rovira, J.; Oses Gomez, S.; Kousta, M.; Drosinos, E.H.; Jasson, V.; Uyttendaele, M.

2011-01-01

Food business operators are challenged to combine requirements from different stakeholders (e.g. government, retailers) into a company specific Food Safety Management System (FSMS). Tools to diagnose the performance of an implemented FSMS (diagnostic tools), tools to help a selection process

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.)

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

International Nuclear Information System (INIS)

Koh, Kwang Yong; Seong, Poong Hyun

2005-01-01

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

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

A study of software safety analysis system for safety-critical software

International Nuclear Information System (INIS)

Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

2004-01-01

The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases

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

International Nuclear Information System (INIS)

2004-01-01

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

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...

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)

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.

Safety analysis of MOX fuels by fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

Plant safety and performance indicators for regulatory use

International Nuclear Information System (INIS)

Ferjancic, M.; Nemec, T.; Cimesa, S.

2004-01-01

Slovenian Nuclear Safety Administration (SNSA) supervises nuclear and radiological safety of Krsko NPP. This SNSA supervision is performed through inspections, safety evaluations of plant modifications and event analyses as well as with the safety and performance indicators (SPI) which are a valuable data source for plant safety monitoring. In the past SNSA relied on the SPI provided by Krsko NPP and did not have a set of SPI which would be more appropriate for regulatory use. In 2003 SNSA started with preparation of a new set of SPI which would be more suitable for performing the regulatory oversight of the plant. New internal SNSA procedure which is under preparation will define use and evaluation of SPI and will include definitions for the proposed set of SPI. According to the evaluation of SPI values in comparison with the limiting values and/or trending, the procedure will define SNSA response and actions. (author)

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.

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.

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

Science.gov (United States)

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

2012-03-01

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

Operational safety performance of Slovak NPPs in 2005

International Nuclear Information System (INIS)

Tomek, J.

2006-01-01

In this presentation author presents operational safety performance of Slovak NPPs in 2005. Operation of Slovak NPPs in 2005 was safe and reliable, with: - high level of performance low risk; - minimal impact on the personnel, environment and public; - positive attitude to safety.

An Empirical Analysis of Human Performance and Nuclear Safety Culture

International Nuclear Information System (INIS)

Jeffrey Joe; Larry G. Blackwood

2006-01-01

The purpose of this analysis, which was conducted for the US Nuclear Regulatory Commission (NRC), was to test whether an empirical connection exists between human performance and nuclear power plant safety culture. This was accomplished through analyzing the relationship between a measure of human performance and a plant's Safety Conscious Work Environment (SCWE). SCWE is an important component of safety culture the NRC has developed, but it is not synonymous with it. SCWE is an environment in which employees are encouraged to raise safety concerns both to their own management and to the NRC without fear of harassment, intimidation, retaliation, or discrimination. Because the relationship between human performance and allegations is intuitively reciprocal and both relationship directions need exploration, two series of analyses were performed. First, human performance data could be indicative of safety culture, so regression analyses were performed using human performance data to predict SCWE. It also is likely that safety culture contributes to human performance issues at a plant, so a second set of regressions were performed using allegations to predict HFIS results

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)

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)

Safety performance indicators used by the Russian Safety Regulatory Authority in its practical activities on nuclear power plant safety regulation

International Nuclear Information System (INIS)

Khazanov, A.L.

2005-01-01

The Sixth Department of the Nuclear, Industrial and Environmental Regulatory Authority of Russia, Scientific and Engineering Centre for Nuclear and Radiation Safety process, analyse and use the information on nuclear power plants (NPPs) operational experience or NPPs safety improvement. Safety performance indicators (SPIs), derived from processing of information on operational violations and analysis of annual NPP Safety Reports, are used as tools to determination of trends towards changing of characteristics of operational safety, to assess the effectiveness of corrective measures, to monitor and evaluate the current operational safety level of NPPs, to regulate NPP safety. This report includes a list of the basic SPIs, those used by the Russian safety regulatory authority in regulatory activity. Some of them are absent in list of IAEA-TECDOC-1141 ('Operational safety performance indicators for nuclear power plants'). (author)

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

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

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

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

Benchmarking Global Food Safety Performances: The Era of Risk Intelligence.

Science.gov (United States)

Valleé, Jean-Charles Le; Charlebois, Sylvain

2015-10-01

Food safety data segmentation and limitations hamper the world's ability to select, build up, monitor, and evaluate food safety performance. Currently, there is no metric that captures the entire food safety system, and performance data are not collected strategically on a global scale. Therefore, food safety benchmarking is essential not only to help monitor ongoing performance but also to inform continued food safety system design, adoption, and implementation toward more efficient and effective food safety preparedness, responsiveness, and accountability. This comparative study identifies and evaluates common elements among global food safety systems. It provides an overall world ranking of food safety performance for 17 Organisation for Economic Co-Operation and Development (OECD) countries, illustrated by 10 indicators organized across three food safety risk governance domains: risk assessment (chemical risks, microbial risks, and national reporting on food consumption), risk management (national food safety capacities, food recalls, food traceability, and radionuclides standards), and risk communication (allergenic risks, labeling, and public trust). Results show all countries have very high food safety standards, but Canada and Ireland, followed by France, earned excellent grades relative to their peers. However, any subsequent global ranking study should consider the development of survey instruments to gather adequate and comparable national evidence on food safety.

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

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

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.

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

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

Science.gov (United States)

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

2010-01-01

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

Multilevel Safety Climate and Safety Performance in the Construction Industry: Development and Validation of a Top-Down Mechanism

Directory of Open Access Journals (Sweden)

Ran Gao

2016-11-01

Full Text Available The character of construction projects exposes front-line workers to dangers and accidents. Safety climate has been confirmed to be a predictor of safety performance in the construction industry. This study aims to explore the underlying mechanisms of the relationship between multilevel safety climate and safety performance. An integrated model was developed to study how particular safety climate factors of one level affect those of other levels, and then affect safety performance from the top down. A questionnaire survey was administered on six construction sites in Vietnam. A total of 1030 valid questionnaires were collected from this survey. Approximately half of the data were used to conduct exploratory factor analysis (EFA and the remaining data were submitted to structural equation modeling (SEM. Top management commitment (TMC and supervisors’ expectation (SE were identified as factors to represent organizational safety climate (OSC and supervisor safety climate (SSC, respectively, and coworkers’ caring and communication (CCC and coworkers’ role models (CRM were identified as factors to denote coworker safety climate (CSC. SEM results show that OSC factor is positively related to SSC factor and CSC factors significantly. SSC factor could partially mediate the relationship between OSC factor and CSC factors, as well as the relationship between OSC factor and safety performance. CSC factors partially mediate the relationship between OSC factor and safety performance, and the relationship between SSC factor and safety performance. The findings imply that a positive safety culture should be established both at the organizational level and the group level. Efforts from all top management, supervisors, and coworkers should be provided to improve safety performance in the construction industry.

Multilevel Safety Climate and Safety Performance in the Construction Industry: Development and Validation of a Top-Down Mechanism.

Science.gov (United States)

Gao, Ran; Chan, Albert P C; Utama, Wahyudi P; Zahoor, Hafiz

2016-11-08

The character of construction projects exposes front-line workers to dangers and accidents. Safety climate has been confirmed to be a predictor of safety performance in the construction industry. This study aims to explore the underlying mechanisms of the relationship between multilevel safety climate and safety performance. An integrated model was developed to study how particular safety climate factors of one level affect those of other levels, and then affect safety performance from the top down. A questionnaire survey was administered on six construction sites in Vietnam. A total of 1030 valid questionnaires were collected from this survey. Approximately half of the data were used to conduct exploratory factor analysis (EFA) and the remaining data were submitted to structural equation modeling (SEM). Top management commitment (TMC) and supervisors' expectation (SE) were identified as factors to represent organizational safety climate (OSC) and supervisor safety climate (SSC), respectively, and coworkers' caring and communication (CCC) and coworkers' role models (CRM) were identified as factors to denote coworker safety climate (CSC). SEM results show that OSC factor is positively related to SSC factor and CSC factors significantly. SSC factor could partially mediate the relationship between OSC factor and CSC factors, as well as the relationship between OSC factor and safety performance. CSC factors partially mediate the relationship between OSC factor and safety performance, and the relationship between SSC factor and safety performance. The findings imply that a positive safety culture should be established both at the organizational level and the group level. Efforts from all top management, supervisors, and coworkers should be provided to improve safety performance in the construction industry.

Performance scorecard for occupational safety and health management systems

Directory of Open Access Journals (Sweden)

Hernâni Veloso Neto

2012-06-01

Full Text Available The pro-active and systematic search for best performances should be the two assumptions of any management system, so safety and health management in organizations must also be guided by these same precepts. However, the scientific production evidences that the performance evaluation processes in safety and health continue to be guided, in their essence, by intermittency, reactivity and negativity, which are not consistent with the assumptions referenced above. Therefore, it is essential that health and safety at work management systems (HSW MS are structured from an active and positive viewpoint, focusing on continuous improvement. This implies considering performance evaluation processes that incorporate, on the one hand, monitoring, measuring and verification procedures, and on the other hand, structured matrixes of results that capture the key factors of success, by mobilizing both reactive and proactive indicators. One of the instruments that can fulfill these precepts of health and safety performance evaluation is the SafetyCard, a performance scorecard for HSW MS that we developed and will seek to outline and demonstrate over this paper.

Improvement programme of safety performance indicators (SPIs) in Korea

International Nuclear Information System (INIS)

Lee, S.Y.

2001-01-01

KINS has developed and used Safety Performance Indicators (SPIs), which are count based and composed of 10 indicators in 8 areas, to monitor the trend of performance of NPPs in Korea since 1997. However, the limited usage of SPIs and the increasing worldwide interest on SPIs became the motivation of the SPI improvement programme in Korea. Korea is planning to establish plant performance evaluation programme through analysis of SPI and result of inspection. The SPI improvement programme is a part of the plant performance evaluation programme and includes study on performance evaluation areas, indicator categories, selection and development of indicators, redefinition of indicators and introduction of graphical display system. The selected performance evaluation areas are general performance, reactor safety and radiation safety. Each area will have categories as sub-areas and a total of six categories are selected. One or two indicators for each category are determined or will be developed to make a set of Safety Performance Indicators. Also, a graphic display system will be introduced to extend the usage of SPIs. (author)

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)

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

International Nuclear Information System (INIS)

Anon.

1992-01-01

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

Fuel supply shutdown facility interim operational safety requirements

International Nuclear Information System (INIS)

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

1995-01-01

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

Lean Six-Sigma in Aviation Safety: An implementation guide for measuring aviation system’s safety performance

OpenAIRE

Panagopoulos, I.; Atkin, C.J.; Sikora, I.

2016-01-01

The paper introduces a conceptual framework that could improve the safety performance measurement process and ultimately the aviation system safety performance. The framework provides an implementation guide on how organisations could design and develop a proactive, measurement tool for assessing and measuring the Acceptable Level of Safety Performance (ALoSP) at sigma (σ) level, a statistical measurement unit. In fact, the methodology adapts and combines quality management tools, a leading i...

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)

Human performance analysis in the frame of probabilistic safety assessment of research reactors

International Nuclear Information System (INIS)

Farcasiu, Mita; Nitoi, Mirela; Apostol, Minodora; Turcu, I.; Florescu, Gh.

2005-01-01

Full text: The analysis of operating experience has identified the importance of human performance in reliability and safety of research reactors. In Probabilistic Safety Assessment (PSA) of nuclear facilities, human performance analysis (HPA) is used in order to estimate human error contribution to the failure of system components or functions. HPA is a qualitative and quantitative analysis of human actions identified for error-likely situations or accident-prone situations. Qualitative analysis is used to identify all man-machine interfaces that can lead to an accident, types of human interactions which may mitigate or exacerbate the accident, types of human errors and performance shaping factors. Quantitative analysis is used to develop estimates of human error probability as effects of human performance in reliability and safety. The goal of this paper is to accomplish a HPA in the PSA frame for research reactors. Human error probabilities estimated as results of human actions analysis could be included in system event tree and/or system fault tree. The achieved sensitivity analyses determine human performance sensibility at systematically variations both for dependencies level between human actions and for operator stress level. The necessary information was obtained from operating experience of research reactor TRIGA from INR Pitesti. The required data were obtained from generic data bases. (authors)

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)

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.

Health, safety and environmental unit performance assessment model under uncertainty (case study: steel industry).

Science.gov (United States)

Shamaii, Azin; Omidvari, Manouchehr; Lotfi, Farhad Hosseinzadeh

2017-01-01

Performance assessment is a critical objective of management systems. As a result of the non-deterministic and qualitative nature of performance indicators, assessments are likely to be influenced by evaluators' personal judgments. Furthermore, in developing countries, performance assessments by the Health, Safety and Environment (HSE) department are based solely on the number of accidents. A questionnaire is used to conduct the study in one of the largest steel production companies in Iran. With respect to health, safety, and environment, the results revealed that control of disease, fire hazards, and air pollution are of paramount importance, with coefficients of 0.057, 0.062, and 0.054, respectively. Furthermore, health and environment indicators were found to be the most common causes of poor performance. Finally, it was shown that HSE management systems can affect the majority of performance safety indicators in the short run, whereas health and environment indicators require longer periods of time. The objective of this study is to present an HSE-MS unit performance assessment model in steel industries. Moreover, we seek to answer the following question: what are the factors that affect HSE unit system in the steel industry? Also, for each factor, the extent of impact on the performance of the HSE management system in the organization is determined.

ON THE IMPACT OF FLIGHT SAFETY CERTIFICATION REQUIREMENTS ON THE AERODYNAMIC EFFICIENCY OF COMMERCIAL AIRPLANES

Directory of Open Access Journals (Sweden)

Vladimir I. Shevyakov

2018-01-01

Full Text Available The article considers the issue of aerodynamics efficiency implementation taking into account certification requirements for flight safety. Aerodynamics efficiency means high aerodynamic performance (depending on the airplane size, aerodynamic performance in cruise flight, high aerodynamic performance at takeoff, as well as lift performance at landing.The author estimated the impact on aerodynamics efficiency of both the requirements for aerodynamics performance and requirements for aircraft systems, noncompliance with which may result in significant change of expected operating conditions. It was shown that the use of supercritical wing profiles may result in flight mode limitations due to failure of the required buffeting capacities. It does not allow engaging all the advantages of aerodynamics layout and requires special design solutions to prevent such cases.There were reviewed certification requirements for flight level pressure altitude accuracy and icing conditions warning sysytem. The research presented the methods of aerodynamic efficiency increase by meeting the requirements for reduced vertical separation minima flights and in icing conditions, including requirements for air data probes. Reduced vertical separation minima flight requirements are met by means of efficient air data probes location. Theoretical methods of flow calculation determine areas on the airplane skin surface where static probes minimize errors depending on angle-of-attack and sideslip. It was shown that if certification requirements are not met and in case of flight out of reduced vertical separation minima area, aerodynamics efficiency is significantly reduced and fuel consumption can be increased by 10% and higher. Suggested approaches implementation allows increasing commercial airplanes competitiveness.

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

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

International Nuclear Information System (INIS)

2000-01-01

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

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

International Nuclear Information System (INIS)

2004-01-01

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

Focus on safety : a comparative analysis of pipeline safety performance 2000-2002

International Nuclear Information System (INIS)

2004-01-01

Canada's National Energy Board (NEB) is responsible for the promotion of safety, environmental protection and economic efficiency in the Canadian public interest in regulating the design, construction, operation and abandonment of interprovincial and international pipelines within Canada. This second annual report provides a review of the safety performance of oil and gas pipeline companies that are regulated by the NEB. The data used to prepare this report originates from two sources: incident reports submitted under the Onshore Pipeline Regulations, 1999, and from information voluntarily provided by pipeline companies under the Safety Performance Indicators (SPI) initiative. Data comparisons with external reference organizations were included. Six key indicators have been identified to provide comprehensive measures of safety performance for pipeline companies: fatalities, ruptures, injury frequencies, liquid releases, gas releases, and unauthorized activities on the right of way. The safety performance of the federally regulated pipeline industry within Canada was satisfactory during this reporting period (2000-2002). The contractor injury frequency rates reported in 2002 were lower than those reported in 2001, and exhibited more consistency with the levels reported in 2000. The NEB is of the opinion that the elevated number of liquid hydrocarbon spills reported in 2000 were a result of elevated construction levels. No fatalities were reported. There was an increase to three from two in the number of ruptures, due in large part to metal loss (corrosion) and cracking, and external interference (third party damage). The number of spills increased to 76 in 2002 from 55 in 2001, which appears to be more in line with industry averages. The volume of hydrocarbon liquid released in 2002 represented one third the volume released in 2001. refs., 5 tabs., 14 figs

Drug Safety Monitoring in Children: Performance of Signal Detection Algorithms and Impact of Age Stratification

NARCIS (Netherlands)

O.U. Osokogu (Osemeke); C. Dodd (Caitlin); A.C. Pacurariu (Alexandra C.); F. Kaguelidou (Florentia); D.M. Weibel (Daniel); M.C.J.M. Sturkenboom (Miriam)

2016-01-01

textabstractIntroduction: Spontaneous reports of suspected adverse drug reactions (ADRs) can be analyzed to yield additional drug safety evidence for the pediatric population. Signal detection algorithms (SDAs) are required for these analyses; however, the performance of SDAs in the pediatric

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

International Nuclear Information System (INIS)

1994-10-01

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

A review on the benchmarking concept in Malaysian construction safety performance

Science.gov (United States)

Ishak, Nurfadzillah; Azizan, Muhammad Azizi

2018-02-01

Construction industry is one of the major industries that propels Malaysia's economy in highly contributes to our nation's GDP growth, yet the high fatality rates on construction sites have caused concern among safety practitioners and the stakeholders. Hence, there is a need of benchmarking in performance of Malaysia's construction industry especially in terms of safety. This concept can create a fertile ground for ideas, but only in a receptive environment, organization that share good practices and compare their safety performance against other benefit most to establish improvement in safety culture. This research was conducted to study the awareness important, evaluate current practice and improvement, and also identify the constraint in implement of benchmarking on safety performance in our industry. Additionally, interviews with construction professionals were come out with different views on this concept. Comparison has been done to show the different understanding of benchmarking approach and how safety performance can be benchmarked. But, it's viewed as one mission, which to evaluate objectives identified through benchmarking that will improve the organization's safety performance. Finally, the expected result from this research is to help Malaysia's construction industry implement best practice in safety performance management through the concept of benchmarking.

Use of safety management practices for improving project performance.

Science.gov (United States)

Cheng, Eddie W L; Kelly, Stephen; Ryan, Neal

2015-01-01

Although site safety has long been a key research topic in the construction field, there is a lack of literature studying safety management practices (SMPs). The current research, therefore, aims to test the effect of SMPs on project performance. An empirical study was conducted in Hong Kong and the data collected were analysed with multiple regression analysis. Results suggest that 3 of the 15 SMPs, which were 'safety committee at project/site level', 'written safety policy', and 'safety training scheme' explained the variance in project performance significantly. Discussion about the impact of these three SMPs on construction was provided. Assuring safe construction should be an integral part of a construction project plan.

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

International Nuclear Information System (INIS)

2014-01-01

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

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

International Nuclear Information System (INIS)

2016-01-01

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

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

International Nuclear Information System (INIS)

2015-01-01

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

Safety performance monitoring of autonomous marine systems

International Nuclear Information System (INIS)

Thieme, Christoph A.; Utne, Ingrid B.

2017-01-01

The marine environment is vast, harsh, and challenging. Unanticipated faults and events might lead to loss of vessels, transported goods, collected scientific data, and business reputation. Hence, systems have to be in place that monitor the safety performance of operation and indicate if it drifts into an intolerable safety level. This article proposes a process for developing safety indicators for the operation of autonomous marine systems (AMS). The condition of safety barriers and resilience engineering form the basis for the development of safety indicators, synthesizing and further adjusting the dual assurance and the resilience based early warning indicator (REWI) approaches. The article locates the process for developing safety indicators in the system life cycle emphasizing a timely implementation of the safety indicators. The resulting safety indicators reflect safety in AMS operation and can assist in planning of operations, in daily operational decision-making, and identification of improvements. Operation of an autonomous underwater vehicle (AUV) exemplifies the process for developing safety indicators and their implementation. The case study shows that the proposed process leads to a comprehensive set of safety indicators. It is expected that application of the resulting safety indicators consequently will contribute to safer operation of current and future AMS. - Highlights: • Process for developing safety indicators for autonomous marine systems. • Safety indicators based on safety barriers and resilience thinking. • Location of the development process in the system lifecycle. • Case study on AUV demonstrating applicability of the process.

Mining Behavior Based Safety Data to Predict Safety Performance

Energy Technology Data Exchange (ETDEWEB)

Jeffrey C. Joe

2010-06-01

The Idaho National Laboratory (INL) operates a behavior based safety program called Safety Observations Achieve Results (SOAR). This peer-to-peer observation program encourages employees to perform in-field observations of each other's work practices and habits (i.e., behaviors). The underlying premise of conducting these observations is that more serious accidents are prevented from occurring because lower level “at risk” behaviors are identified and corrected before they can propagate into culturally accepted “unsafe” behaviors that result in injuries or fatalities. Although the approach increases employee involvement in safety, the premise of the program has not been subject to sufficient empirical evaluation. The INL now has a significant amount of SOAR data on these lower level “at risk” behaviors. This paper describes the use of data mining techniques to analyze these data to determine whether they can predict if and when a more serious accident will occur.

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

African Journals Online (AJOL)

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

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

Directory of Open Access Journals (Sweden)

Marek L. Kowalski

2016-10-01

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

Nuclear Safety Regulations

International Nuclear Information System (INIS)

Novosel, N.; Prah, M.

2008-01-01

Beside new Ordinance on the control of nuclear material and special equipment ('Official Gazette' No. 15/08), from 2006 State Office for Nuclear Safety (SONS) adopted Ordinance on performing nuclear activities ('Official Gazette' No. 74/06) and Ordinance on special requirements which expert organizations must fulfil in order to perform certain activities in the field of nuclear safety ('Official Gazette' No. 74/06), based on Nuclear Safety Act ('Official Gazette' No. 173/03). The Ordinance on performing nuclear activities regulates the procedure of notification of the intent to perform nuclear activities, submitting the application for the issue of a licence to perform nuclear activities, and the procedure for issuing decisions on granting a licence to perform a nuclear activity. The Ordinance also regulates the content of the forms for notification of the intent to perform nuclear activities, as well as of the application for the issue of a licence to perform the nuclear activity and the method of keeping the register of nuclear activities. According to the Nuclear Safety Act, nuclear activities are the production, processing, use, storage, disposal, transport, import, export, possession or other handling of nuclear material or specified equipment. The Ordinance on special requirements which expert organizations must fulfil in order to perform certain activities in the field of nuclear safety regulates these mentioned conditions, whereas compliance is established by a decision passed by the SONS. Special requirements which expert organizations must fulfil in order to perform certain activities in the field of nuclear safety are organizational, technical, technological conditions and established system of quality assurance. In 2007, State Office for Nuclear Safety finalized the text of new Ordinance on conditions for nuclear safety and protection with regard to the siting, design, construction, use and decommissioning of a facility in which a nuclear activity is

Technical Safety Requirements for the Gamma Irradiation Facility (GIF)

CERN Document Server

Mahn, J A E M J G

2003-01-01

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

Evaluating Performance of Safety Management and Occupational Health Using Total Quality Safety Management Model (TQSM

Directory of Open Access Journals (Sweden)

E Mohammadfam

2015-11-01

Full Text Available Introduction: All organizations, whether public or private, necessitate performance evaluation systems in regard with growth, stability, and development in the competitive fields. One of the existing models for performance evaluation of occupational health and safety management is Total Quality Safety Management model (TQSM. Therefore, the present study aimed to evaluate performance of safety management and occupational health utilizing TQSM model. Methods: In this descriptive-analytic study, the population consisted of 16 individuals, including managers, supervisors, and members of technical protection and work health committee. Then the participants were asked to respond to TQSM questionnaire before and after the implementation of Occupational Health & Safety Advisory Services 18001 (OHSAS18001. Ultimately, the level of each program as well as the TQSM status were determined before and after the implementation of OHSAS18001. Results: The study results showed that the scores obtained by the company before OHSAS 18001’s implementation, was 43.7 out of 312. After implementing OHSAS 18001 in the company and receiving the related certificate, the total score of safety program that company could obtain was 127.12 out of 312 demonstrating a rise of 83.42 scores (26.8%. The paired t-test revealed that mean difference of TQSM scores before and after OHSAS 18001 implementation was proved to be significant (p> 0.05. Conclusion: The study findings demonstrated that TQSM can be regarded as an appropriate model in order to monitor the performance of safety management system and occupational health, since it possesses the ability to quantitatively evaluate the system performance.

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

Science.gov (United States)

2012-12-20

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

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

Safety evaluation report on Tennessee Valley Authority: Browns Ferry nuclear performance plan

International Nuclear Information System (INIS)

1989-10-01

This safety evaluation report (SER) on the information submitted by the Tennessee Valley Authority (TVA) in its Nuclear Performance Plan, through Revision 2, for the Browns Ferry Nuclear Plant and in supporting documents has been prepared by the US Nuclear Regulatory commission staff. The Browns Ferry Nuclear Plant consists of three boiling-water reactors at a site in Limestone County, Alabama. The plan addresses the plant-specific concerns requiring resolution before the startup of Unit 2. The staff will inspect implementation of those TVA programs that address these concerns. Where systems are common to Units 1 and 2 or to Units 2 and 3, the staff safety evaluations of those systems are included herein. 85 refs

Safety evaluation report on Tennessee Valley Authority: Watts Bar Nuclear Performance Plan

International Nuclear Information System (INIS)

1990-01-01

This safety evaluation report on the information submitted by the Tennessee Valley Authority in its Nuclear Performance Plan for the Watts Bar Nuclear Plant and in supporting documents has been prepared by the US Nuclear Regulatory Commission staff. The plan addresses the plant-specific corrective actions as part of the recovery program for licensing of Unit 1. The staff will be monitoring and inspecting the implementation of the programs. The plan does not address all licensing matters that will be required for fuel load and operation of Unit 1. Those remaining licensing matters have been addressed in previous safety evaluations or will be addressed in accordance with routing NRC licensing practices. 97 refs

Nuclear safety review requirements for launch approval

International Nuclear Information System (INIS)

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

1992-01-01

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

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

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.

Operational Safety Performance Indicators and Balanced Scorecard in HANARO

International Nuclear Information System (INIS)

Wu, Jong-Sup; Jung, Hoan-Sung; Ahn, Guk-Hoon; Lee, Kye-Hong; Lim, In-Cheol; Kim, Hark-Rho

2007-01-01

Research reactors need an extensive basis for ensuring their safety. The importance of a safety management in nuclear facilities and activities has been emphasized. The safety activities in HANARO have been continuously conducted to enhance its safe operation. Last year, HANARO prepared two indicator sets to measure and assess the safety status of the reactor's operation and utilization. One is Safety Performance Indicators (SPI) and the other is Balanced Scorecard (BSC). Through reviewing these indicators, we can obtain the following information; - Plant safety status - Safety parameter trends - Safety information, for example, reactor operation status and radiation safety HANARO will continuously pursue the trends of SPI and BSC

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)

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

International Nuclear Information System (INIS)

Gruendler, D.; Wurtinger, W.

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1993-12-01

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

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Lee, Jang-Soo; Jee, Eunkyoung

2016-01-01

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee-Choon; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jee, Eunkyoung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents.

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

International Nuclear Information System (INIS)

Saha, D.; Sinha, R.K.

2002-01-01

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

Performance Test Results of Safety I and C Systems of SMART MMIS

International Nuclear Information System (INIS)

Suh, Yong Suk; Keum, Jong Yong; Jeong, Kwang Il; Lee, Joon Ku; Lee, Sang Seok; Kim, Kwan Woong

2011-01-01

KAERI has developed SMART (System-integrated Modular Advanced ReacTor), a 330MWt integral pressurized light water reactor that integrates four reactor coolant pumps, one pressurizer, eight steam generators, and one reactor core into a reactor vessel, since 1997 and submitted a SSAR (Standard design Safety Analysis Report) to Korea institute of nuclear safety (KINS) at the end of 2010 for the purpose of achieving the standard design approval (SDA) by the end of 2011. SMART MMIS has been designed with fully digitalized systems. Non-safety instrumentation and control (I and C) systems are designed based on the commercial distributed control systems. The safety I and C systems are designed using a new platform that was developed and validated by KAERI. Safety I and C systems are modularized using the platform. In the protection systems (PSs), datalinks are used to transmit data in a one-way direction in order to meet the independency requirement. In the engineered safety features-component control system (ESF-CCS), network switch devices (NSDs) are used to connect the group and loop controllers. The NSD was also newly developed and validated by KAERI. After validating the platform and NSD, a test facility was developed using the platform and NSDs to validate the performance of safety I and C systems. This paper presents the development and test results from the test facility

Road safety performance indicators for the interurban road network.

NARCIS (Netherlands)

Yannis, G. Weijermars, W.A.M. Gitelman, V. Vis, M. Chazirisa, A. Papadimitriou, E. & Lima Azevedo, C.

2013-01-01

Various road safety performance indicators (SPIs) have been proposed for different road safety research areas, mainly as regards driver behaviour (e.g. seat belt use, alcohol, drugs, etc.) and vehicles (e.g. passive safety); however, no SPIs for the road network and design have been developed. The

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

Science.gov (United States)

2010-10-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-05-01

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

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

Science.gov (United States)

2010-10-01

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

Managing patient safety through NPSGs and employee performance.

Science.gov (United States)

Adair, Liberty

2010-01-01

Patient safety can only exist in a culture of patient safety, which implies it is a value perceived by all. Culture predicts safety outcomes and leadership predicts the culture. Leaders are obligated to continually mitigate hazard and take action consciously. Healthcare workers should focus on preventing and reporting mistakes with the National Patient Safety Goals (NPSGs) in mind. These include: accuracy of patient identification, effectiveness of communication among caregivers, improving safety of medications, reducing infections, reducing risk of falls, and encouraging patients to be involved in care. Poor performers and reckless behavior need to be mitigated. If employees recognize their roles in the process, feel empowered,and have appropriate tools, resources,and data to implement solutions, errors can be avoided and patient safety becomes paramount.

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

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)

Development of Occupational Safety and Health Requirement Management System (OSHREMS Software Using Adobe Dreamweaver CS5 for Building Construction Project

Directory of Open Access Journals (Sweden)

Abas Nor Haslinda

2017-01-01

Full Text Available The construction industry sector is considered as being risky with frequent and high accident rate. According to Social Security Organization (SOCSO, the construction accidents has arisen from time to time. Construction Industry Development Board (CIDB has developed the Safety and Health Assessment System in Construction (SHASSIC for evaluating the performance of a contractor in construction project by setting out the safety and health management and practices, however the requirement checklist provided is not comprehensive. Therefore, this study aims to develop a software system for facilitating OSH in building construction project, namely OSH requirements management system (OSHREMS, using Adobe Dreamweaver CS5 and Sublime Text as PHP editor. The results from a preliminary study which was conducted through interviews showed that, the respondents were only implementing the basic requirements that comply with legislations, with the absence of appropriate and specific guideline in ensuring occupational safety and health (OSH at the workplace. The tool will be benefits for contractors and other parties to effectively manage the OSH requirements for their projects based on project details.

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

Operational safety performance indicators for nuclear power plants

International Nuclear Information System (INIS)

2000-05-01

Since the late 1980s, the IAEA has been actively sponsoring work in the area of indicators to monitor nuclear power plant (NPP) operational safety performance. The early activities were mainly focused on exchanging ideas and good practices in the development and use of these indicators at nuclear power plants. Since 1995 efforts have been directed towards the elaboration of a framework for the establishment of an operational safety performance indicator programme. The result of this work, compiled in this publication, is intended to assist NPPs in developing and implementing a monitoring programme, without overlooking the critical aspects related to operational safety performance. The framework proposed in this report was presented at two IAEA workshops on operational safety performance indicators held in Ljubljana, Slovenia, in September 1998 and at the Daya Bay NPP, Szenzhen, China, in December 1998. During these two workshops, the participants discussed and brainstormed on the indicator framework presented. These working sessions provided very useful insights and ideas which where used for the enhancement of the framework proposed. The IAEA is acknowledging the support and contribution of all the participants in these two activities. The programme development was enhanced by pilot plant studies. Four plants from different countries with different designs participated in this study with the objective of testing the applicability, usefulness and viability of this approach

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

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.

2015-01-01

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

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.

Performance Measurement and Target-Setting in California's Safety Net Health Systems.

Science.gov (United States)

Hemmat, Shirin; Schillinger, Dean; Lyles, Courtney; Ackerman, Sara; Gourley, Gato; Vittinghoff, Eric; Handley, Margaret; Sarkar, Urmimala

Health policies encourage implementing quality measurement with performance targets. The 2010-2015 California Medicaid waiver mandated quality measurement and reporting. In 2013, California safety net hospitals participating in the waiver set a voluntary performance target (the 90th percentile for Medicare preferred provider organization plans) for mammography screening and cholesterol control in diabetes. They did not reach the target, and the difference-in-differences analysis suggested that there was no difference for mammography ( P = .39) and low-density lipoprotein control ( P = .11) performance compared to measures for which no statewide quality improvement initiative existed. California's Medicaid waiver was associated with improved performance on a number of metrics, but this performance was not attributable to target setting on specific health conditions. Performance may have improved because of secular trends or systems improvements related to waiver funding. Relying on condition-specific targets to measure performance may underestimate improvements and disadvantage certain health systems. Achieving ambitious targets likely requires sustained fiscal, management, and workforce investments.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

SAFETY PERFORMANCE OF SUBCONTRACTORS IN THE PALESTINIAN CONSTRUCTION INDUSTRY

Directory of Open Access Journals (Sweden)

Adnan Enshassi

2008-06-01

Full Text Available Subcontractors perform most of the construction works and their effect on industry are apparent in different activities of construction. Therefore, subcontractors need more attention from government and contractors union. The aim of this paper is to identify, evaluate, and rank factors that influence safety performance of subcontractors in the Gaza Strip (Palestine according to their relative importance. The study concluded that reported accident rates will decrease among subcontractors and their workers if new workers are trained well in the work site and they are informed about dangerous places, and if a workable safety plan is well preplanned. The results also showed that reported accident rates increased among subcontractors when using old, unsafe equipment and due to the complexity or difficulty in the construction sites features. Owners and general contractors need to stipulate strict clauses for safety in the contract for improving safety record of subcontractors. Construction workers must receive proper job related safety and health training with a safety logbook. It is recommended that the subcontractors and workers should attend continuing safety programs on regular basis as part of their perquisite to work in construction sites.

Requirements of radiation protection and safety for nuclear medicine services

International Nuclear Information System (INIS)

1989-01-01

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

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

Science.gov (United States)

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

2008-05-01

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

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

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

International Nuclear Information System (INIS)

2005-01-01

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

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...

National Ignition Facility Project Site Safety Program

International Nuclear Information System (INIS)

Dun, C

2003-01-01

This Safety Program for the National Ignition Facility (NIF) presents safety protocols and requirements that management and workers shall follow to assure a safe and healthful work environment during activities performed on the NIF Project site. The NIF Project Site Safety Program (NPSSP) requires that activities at the NIF Project site be performed in accordance with the ''LLNL ES and H Manual'' and the augmented set of controls and processes described in this NIF Project Site Safety Program. Specifically, this document: (1) Defines the fundamental NIF site safety philosophy. (2) Defines the areas covered by this safety program (see Appendix B). (3) Identifies management roles and responsibilities. (4) Defines core safety management processes. (5) Identifies NIF site-specific safety requirements. This NPSSP sets forth the responsibilities, requirements, rules, policies, and regulations for workers involved in work activities performed on the NIF Project site. Workers are required to implement measures to create a universal awareness that promotes safe practice at the work site and will achieve NIF management objectives in preventing accidents and illnesses. ES and H requirements are consistent with the ''LLNL ES and H Manual''. This NPSSP and implementing procedures (e.g., Management Walkabout, special work procedures, etc.,) are a comprehensive safety program that applies to NIF workers on the NIF Project site. The NIF Project site includes the B581/B681 site and support areas shown in Appendix B

Safety performance evaluation using proactive indicators in a selected industry

Directory of Open Access Journals (Sweden)

Abolfazl Barkhordari

2015-03-01

Full Text Available Background & Objectives: Quality and effectiveness of safety systems are critical factors in achieving their goals. This study was aimed to represent a method for performance evaluation of safety systems by proactive indicators using different updated models in the field of safety which will be tested in a selected industry. Methods: This study is a cross-sectional study. Proactive indicators used in this study were: Unsafe acts rate, Safety Climate, Accident Proneness, and Near-miss incident rate. The number of in 1473 safety climate questionnaires and 543 Accident Proneness questionnaires was completed. Results: The minimum and maximum safety climate score were 56.88 and 58.2, respectively, and the minimum and maximum scores of Accident Proneness were 98.2 and 140.7, respectively. The maximum number of Near-miss incident rate were 408 and the minimum of that was 196. The maximum number of unsafe acts rate was 43.8 percent and the minimum of that was 27.2 percent. In nine dimensions of Safety climate the eighth dimension (personal perception of risk with the score of 4.07 has the lowest score and the fourth (laws and safety regulations dimension with 8.05 has the highest score. According to expert opinions, the most important indicator in the assessment of safety performance was unsafe acts rate, while near-miss incident rate was the least important one. Conclusion: The results of this survey reveal that using proactive (Prospective indicators could be an appropriate method in organizations safety performance evaluation.

Indicators to monitor NPP safety performance. Working material

International Nuclear Information System (INIS)

1997-01-01

Numerical indicators to monitor safety status and overall safety performance of nuclear power plants (NPPs) are used by operators and some regulators worldwide. During the last few years, the IAEA, through Technical Committee Meetings and Consultants' Meetings has worked on this area. This report presents a framework for nuclear power plant safety performance indicators that was developed during two consultant meetings held at the IAEA headquarters in Vienna in December 1995 and November 1996. Annex 2 on risk based indicators was prepared during a consultants' meeting held in Vienna in July 1996. An additional outcome from these activities, was the recommendation that the IAEA conduce pilot exercises at several nuclear power plants that might be interested to participate, in order to test the validity of the concept and its usefulness. 6 figs

Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)

Core size effects on safety performances of LMRs

Energy Technology Data Exchange (ETDEWEB)

Na, Byung Chan; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

An oxide fuel small size core (1200 MWt) was analyzed in comparison with a large size core (3600 MWt) in order to evaluate the size effects on transient safety performances of liquid-metal reactors (LMRs). In the first part of the study, main static safety parameters (i.e., Doppler coefficient, sodium void effect, etc.) of the two cores were characterized, and the second part of the study was focused on the dynamic behavior of the cores in two representative transient events: the unprotected loss-of-flow (ULOF) and the unprotected transient overpower (UTOP). Margins to fuel melting and sodium boiling have been evaluated for these representative transients. Results show that the small core has a generally better or equivalent level of safety performances during these events. 6 refs., 4 figs., 2 tabs. (Author)

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.

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

Science.gov (United States)

2010-09-29

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

Technical Safety Requirements for the Waste Storage Facilities May 2014

Energy Technology Data Exchange (ETDEWEB)

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

2014-04-16

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

Technical Safety Requirements for the Waste Storage Facilities May 2014

International Nuclear Information System (INIS)

Laycak, D. T.

2014-01-01

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

Preparation of safety regulatory requirements for new technology like digital system

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The current regulatory requirements on digital instrumentation and control system have been reviewed by JNES, considering international trend discussed in DICWG of MDEP. MDEP DICWG held in OECD/NEA gives the opportunity to identify the convergence of applicable standards. The working group's activities include: identifying and prioritising the member countries' challenges, practices, and needs regarding standards and regulatory guidance on digital instrumentation and control; identifying areas of importance and needs for convergence of existing standards and guidance or development of new standards; sharing of information; and identifying common positions among the member countries for areas of particular importance and need. The DICWG drafted common positions on specific issues which are based on the existing standards, national regulatory guidance, best practices, and group inputs using an agreed process and framework. The following four general common positions have been discussed in this fiscal year. The Treatment of Common Cause Failure Resulting from Software within Digital Safety Systems, The Treatment of Hardware Description Language(HDL) Programmed Devices for Use in Nuclear Safety System, Factory Acceptance Test and Site Acceptance Test, The Use of Automatic Tests to Perform Surveilance for Digital Systems. (author)

The spin project: safety and performance indicators in different time frames

International Nuclear Information System (INIS)

Storck, R.; Becker, D.A.

2002-01-01

Safety and performance indicators have been under discussion for many years in several countries and international organisations. If those indicators refer to the long term safety of the total disposal system, they are often called safety indicators. If they refer to the performance of subsystems or the total system from a more technical point of view, they are sometimes called performance indicators. The need for indicators other than dose rates derives e.g. from the long time frames involved in safety assessments of waste disposal systems and the increasing uncertainty in dose rate calculations over time due to uncertainty in evolution of the surface environment and of behaviour of man. Before introducing additional indicators into a safety case of a potential repository site, the applicability and usefulness of different indicators have to be investigated and evaluated. The systematic analysis and testing of safety and performance indicators for use in different time horizons after closure of the disposal facility is the task of the SPIN project. This is done by re-calculating four recent studies concerning repository projects in granite formations. (authors)

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

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

International Nuclear Information System (INIS)

2016-01-01

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

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

International Nuclear Information System (INIS)

2016-01-01

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

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

International Nuclear Information System (INIS)

2016-01-01

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

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.)

Safety and performance of a long life lithium-thionyl chloride battery

Science.gov (United States)

Cieslak, W. R.; Street, H. K.

A Li/SOCl2 'D' cell for applications requiring 10 to 15 years life at very low drain rates, typically less than 150 microA, were developed. Maximizing cell safety and reliability, while delivering very good energy density, were the goals. These goals were achieved by designing the cell to be application specific. The low-rate cell was optimized to deliver up to 16 Ah at drain rates of less than 70 mA. By virtue of its low surface area, 145 cm(sup 2), the cell demonstrated excellent safety behavior. Safety testing was performed on individual cells as well as on two-cell and four-cell batteries. Single cells did not vent when short-circuited. Benign venting in a two cell string was produced, but only when the string was partially discharged before shorting. The vent mechanism is a 300 psi rupture pressure burst disc manufactured by BS&B Safety Systems. Benign venting is defined as full opening of the 3/8 in. dia vent hole without deformation of the case. Material is expelled from the cell without flame, and the cell stack remains largely intact. Venting of the Sandia-designed low rate cell was not produced under any other abuse test conditions. The vent functions as an ultimate safety mechanism in the case of severe abuse, but resistance to venting under normal use and mild abuse conditions is key to the achievement of high reliability.

Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Yun Goo; Oh, Eung Se [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2016-05-15

During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

International Nuclear Information System (INIS)

Kim, Yun Goo; Oh, Eung Se

2016-01-01

During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

Multiscale modeling and characterization for performance and safety of lithium-ion batteries

International Nuclear Information System (INIS)

Pannala, S.; Turner, J. A.; Allu, S.; Elwasif, W. R.; Kalnaus, S.; Simunovic, S.; Kumar, A.; Billings, J. J.; Wang, H.; Nanda, J.

2015-01-01

Lithium-ion batteries are highly complex electrochemical systems whose performance and safety are governed by coupled nonlinear electrochemical-electrical-thermal-mechanical processes over a range of spatiotemporal scales. Gaining an understanding of the role of these processes as well as development of predictive capabilities for design of better performing batteries requires synergy between theory, modeling, and simulation, and fundamental experimental work to support the models. This paper presents the overview of the work performed by the authors aligned with both experimental and computational efforts. In this paper, we describe a new, open source computational environment for battery simulations with an initial focus on lithium-ion systems but designed to support a variety of model types and formulations. This system has been used to create a three-dimensional cell and battery pack models that explicitly simulate all the battery components (current collectors, electrodes, and separator). The models are used to predict battery performance under normal operations and to study thermal and mechanical safety aspects under adverse conditions. This paper also provides an overview of the experimental techniques to obtain crucial validation data to benchmark the simulations at various scales for performance as well as abuse. We detail some initial validation using characterization experiments such as infrared and neutron imaging and micro-Raman mapping. In addition, we identify opportunities for future integration of theory, modeling, and experiments

The safety performance management system: A tool for diagnosis, intervention and measurement

International Nuclear Information System (INIS)

Haber, S.B.; Shurberg, D.A.

2002-01-01

Many organizations depend on human performance to avoid incidents involving significant adverse consequences. Such organizations are typically termed high reliability organizations (HROs). While heavy emphasis has been placed on designing system hardware and software to intercept and mitigate events that could cause adverse consequences, dealing with the design of the human component has proven to be more complicated. Examination of various safety-related incidents makes it clear that human performance, and in particular organizational processes, plays a dominant role. The human errors are of various origins and are typically part of larger organizational processes that encourage unsafe acts that ultimately produce system failures. It is generally postulated that without an effective organizational safety culture, a safe working environment is impossible. While many different perspectives exist from which safety issues might be addressed, a method that allows the quantitative measurement of organizational processes deemed to impact overall safety performance is considered useful to understand the potential for future inadequate safety performance. This paper describes the Safety Performance Management System, a method useful for diagnosis, subsequent intervention and follow-on measurement. Implications for use of this method are presented and the concluding discussion includes insights regarding the general application of the method to improved facility safety performance. (author)

Ferrocyanide Safety Program: Safety criteria for ferrocyanide watch list tanks

International Nuclear Information System (INIS)

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

1994-01-01

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

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

NARCIS (Netherlands)

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

2018-01-01

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

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

Science.gov (United States)

2013-10-31

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

Can Leader–Member Exchange Contribute to Safety Performance in An Italian Warehouse?

Directory of Open Access Journals (Sweden)

Marco G. Mariani

2017-05-01

Full Text Available Introduction: The research considers safety climate in a warehouse and wants to analyze the Leader–Member Exchange (LMX role in respect to safety performance. Griffin and Neal’s safety model was adopted and Leader-Member Exchange was inserted as moderator in the relationships between safety climate and proximal antecedents (motivation and knowledge of safety performance constructs (compliance and participation.Materials and Methods: Survey data were collected from a sample of 133 full-time employees in an Italian warehouse. The statistical framework of Hayes (2013 was adopted for moderated mediation analysis.Results: Proximal antecedents partially mediated the relationship between Safety climate and safety participation, but not safety compliance. Moreover, the results from the moderation analysis showed that the Leader–Member Exchange moderated the influence of safety climate on proximal antecedents and the mediation exist only at the higher level of LMX.Conclusion: The study shows that the different aspects of leadership processes interact in explaining individual proficiency in safety practices.Practical Implications: Organizations as warehouses should improve the quality of the relationship between a leader and a subordinate based upon the dimensions of respect, trust, and obligation for high level of safety performance.

Safety management systems and their role in achieving high standards of operational safety

International Nuclear Information System (INIS)

Coulston, D.J.; Baylis, C.C.

2000-01-01

Achieving high standards of operational safety requires a robust management framework that is visible to all personnel with responsibility for its implementation. The structure of the management framework must ensure that all processes used to manage safety interlink in a logical and coherent manner, that is, they form a management system that leads to continuous improvement in safety performance. This Paper describes BNFL's safety management system (SMS). The SMS has management processes grouped within 5 main elements: 1. Policy, 2. Organisation, 3. Planning and Implementation, 4. Measuring and Reviewing Performance, 5. Audit. These elements reflect the overall process of setting safety objective (from Policy), measuring success and reviewing the performance. Effective implementation of the SMS requires senior managers to demonstrate leadership through their commitment and accountability. However, the SMS as a whole reflects that every employee at every level within BNFL is responsible for safety of operations under their control. The SMS therefore promotes a proactive safety culture and safe operations. The system is formally documented in the Company's Environmental, Health and Safety (EHS) Manual. Within in BNFL Group, the Company structures enables the Manual to provide overall SMS guidance and co-ordination to its range of nuclear businesses. Each business develops the SMS to be appropriate at all levels of its organisation, but ensuring that each level is consistent with the higher level. The Paper concludes with a summary of BNFL's safety performance. (author)

Technical Safety Requirements for the Waste Storage Facilities

International Nuclear Information System (INIS)

Larson, H L

2007-01-01

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

Technical Safety Requirements for the Waste Storage Facilities

Energy Technology Data Exchange (ETDEWEB)

Larson, H L

2007-09-07

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

Validation of risk-based performance indicators: Safety system function trends

International Nuclear Information System (INIS)

Boccio, J.L.; Vesely, W.E.; Azarm, M.A.; Carbonaro, J.F.; Usher, J.L.; Oden, N.

1989-10-01

This report describes and applies a process for validating a model for a risk-based performance indicator. The purpose of the risk-based indicator evaluated, Safety System Function Trend (SSFT), is to monitor the unavailability of selected safety systems. Interim validation of this indicator is based on three aspects: a theoretical basis, an empirical basis relying on statistical correlations, and case studies employing 25 plant years of historical data collected from five plants for a number of safety systems. Results using the SSFT model are encouraging. Application of the model through case studies dealing with the performance of important safety systems shows that statistically significant trends in, and levels of, system performance can be discerned which thereby can provide leading indications of degrading and/or improving performances. Methods for developing system performance tolerance bounds are discussed and applied to aid in the interpretation of the trends in this risk-based indicator. Some additional characteristics of the SSFT indicator, learned through the data-collection efforts and subsequent data analyses performed, are also discussed. The usefulness and practicality of other data sources for validation purposes are explored. Further validation of this indicator is noted. Also, additional research is underway in developing a more detailed estimator of system unavailability. 9 refs., 18 figs., 5 tabs

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.

A road safety performance indicator for vehicle fleet compatibility.

NARCIS (Netherlands)

Christoph, M. Vis, M.A. Rackliff, L. & Stipdonk, H.

2013-01-01

This paper discusses the development and the application of a safety performance indicator which measures the intrinsic safety of a country's vehicle fleet related to fleet composition. The indicator takes into account both the ‘relative severity’ of individual collisions between different vehicle

Strengthening the culture of safety and performance in nuclear installations

International Nuclear Information System (INIS)

Briant, V.S.; Germann, R.P.

1997-01-01

In mid-1995, the International Atomic Energy Agency (IAEA) in Vienna brought together a group of safety culture experts from around the world to explore and summarize those practices they viewed as important in establishing sound safety cultures in nuclear installations. This paper will summarize key findings of the Vienna team and also expand those ideas based on related work in which the authors are engaged. The paper includes a definition of safety culture, a description of three stages of safety culture, and five key practices essential to establishing and maintaining a sound safety culture. Additionally, the authors contradicts the conventional view of safety and production as trade-offs, supporting the Vienna team's conclusion that the principles, attitudes, and practices which bring about sustained levels of high performance are the same as those which enhance safety. Based on input from colleagues in several countries, this appears to hold true across geographical and ethnic boundaries. The authors also discuss how this information can be put to practical use to obtain an objective, measurable, and repeated assessment of the current state of the safety culture within a company, plant or work unit. With that information, leaders are then in the position to act on any of the several parameters which affect both safety and performance effectiveness. (author)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Transformational leadership and employee safety performance: a within-person, between-jobs design.

Science.gov (United States)

Inness, Michelle; Turner, Nick; Barling, Julian; Stride, Chris B

2010-07-01

We investigated the extent to which the safety performance (i.e., self-reported safety compliance and safety participation) of employees with 2 jobs was predicted by their respective supervisors' transformational leadership behaviors. We compared 2 within-person models: a context-specific model (i.e., transformational leadership experienced by employees in 1 context related to those same employees' safety performance only in that context) and a context-spillover model (i.e., transformational leadership experienced by employees in 1 context related to those same employees' safety performance in the same and other contexts). Our sample comprised 159 "moonlighters" (73 men, 86 women): employees who simultaneously hold 2 different jobs, each with a different supervisor, providing within-person data on the influence of different supervisors on employee safety performance across 2 job contexts. Having controlled for individual differences (negative affectivity and conscientiousness) and work characteristics (e.g., hours worked and length of relationship with supervisor), the context-specific model provided the best fit to the data among alternative nested models. Implications for the role of transformational leadership in promoting workplace safety are discussed.

Comparative studies of CERCER and CERMET fuels for EFIT from the viewpoint of core performance and safety

International Nuclear Information System (INIS)

Chen, X.N.; Rineiski, A.; Maschek, W.; Liu, P.; Boccaccini, C.M.; Sobolev, V.; Delage, F.; Rimpault, G.

2011-01-01

The European Facility for Industrial Transmutation (EFIT) has been developed within the 6. EU Framework by the EUROTRANS Program, aiming at a generic conceptual design of an accelerator driven transmuter. This paper deals with assessments of EFIT cores with CERCER and CERMET fuels from the viewpoint of core performance and safety. The conclusive remarks can be drawn as follows. Because of its much better thermal conductivity, the CERMET core can be designed by using thicker pins, so that it has the same or even better transmutation performance compared to the CERCER core. Both CERCER and CERMET fuels fulfill safety requirements. Moreover the CERMET fuel has higher fuel safety margins than the CERCER one. Preliminary analyses show that the CERMET total core power can be further increased by 50% at least without exceeding fuel and clad temperature limits. (authors)

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

International Nuclear Information System (INIS)

Mundigl, S.

2009-01-01

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

Westinghouse Hanford Company health and safety performance report

International Nuclear Information System (INIS)

Rogers, L.

1996-01-01

Topping the list of WHC Safety recognition during this reporting period is a commendation received from the National Safety Council (NSC). The NSC bestowed their Award of Honor upon WHC for significant reduction of incidence rates during CY 1995. The award is based upon a reduction of 48 % or greater in cases involving days away from work, a 30 % or greater reduction in the number of days away, and a 15% or greater reduction in the total number of occupational injuries and illnesses. (page 2-1). A DOE-HQ review team representing the Office of Envirorunent, Safety and Health (EH), visited the Hanford Site during several weeks of the quarter. Ile 40-member Safety Management Evaluation Team (SMET) assessed WHC in the areas of management responsibility, comprehensive requirements, and competence commensurate with responsibility. As part of their new approach to oversight, they focused on the existence of management systems and programs (comparable approach to VPP). Plant/project areas selected for review within WHC were PFP, B Plant/WESF, Tank Farms, and K-Basins (page 2-2). Effective safety meetings, prejob safety meetings, etc., are a cornerstone of any successful safety program. In an effort to improve the reporting of safety meetings, the Safety/Security Meeting Report form was revised. It now provides a mechanism for recording and tracking safety issues (page 2-4). WHC has experienced an increase in the occupational injury and illness incidence rates during the first quarter of CY 1996. Trends show this increase can be partially attributed to inattention to workplace activities due 0999to the uncertainty Hanford employees currently face with recent reduction of force, reorganization, and reengineering efforts (page 2-7). The cumulative CY 1995 lost/restricted workday case incidence rate for the first quarter of CY 1996 (1.28) is 25% below the DOE CY 1991-93 average (1.70). However, the incidence rate increased 24% from the CY 1995 rate of 1.03 (page 2-8). The

Westinghouse Hanford Company health and safety performance report

Energy Technology Data Exchange (ETDEWEB)

Rogers, L.

1996-05-15

Topping the list of WHC Safety recognition during this reporting period is a commendation received from the National Safety Council (NSC). The NSC bestowed their Award of Honor upon WHC for significant reduction of incidence rates during CY 1995. The award is based upon a reduction of 48 % or greater in cases involving days away from work, a 30 % or greater reduction in the number of days away, and a 15% or greater reduction in the total number of occupational injuries and illnesses. (page 2-1). A DOE-HQ review team representing the Office of Envirorunent, Safety and Health (EH), visited the Hanford Site during several weeks of the quarter. Ile 40-member Safety Management Evaluation Team (SMET) assessed WHC in the areas of management responsibility, comprehensive requirements, and competence commensurate with responsibility. As part of their new approach to oversight, they focused on the existence of management systems and programs (comparable approach to VPP). Plant/project areas selected for review within WHC were PFP, B Plant/WESF, Tank Farms, and K-Basins (page 2-2). Effective safety meetings, prejob safety meetings, etc., are a cornerstone of any successful safety program. In an effort to improve the reporting of safety meetings, the Safety/Security Meeting Report form was revised. It now provides a mechanism for recording and tracking safety issues (page 2-4). WHC has experienced an increase in the occupational injury and illness incidence rates during the first quarter of CY 1996. Trends show this increase can be partially attributed to inattention to workplace activities due 0999to the uncertainty Hanford employees currently face with recent reduction of force, reorganization, and reengineering efforts (page 2-7). The cumulative CY 1995 lost/restricted workday case incidence rate for the first quarter of CY 1996 (1.28) is 25% below the DOE CY 1991-93 average (1.70). However, the incidence rate increased 24% from the CY 1995 rate of 1.03 (page 2-8). The

Evaluation of Four Bedside Test Systems for Card Performance, Handling and Safety.

Science.gov (United States)

Giebel, Felix; Picker, Susanne M; Gathof, Birgit S

2008-01-01

SUMMARY: OBJECTIVE: Pretransfusion ABO compatibility testing is a simple and required precaution against ABO-incompatible transfusion, which is one of the greatest threats in transfusion medicine. While distinct agglutination is most important for correct test interpretation, protection against infectious diseases and ease of handling are crucial for accurate test performance. Therefore, the aim of this study was to evaluate differences in test card design, handling, and user safety. DESIGN: Four different bedside test cards with pre-applied antibodies were evaluated by 100 medical students using packed red blood cells of different ABO blood groups. Criteria of evaluation were: agglutination, labelling, handling, and safety regarding possible user injuries. Criteria were rated subjectively according to German school notes ranging from 1 = very good to 6 = very bad/insufficient. RESULTS: Overall, all cards received very good/good marks. The ABO blood group was identified correctly in all cases. Three cards (no. 1, no. 3, no. 4) received statistically significant (p labelling (1.5 vs. 2.2-2.4), handling (1.9-2.0 vs. 2.5), and user safety (2.5 vs. 3.4). Analysis of card self-explanation revealed no remarkable differences. CONCLUSION: Despite good performance of all card systems tested, the best results when including all criteria evaluated were obtained with card no. 4 (particularly concerning clear agglutination), followed by cards no. 2, no. 1, and no. 3.

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

Science.gov (United States)

Wahl, Ray

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

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)

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

Science.gov (United States)

2010-10-01

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

Engineering Solutions to Enhance Traffic Safety Performance on Two-Lane Highways

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Lina Wu

2015-01-01

Full Text Available Improving two-lane highway traffic safety conditions is of practical importance to the traffic system, which has attracted significant research attention within the last decade. Many cost-effective and proactive solutions such as low-cost treatments and roadway safety monitoring programs have been developed to enhance traffic safety performance under prevailing conditions. This study presents research perspectives achieved from the Highway Safety Enhancement Project (HSEP that assessed safety performance on two-lane highways in Beijing, China. Potential causal factors are identified based on proposed evaluation criteria, and primary countermeasures are developed against inferior driving conditions such as sharp curves, heavy gradients, continuous downgrades, poor sight distance, and poor clear zones. Six cost-effective engineering solutions were specifically implemented to improve two-lane highway safety conditions, including (1 traffic sign replacement, (2 repainting pavement markings, (3 roadside barrier installation, (4 intersection channelization, (5 drainage optimization, and (6 sight distance improvement. The effectiveness of these solutions was examined and evaluated based on Empirical Bayes (EB models. The results indicate that the proposed engineering solutions effectively improved traffic safety performance by significantly reducing crash occurrence risks and crash severities.

Impact of individual resilience and safety climate on safety performance and psychological stress of construction workers: A case study of the Ontario construction industry.

Science.gov (United States)

Chen, Yuting; McCabe, Brenda; Hyatt, Douglas

2017-06-01

The construction industry has hit a plateau in terms of safety performance. Safety climate is regarded as a leading indicator of safety performance; however, relatively little safety climate research has been done in the Canadian construction industry. Safety climate may be geographically sensitive, thus it is necessary to examine how the construct of safety climate is defined and used to improve safety performance in different regions. On the other hand, more and more attention has been paid to job related stress in the construction industry. Previous research proposed that individual resilience may be associated with a better safety performance and may help employees manage stress. Unfortunately, few empirical research studies have examined this hypothesis. This paper aims to examine the role of safety climate and individual resilience in safety performance and job stress in the Canadian construction industry. The research was based on 837 surveys collected in Ontario between June 2015 and June 2016. Structural equation modeling (SEM) techniques were used to explore the impact of individual resilience and safety climate on physical safety outcomes and on psychological stress among construction workers. The results show that safety climate not only affected construction workers' safety performance but also indirectly affected their psychological stress. In addition, it was found that individual resilience had a direct negative impact on psychological stress but had no impact on physical safety outcomes. These findings highlight the roles of both organizational and individual factors in individual safety performance and in psychological well-being. Construction organizations need to not only monitor employees' safety performance, but also to assess their employees' psychological well-being. Promoting a positive safety climate together with developing training programs focusing on improving employees' psychological health - especially post-trauma psychological

Identifying performance gaps in hydrogen safety sensor technology for automotive and stationary applications

International Nuclear Information System (INIS)

Boon-Brett, L.; Bousek, J.; Black, G.; Moretto, P.; Castello, P.; Huebert, T.; Banach, U.

2010-01-01

A market survey has been performed of commercially available hydrogen safety sensors, resulting in a total sample size of 53 sensors from 21 manufacturers. The technical specifications, as provided by the manufacturer, have been collated and are displayed herein as a function of sensor working principle. These specifications comprise measuring range, response and recovery times, ambient temperature, pressure and relative humidity, power consumption and lifetime. These are then compared against known performance targets for both automotive and stationary applications in order to establish in how far current technology satisfies current requirements of sensor end users. Gaps in the performance of hydrogen sensing technologies are thus identified and areas recommended for future research and development. (author)

Transient performance analysis of pressurized safety injection tank with a partition

International Nuclear Information System (INIS)

Bae, Youngmin; Kim, Young In; Kim, Keung Koo

2015-01-01

Highlights: • Functional performance of safety injection tanks with a partition is evaluated. • Effects of key design parameters are scrutinized. • Distinctive features of the flow in multi-unit safety injection tanks are explored. - Abstract: A parametric study has been performed to evaluate the functional performance of a pressurized multi-unit safety injection tank, which would be considered as one of the candidates for a passive safety injection system in a nuclear power plant. The influences of key design parameters including the orifice size, initial gas fraction, and resistance coefficients and operating condition on the injection flow rate are scrutinized with a discussion of the relevant flow features such as the choked flow of gas through an orifice and two interconnected regions of differing gaseous pressure. The obtained results indicate that a multi-unit safety injection tank can passively control the injection flow rate and provide a stable safety injection over a relatively long period even in the case of drastic depressurization of a reactor coolant system

Organizing Performance Requirements For Dynamical Systems

Science.gov (United States)

Malchow, Harvey L.; Croopnick, Steven R.

1990-01-01

Paper describes methodology for establishing performance requirements for complicated dynamical systems. Uses top-down approach. In series of steps, makes connections between high-level mission requirements and lower-level functional performance requirements. Provides systematic delineation of elements accommodating design compromises.

Safety and performance of a long life lithium-thionyl chloride battery

Energy Technology Data Exchange (ETDEWEB)

Cieslak, W.R.; Street, H.K.

1992-12-31

We have developed a Li/SOCl{sub 2} ``D`` cell for applications requiring 10 to 15 years life at very low drain rates, typically less than 150 {mu}A. Maximizing cell safety and reliability, while delivering very good energy density, have been the goals of our study. We have achieved these goals by designing the cell to be application specific. The low-rate cell has been optimized to deliver up to 16 Ah at drain rates of less than 70 mA. By virtue of its low surface area, 145 cm{sub 2}, the cell has demonstrated excellent safety behavior. Safety testing has been performed on individual cells as well as on two-cell and four-cell batteries. Single cells did not vent when short-circuited. We were able to produce benign venting in a two cell string, but only when the string was partially discharged before shorting. The vent mechanism is a 300 psi rupture pressure burst disc manufactured by BS&B Safety Systems. We define benign venting as full opening of the 3/8 in. dia vent hole without deformation of the case. Material is expelled from the cell without flame, and the cell stack remains largely intact. We have not produced venting of the Sandia-designed low rate cell under any other abuse test conditions. The vent functions as an ultimate safety mechanism in the case of severe abuse, but resistance to venting under normal use and mild abuse conditions is key to the achievement of high reliability.

Safety and performance of a long life lithium-thionyl chloride battery

Energy Technology Data Exchange (ETDEWEB)

Cieslak, W.R.; Street, H.K.

1992-01-01

We have developed a Li/SOCl[sub 2] D'' cell for applications requiring 10 to 15 years life at very low drain rates, typically less than 150 [mu]A. Maximizing cell safety and reliability, while delivering very good energy density, have been the goals of our study. We have achieved these goals by designing the cell to be application specific. The low-rate cell has been optimized to deliver up to 16 Ah at drain rates of less than 70 mA. By virtue of its low surface area, 145 cm[sub 2], the cell has demonstrated excellent safety behavior. Safety testing has been performed on individual cells as well as on two-cell and four-cell batteries. Single cells did not vent when short-circuited. We were able to produce benign venting in a two cell string, but only when the string was partially discharged before shorting. The vent mechanism is a 300 psi rupture pressure burst disc manufactured by BS B Safety Systems. We define benign venting as full opening of the 3/8 in. dia vent hole without deformation of the case. Material is expelled from the cell without flame, and the cell stack remains largely intact. We have not produced venting of the Sandia-designed low rate cell under any other abuse test conditions. The vent functions as an ultimate safety mechanism in the case of severe abuse, but resistance to venting under normal use and mild abuse conditions is key to the achievement of high reliability.

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

Science.gov (United States)

2010-01-01

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

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

International Nuclear Information System (INIS)

Lipar, M.

1997-01-01

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

Building quality into performance and safety assessment software

International Nuclear Information System (INIS)

Wojciechowski, L.C.

2011-01-01

Quality assurance is integrated throughout the development lifecycle for performance and safety assessment software. The software used in the performance and safety assessment of a Canadian deep geological repository (DGR) follows the CSA quality assurance standard CSA-N286.7 [1], Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear Power Plants. Quality assurance activities in this standard include tasks such as verification and inspection; however, much more is involved in producing a quality software computer program. The types of errors found with different verification methods are described. The integrated quality process ensures that defects are found and corrected as early as possible. (author)

Health and safety plan for operations performed for the Environmental Restoration Program. Task, OU 1-03 and OU 4-10 Track 2 investigations

Energy Technology Data Exchange (ETDEWEB)

Trippet, W.A. II [IT Corp., (United States); Reneau, M.; Morton, S.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1992-04-01

This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG&G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the EPR. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

Strengthening the culture of safety and performance in nuclear installations

Energy Technology Data Exchange (ETDEWEB)

Briant, V S [GPU Nuclear, Parsippany (United States); Germann, R P [Aberdeen Center for Team Learning, Matawan (United States)

1997-07-01

In mid-1995, the International Atomic Energy Agency (IAEA) in Vienna brought together a group of safety culture experts from around the world to explore and summarize those practices they viewed as important in establishing sound safety cultures in nuclear installations. This paper will summarize key findings of the Vienna team and also expand those ideas based on related work in which the authors are engaged. The paper includes a definition of safety culture, a description of three stages of safety culture, and five key practices essential to establishing and maintaining a sound safety culture. Additionally, the authors contradicts the conventional view of safety and production as trade-offs, supporting the Vienna team`s conclusion that the principles, attitudes, and practices which bring about sustained levels of high performance are the same as those which enhance safety. Based on input from colleagues in several countries, this appears to hold true across geographical and ethnic boundaries. The authors also discuss how this information can be put to practical use to obtain an objective, measurable, and repeated assessment of the current state of the safety culture within a company, plant or work unit. With that information, leaders are then in the position to act on any of the several parameters which affect both safety and performance effectiveness. (author) 9 refs., 5 tabs.

Incorporating Workplace Injury to Measure the Safety Performance of Industrial Sectors in Taiwan

Directory of Open Access Journals (Sweden)

Li-Ting Yeh

2017-12-01

Full Text Available The severity of workplace injuries varies by industry. Information on workplace injuries can enable firms and governments to effectively improve their safety performance based on the specific contexts of each industry. Incorporating the three workplace injury rates (being wounded or ill, disability, and death, a data envelopment analysis (DEA model is developed to evaluate the safety performance of 17 industrial sectors in Taiwan. The results suggest that the Taiwanese government should pay particular attention to the mining and quarrying industry, which has the lowest safety performance. Additionally, the results provide abundant information for the Taiwanese government to design industry safety regulations in a way that may prompt firms to develop a sustainable economy by improving their health and safety practices and enhancing their overall safety performance.

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)

The safety culture change process performed in Polish research reactor MARIA

International Nuclear Information System (INIS)

Golab, Andrzej

2002-01-01

The Safety Culture Change Process Performed in research reactor MARIA is described in this paper. The essential issues fulfilled in realization of the Safety Culture Enhancement Programme are related to the attitude and behaviour of top management, co-operating groups, operational personnel, relations between the operating organization and the supervising and advising organizations. Realization of this programme is based on changing the employees understanding of safety, changing their attitudes and behaviours by means of adequate training, requalification process and performing the broad self-assessment programme. Also a high level Quality Assurance Programme helps in development of the Safety Culture. (author)

Performance and safety of hydraulic turbines

International Nuclear Information System (INIS)

Brekke, H

2010-01-01

The first part of the paper contains the choice of small turbines for run of the river power plants. Then a discussion is given on the optimization of the performance of different types of large turbines. Finally a discussion on the safety and necessary maintenance of turbines is given with special attention to bolt connections.

Risk-based safety performance indicators for nuclear power plants

International Nuclear Information System (INIS)

Chakraborty, S.; Prohaska, G.; Flodin, Y.; Grint, G.; Habermacher, H.; Hallman, A.; Isasia, R.; Melendez, E.; Verduras, E.; Karsa, Z.; Khatib-Rahbar, M.; Koeberlein, K.; Schwaeger, C.; Matahri, N.; Moravcik, I.; Tkac, M.; Preston, J.

2003-01-01

In a Concerted Action (CA), sponsored by the European Commission within its 5th Framework Program, a consortium of eleven partners from eight countries has reviewed and evaluated the application of Safety Performance Indicators (SPIs), which - in combination with other tools - can be used to monitor and improve the safety of nuclear power plants. The project was aimed at identification of methods that can be used in a risk-informed regulatory system and environment, and to exploit PSA techniques for the development and use of meaningful additional/alternative SPIs. The CA included the review of existing indicator systems, and the collection of information on the experience from indicator systems by means of a specific questionnaire. One of the most important and challenging issues for nuclear plant owners and/or regulators is to recognize early signs of deterioration in safety performance, caused by influences from management, organization and safety culture (MOSC), before actual events and/or mishaps take place. Most of the existing SPIs as proposed by various organizations are considered as 'lagging' indicators, that is, they are expected to show an impact only when a downward trend has already started. Furthermore, most of the available indicators are at a relatively high level, such that they will not provide useful information on fundamental weaknesses causing the problem in the first place. Regulators' and utilities' views on the use of a Safety Performance Indicator System have also been a part of the development of the CA. (author)

Safety performance assessment of food industry facilities using a fuzzy approach

Directory of Open Access Journals (Sweden)