Influence of Non-safety Important Component on Maintenance Rule

International Nuclear Information System (INIS)

Ju, Tae Young; Kim, Wang Bae

2016-01-01

The Maintenance Rule (MR) programs in KHNP have been implemented since Jan 2009. KHNP is currently developing MR program for new built plant which has been constructed from December 2011. It is required to utilize plant-specific probabilistic safety analysis (PSA) result as risk significant criteria to determine which components are significantly important to safety. The criteria consist of three PSA risk values which are risk reduction worth (RRW), risk achievement worth (RAW) and core damage frequency (CDF) contribution. Most safety related components are classified as high risk significant, and non-safety related components as low safety significant in MR program. This paper presents the influence of the non-safety related component which has high PSA risk value on MR program of new built plant. It is considered that safety related system has at least one or more safety functions and some non-safety functions, but non-safety system doesn't have any safety function. The safety functions are defined as three functions which are required to maintain 1) integrity of reactor coolant pressure boundary, 2) capability to shut-down the reactor and maintain it in a safe shutdown, and 3) capability to prevent or mitigate the accident that could result in potential offsite exposure. The Maintenance Rule program is developed based on PSA result. Safety functions have high risk value in PSA program and considered HSS function in MR program. On the contrary, non-safety functions are generally has low risk value in PSA program and they are determined as LSS function in MR program. The AAC DG and its supporting systems are designed as non-safety systems which mean they don't have any safety function. But, AAC DG is treated as an important measure to mitigate accident in PSA program. It is determined as HSS function in MR program because it has high risk value in PSA program. AAC DG supporting systems does not have high risk value in operating plant's PSA program

Optimal replacement policy for safety-related multi-component multi-state systems

International Nuclear Information System (INIS)

Xu Ming; Chen Tao; Yang Xianhui

2012-01-01

This paper investigates replacement scheduling for non-repairable safety-related systems (SRS) with multiple components and states. The aim is to determine the cost-minimizing time for replacing SRS while meeting the required safety. Traditionally, such scheduling decisions are made without considering the interaction between the SRS and the production system under protection, the interaction being essential to formulate the expected cost to be minimized. In this paper, the SRS is represented by a non-homogeneous continuous time Markov model, and its state distribution is evaluated with the aid of the universal generating function. Moreover, a structure function of SRS with recursive property is developed to evaluate the state distribution efficiently. These methods form the basis to derive an explicit expression of the expected system cost per unit time, and to determine the optimal time to replace the SRS. The proposed methodology is demonstrated through an illustrative example.

Unresolved safety issues summary: aqua book

International Nuclear Information System (INIS)

1983-06-01

The unresolved safety issues summary is designed to provide the management of the nuclear regulatory commission with a quarterly overview of the progress and plans for completion of generic tasks addressing unresolved safety issues reported to congress pursuant to section 210 of the Energy Reorganization Act of 1974 as amended. The schedules in this book include a milestone at the end of each action plan which represents the initiation of the implementation process both with respect to incorporation of the technical resolution in the NRC official guidance or requirements and also the application of changes to individual operating plants. The schedule for implementation will not normally be included in the task action plan(s) for the resolution of a USI since the nature and extent of the activities necessary to accomplish the implementation cannot normally be reasonably determined prior to the determination of a technical resolution. The progress and status for implementation of unresolved safety issues for which a technical resolution has been completed are reported specifically in a separate table provided in this summary

78 FR 29392 - Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied...

Science.gov (United States)

2013-05-20

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0098] Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied on for Safety AGENCY: Nuclear Regulatory Commission. ACTION... (NRC) is issuing for public comment Draft Regulatory Issue Summary (RIS) 2013-XX, ``Embedded Digital...

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

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 functions and component classification for BWR, PWR and PTR

International Nuclear Information System (INIS)

1979-01-01

The Safety Guide forms part of the IAEA programme, referred to as the NUSS programme (Nuclear Safety Standards), for establishing Codes of Practice and Safety Guides relating to thermal neutron power plants. The present Safety Guide has the following chapters: safety functions, ranking of safety functions, assignment of safety class requirements. Design requirements for structural integrity of boundaries of fluid-retaining components are also discussed

Summary of fuel safety research meeting 2004

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Hidaka, Akihide; Nakamura, Jinichi; Suzuki, Motoe; Nagase, Fumihisa; Sasajima, Hideo; Fujita, Misao; Otomo, Takashi; Kudo, Tamotsu; Amaya, Masaki; Sugiyama, Tomoyuki; Ikehata, Hisashi; Iwasaki, Ryo; Ozawa, Masaaki; Kida, Mitsuko

2004-10-01

Fuel Safety Research Meeting 2004, which was organized by the Japan Atomic Energy Research Institute, was held on March 1-2, 2004 at Toranomon Pastoral, Tokyo. The purposes of the meeting are to present and discuss the results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. The technical topics of the meeting covered the status of fuel safety research activities, fuel behavior under RIA and LOCA conditions, high burnup fuel behavior, and radionuclides release under severe accident conditions. This summary contains all the abstracts and OHP sheets presented in the meeting. (author)

Unresolved safety issues summary. Aqua Book. Volume 6, No. 3

International Nuclear Information System (INIS)

Butts, J.

1984-01-01

The unresolved safety issues summary is designed to provide the management of the Nuclear Regulatory Commission with a quarterly overview of the progress and plans for completion of generic tasks addressing unresolved safety issues reported to Congress pursuant to Section 210 of the Energy Reorganization Act of 1974 as amended. This summary utilizes data collected from the Office of Nuclear Reactor Regulation, Office of Nuclear Regulatory Research, and the national laboratories and is prepared by the Office of Nuclear Reactor Regulation

Summary of FY 1997 work related to JAPC-U.S. DOE contract study on improvement of core safety - study on GEM (III)

International Nuclear Information System (INIS)

Burke, T.M.

1998-01-01

FFTF was originally designed/constructed/operated to develop LMFBR fuels and materials. Inherent safety became a major focus of the US nuclear industry in the mid 1980's. The inherent safety characteristics of LMFBRs were recognized but additional enhancement was desired. The presentation contents are: Fast Flux Test Facility history and status; Overview of contract activities; Summary of loss of flow without scram with GEMs testing; and Summary of pump start with GEMs testing

Interoperability Assets for Patient Summary Components: A Gap Analysis.

Science.gov (United States)

Heitmann, Kai U; Cangioli, Giorgio; Melgara, Marcello; Chronaki, Catherine

2018-01-01

The International Patient Summary (IPS) standards aim to define the specifications for a minimal and non-exhaustive Patient Summary, which is specialty-agnostic and condition-independent, but still clinically relevant. Meanwhile, health systems are developing and implementing their own variation of a patient summary while, the eHealth Digital Services Infrastructure (eHDSI) initiative is deploying patient summary services across countries in the Europe. In the spirit of co-creation, flexible governance, and continuous alignment advocated by eStandards, the Trillum-II initiative promotes adoption of the patient summary by engaging standards organizations, and interoperability practitioners in a community of practice for digital health to share best practices, tools, data, specifications, and experiences. This paper compares operational aspects of patient summaries in 14 case studies in Europe, the United States, and across the world, focusing on how patient summary components are used in practice, to promote alignment and joint understanding that will improve quality of standards and lower costs of interoperability.

Safety engineering with COTS components

International Nuclear Information System (INIS)

O'Halloran, Mark; Hall, Jon G.; Rapanotti, Lucia

2017-01-01

Safety-critical systems are becoming more widespread, complex and reliant on software. Increasingly they are engineered through (COTS) (Commercial Off The Shelf) components to alleviate the spiralling costs and development time, often in the context of complex supply chains. A parallel increased concern for safety has resulted in a variety of safety standards, with a growing consensus that a safety life cycle is needed which is fully integrated with the design and development life cycle, to ensure that safety has appropriate influence on the design decisions as system development progresses. In this article we explore the application of an integrated approach to safety engineering in which assurance drives the engineering process. The paper reports on the outcome of a case study on a live industrial project with a view to evaluate: its suitability for application in a real-world safety engineering setting; its benefits and limitations in counteracting some of the difficulties of safety engineering with (COTS) components across supply chains; and, its effectiveness in generating evidence which can contribute directly to the construction of safety cases. - Highlights: • Assurance as effective driver for COTS-based safety-critical system development. • Engages stakeholders, captures requirements and provides rich traceability. • Shares appropriate safety requirements across the supply chain.

Safety-related control air systems

International Nuclear Information System (INIS)

Anon.

1977-01-01

This Standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This Standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this Standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements

Project W-030 safety class upgrade summary report

International Nuclear Information System (INIS)

Kriskovich, J.R.

1998-01-01

This document presents a summary of safety class criteria for the 241-AY/AZ Tank Farm primary ventilation system upgrade under Project W-030, and recommends acceptance of the system as constructed, based on a review of supporting documentation

Hanford Site Environmental Safety and Health Fiscal Year 2001 Budget-Risk management summary

Energy Technology Data Exchange (ETDEWEB)

REEP, I.E.

1999-05-12

The Hanford Site Environment, Safety and Health (ES&H) Budget-Risk Management Summary report is prepared to support the annual request to sites in the U.S. Department of Energy (DOE) Complex by DOE, Headquarters. The request requires sites to provide supplementary crosscutting information related to ES&H activities and the ES&H resources that support these activities. The report includes the following: (1) A summary status of fiscal year (FY) 1999 ES&H performance and ES&H execution commitments; (2)Status and plans of Hanford Site Office of Environmental Management (EM) cleanup activities; (3) Safety and health (S&H) risk management issues and compliance vulnerabilities of FY 2001 Target Case and Below Target Case funding of EM cleanup activities; (4) S&H resource planning and crosscutting information for FY 1999 to 2001; and (5) Description of indirect-funded S&H activities.

Advances in safety related maintenance

International Nuclear Information System (INIS)

2000-03-01

The maintenance of systems, structures and components in nuclear power plants (NPPs) plays an important role in assuring their safe and reliable operation. Worldwide, NPP maintenance managers are seeking to reduce overall maintenance costs while maintaining or improving the levels of safety and reliability. Thus, the issue of NPP maintenance is one of the most challenging aspects of nuclear power generation. There is a direct relation between safety and maintenance. While maintenance alone (apart from modifications) will not make a plant safer than its original design, deficient maintenance may result in either an increased number of transients and challenges to safety systems or reduced reliability and availability of safety systems. The confidence that NPP structures, systems and components will function as designed is ultimately based on programmes which monitor both their reliability and availability to perform their intended safety function. Because of this, approaches to monitor the effectiveness of maintenance are also necessary. An effective maintenance programme ensures that there is a balance between the improvement in component reliability to be achieved and the loss of component function due to maintenance downtime. This implies that the safety level of an NPP should not be adversely affected by maintenance performed during operation. The nuclear industry widely acknowledges the importance of maintenance in NPP safety and operation and therefore devotes great efforts to develop techniques, methods and tools to aid in maintenance planning, follow-up and optimization, and in assuring the effectiveness of maintenance

Configuration control during maintenance of safety related equipment

International Nuclear Information System (INIS)

Irish, C.S.

2001-01-01

Possibly the most important aspect of performing maintenance of safety related equipment is maintaining the component's original design basis. Assuring that the repaired item will perform the same safety function within the original performance and equipment qualification parameters is commonly referred to as configuration control. Maintaining configuration control of a technologically current well documented item is easy. Unfortunately, this does not describe most safety related items requiring maintenance within the global nuclear industry. Items such as motors, transformers, metal clad switchgear (low and medium voltage circuit breakers), refrigeration compressors, and electronic components (i.e. circuit boards, power supplies, regulators, etc.) which routinely require repair have been in service for twenty plus years. As a result, finding replacement parts and or material to repair the items to the original condition is becoming more and more difficult. An added difficulty is the lack of original technical documentation available on the item which is being repaired. The lack of technical documentation makes it difficult to identify replacement material and parts when the original part or material is not available. The lack of documentation also makes it difficult to test the repaired item to make sure that the original configuration has been maintained after the repair. The presentation will discuss the details of repairing various items including motors, metal clad switchgear, refrigeration compressors and power supplies and the controls which are necessary to maintain the configuration of the original item. The discussion will include the Quality Assurance and engineering necessary to identify and evaluate replacement material and parts necessary to perform repairs on safety related equipment when the original material or part is not available. Examples of repairs which required different parts or materials than the original to be used in the repair will be

Configuration control during maintenance of safety related equipment

International Nuclear Information System (INIS)

Irish, C.S.

2003-01-01

Possibly the most important aspect of performing maintenance of safety related equipment is maintaining the component's original design basis. Assuring that the repaired item will perform the same safety function within the original performance and equipment qualification parameters is commonly referred to as configuration control. Maintaining configuration control of a technologically current well documented item is easy. Unfortunately, this does not describe most safety related items requiring maintenance within the global nuclear industry. Items such as motors, transformers, metal clad switchgear (low and medium voltage circuit breakers), refrigeration compressors, and electronic components (i.e. circuit boards, power supplies, regulators, etc.) which routinely require repair have been in service for twenty plus years. As a result, finding replacement parts and or material to repair the items to the original condition is becoming more and more difficult. An added difficulty is the lack of original technical documentation available on the item which is being repaired. The lack of technical documentation makes it difficult to identify replacement material and parts when the original part or material is not available. The lack of documentation also makes it difficult to test the repaired item to make sure that the original configuration has been maintained after the repair. The presentation will discuss the details of repairing various items including motors, metal clad switchgear, refrigeration compressors and power supplies and the controls which are necessary to maintain the configuration of the original item. The discussion will include the Quality Assurance and engineering necessary to identify and evaluate replacement material and parts necessary to perform repairs on safety related equipment when the original material or part is not available. Examples of repairs which required different parts or materials than the original to be used in the repair will be

Interim summary report of the safety case 2009

International Nuclear Information System (INIS)

2010-03-01

intrinsic properties of the main components of the repository and from the understanding of their evolution gained from extensive site- and concept-specific field, laboratory and modelling studies and from studies of natural and anthropogenic analogues. For any canisters that fail over this time window, the low radionuclide calculated release rates to the biosphere and resultant annual effective doses to humans and absorbed dose rates to other species of flora and fauna imply that any radiological consequences of these releases will be negligible. Furthermore, the calculation results indicate that, in general, differences in the geometry and transport paths considered in the analyses of the KBS-3V and KBS-3H design variants have only a minor impact on calculated releases and doses. Work carried out to date indicates that a geological repository for the final disposal of spent fuel, implemented as planned at the Olkiluoto site, will conform to Finnish regulatory requirements and provide an adequate level of longterm safety. This conclusion is based on the findings of safety assessments, the systematic treatment of uncertainty in these assessments and the quality measures that have been applied in the development and application of models, data and computer codes. Plans are in place to manage remaining safety-related issues and uncertainties, as given in the report TKS-2009. In implementing TKS-2009, quality assurance measures will be applied in the various production steps of the safety case, including and tests and experiments to demonstrate the feasibility and quality of technical solutions. In this way, a comprehensive safety case will be developed to support the licensing process. (orig.)

Summary of the nuclear safety in operation

International Nuclear Information System (INIS)

2004-01-01

This summary is a collection of general information about nuclear safety of PWR type reactors exploited by EDF. Teaching aid, this work has been conceived by operators for operators, it must not be considered nor used as a doctrine document with a regulatory or prescriptive characteristic. it summarizes the great principles of nuclear safety, places them in a global approach and shows their coherence. It consists in 6 chapters and 6 annexes. The news of this edition are the chapter 2 devoted to the safety management and the annexe 6 devoted to the principal teaching coming from the feedback. At the end a glossary explains the signs and abbreviations and an index allows to find themes in the memento text from keywords. (N.C.)

Summary of fuel safety research meeting 2005

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Nakamura, Takehiko; Nagase, Fumihisa; Nakamura, Jinichi; Suzuki, Motoe; Sasajima, Hideo; Sugiyama, Tomoyuki; Amaya, Masaki; Kudo, Tamotsu; Chuto, Toshinori; Tomiyasu, Kunihiko; Udagawa, Yutaka; Ikehata, Hisashi; Kida, Mitsuko; Ikatsu, Nobuhiko; Hosoyamada, Ryuji; Hamanishi, Eizou; Iwasaki, Ryo; Ozawa, Masaaki

2006-03-01

Fuel Safety Research Meeting 2005, which was organized by the Japan Atomic Energy Agency (Establishment of the new organization in Oct. 1, 2005 integrated of JAERI and JNC) was held on March 2-3, 2005 at Toshi Center Hotel, Tokyo. The purposes of the meeting are to present and discuss the results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. The technical topics of the meting covered the status of fuel safety research activities, fuel behavior under Reactivity Initiated Accident (RIA) and Loss of coolant accident (LOCA) conditions, high fuel behavior, and radionuclide release under severe accident conditions. This summary contains all the abstracts and sheets of viewgraph presented in the meeting. (author)

77 FR 60704 - Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval...

Science.gov (United States)

2012-10-04

... regulations provide that FDA publish a quarterly list of available safety and effectiveness summaries of PMA...-0638] Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval... Administration (FDA) is publishing a list of premarket approval applications (PMAs) that have been approved. This...

75 FR 38532 - Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval...

Science.gov (United States)

2010-07-02

... summaries of safety and effectiveness data to the Division of Dockets Management (HFA-305), Food and Drug... Safety and Effectiveness Summaries for Premarket Approval Applications AGENCY: Food and Drug... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket Nos. FDA-2010-M-0068...

78 FR 35284 - Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval...

Science.gov (United States)

2013-06-12

... summaries of safety and effectiveness data to the Division of Dockets Management (HFA-305), Food and Drug... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket Nos. FDA-2013-M-0036... Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval Applications AGENCY...

76 FR 31965 - Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval...

Science.gov (United States)

2011-06-02

... summaries of safety and effectiveness data to the Division of Dockets Management (HFA-305), Food and Drug... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket Nos. FDA-2011-M-0034... Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval Applications AGENCY...

78 FR 50422 - Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval...

Science.gov (United States)

2013-08-19

... summaries of safety and effectiveness data to the Division of Dockets Management (HFA-305), Food and Drug... of Safety and Effectiveness Summaries for Premarket Approval Applications AGENCY: Food and Drug... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket Nos. FDA-2013-M-0462...

75 FR 36099 - Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval...

Science.gov (United States)

2010-06-24

... summaries of safety and effectiveness data to the Division of Dockets Management (HFA-305), Food and Drug...; Availability of Safety and Effectiveness Summaries for Premarket Approval Applications AGENCY: Food and Drug... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket Nos. FDA-2009-M-0317...

Qualification of safety-related valve actuators

International Nuclear Information System (INIS)

Anon.

1981-01-01

This Standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. It may also be used to separately qualify actuator components. This Standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

Unresolved safety issues summary. Volume 3, Number 3. Aqua book

International Nuclear Information System (INIS)

1981-01-01

The 'Unresolved Safety Issues' summary is designed to provide the management of the Nuclear Regulatory Commission with a quarterly overview of the progress and plans for completion of generic tasks addressing Unresolved Safety Issues reported to Congress pursuant to section 210 of The Energy Reorganization Act of 1974 as amended. This summary utilizes data collected from the Office of Nuclear Reactor Regulation, Office of Nuclear Regulatory Research, and the National Laboratories and is prepared by the office of Management and Program Analysis. The definition of what constitutes completion of an unresolved safety issue (USI) has recently been expanded to include the implementation of the technical resolution. This is in acknowledgement of the fact that real safety benefits occur only after the implementation has taken place. The schedules in this book include a milestone at the end of each action plan which represents the initiation of the implementation process both with respect to incorporation of the technical resolution in the NRC official guidance or requirements and also the application of changes to individual operating plants. The schedule for implementation will not normally be included in the task action plan(s) for the resolution of a USI since the nature and extent of the activities necessary to accomplish the implementation cannot normally be reasonably determined prior to the determination of a technical resolution. The progress and status for implementation of unresolved safety issues for which a technical resolution has been completed are reported specifically in a separate table provided in this summary

Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

International Nuclear Information System (INIS)

Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

1996-01-01

The rules that are currently under application to verify the acceptance of flaws in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation to reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R6 procedure for assessing the integrity of the structure. (authors). 5 refs., 5 figs

Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

International Nuclear Information System (INIS)

Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

1995-01-01

The current rules applied to verify the flaws acceptance in nuclear components rely on deterministic criteria supposed to ensure the plant safe operation. The interest in have a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation do reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R 6 procedure for assessing the structure integrity. (author). 5 refs., 5 figs., 1 tab

29 CFR 1905.41 - Summary decision.

Science.gov (United States)

2010-07-01

... OCCUPATIONAL SAFETY AND HEALTH ACT OF 1970 Summary Decisions § 1905.41 Summary decision. (a) No genuine issue... 29 Labor 5 2010-07-01 2010-07-01 false Summary decision. 1905.41 Section 1905.41 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR RULES OF...

Critical Characteristics of Radiation Detection System Components to be Dedicated for use in Safety Class and Safety Significant System

International Nuclear Information System (INIS)

DAVIS, S.J.

2000-01-01

This document identifies critical characteristics of components to be dedicated for use in Safety Significant (SS) Systems, Structures, or Components (SSCs). This document identifies the requirements for the components of the common, radiation area, monitor alarm in the WESF pool cell. These are procured as Commercial Grade Items (CGI), with the qualification testing and formal dedication to be performed at the Waste Encapsulation Storage Facility (WESF) for use in safety significant systems. System modifications are to be performed in accordance with the approved design. Components for this change are commercially available and interchangeable with the existing alarm configuration This document focuses on the operational requirements for alarm, declaration of the safety classification, identification of critical characteristics, and interpretation of requirements for procurement. Critical characteristics are identified herein and must be verified, followed by formal dedication, prior to the components being used in safety related applications

Summary of ORNL work on NRC-sponsored HTGR safety research, July 1974-September 1980

International Nuclear Information System (INIS)

Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Delene, J.G.; Harrington, R.M.; Hatta, M.; Hedrick, R.A.; Johnson, L.G.; Sanders, J.P.

1982-03-01

A summary is presented of the major accomplishments of the Oak Ridge National Laboratory (ORNL) research program on High-Temperature Gas-Cooled Reactor (HTGR) safety. This report is intended to help the nuclear Regulatory Commission establish goals for future research by comparing the status of the work here (as well as at other laboratories) with the perceived safety needs of the large HTGR. The ORNL program includes extensive work on dynamics-related safety code development, use of codes for studying postulated accident sequences, and use of experimental data for code verification. Cooperative efforts with other programs are also described. Suggestions for near-term and long-term research are presented

Structural Aging Program to evaluate continued performance of safety-related concrete structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

1994-01-01

This report discusses the Structural Aging (SAG) Program which is being conducted at the Oak Ridge National Laboratory (ORNL) for the United States Nuclear Regulatory commission (USNRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into three technical tasks: Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for continued Service Determinations. Objectives and a summary of recent accomplishments under each of these tasks are presented

Benefits of a systematic approach to maintenance for safety and safety related systems

International Nuclear Information System (INIS)

Dam, R.F.; Ayazzudin, S.; Nickerson, J.H.

2003-01-01

For safety and safety-related systems, nuclear plants have to balance the requirements of demonstrating the reliability of each system, while maintaining the system and plant availability. With the goal of demonstrating statistical reliability, these systems have extensive testing programs, which often results in system unavailability and this can impact the plant capacity. The inputs to the process are often safety and regulatory related, resulting in programs that provide a high level of scrutiny. In such cases, the value of the application of a Systematic Assessment of Maintenance (SAM) process, such as Reliability Centered Maintenance (RCM), is questioned. The special case of Standby-Safety systems was discussed in a previous paper, where it was demonstrated how SAM techniques provide useful insight into current system performance, the impact of testing on component and system reliability, and how PSA considerations can be integrated into a comprehensive Maintenance, Surveillance, and Inspection (MSI) strategy. Although the system reliability requirements are an important part of the strategy evaluation, SAM techniques provide a systematic assessment within a broader context. Testing is only one part of an overall strategy focused on ensuring that component function is maintained through a combination of monitoring technologies (including testing), predictive techniques, and intrusive maintenance strategies. Each strategy is targeted to known component degradation mechanisms. This thinking can be extended to safety and safety related systems in general. Over the past 6 years, AECL has been working with CANDU utilities in the development and implementation of a comprehensive and integrated Plant Life Management (PLiM) program. As part of developing a comprehensive plant asset management approach, SAM techniques are used to develop a technical basis that not only works towards ensuring reliable operation of plant systems, but also facilitates the optimization and

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1995-12-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1991-09-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

A plan for safety and integrity of research reactor components

International Nuclear Information System (INIS)

Moatty, Mona S. Abdel; Khattab, M.S.

2013-01-01

Highlights: ► A plan for in-service inspection of research reactor components is put. ► Section XI of the ASME Code requirements is applied. ► Components subjected to inspection and their classes are defined. ► Flaw evaluation and its acceptance–rejection criteria are reviewed. ► A plan of repair or replacement is prepared. -- Abstract: Safety and integrity of a research reactor that has been operated over 40 years requires frequent and thorough inspection of all the safety-related components of the facility. The need of increasing the safety is the need of improving the reliability of its systems. Diligent and extensive planning of in-service inspection (ISI) of all reactor components has been imposed for satisfying the most stringent safety requirements. The Safeguards Officer's responsibilities of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code ASME Code have been applied. These represent the most extensive and time-consuming part of ISI program, and identify the components subjected to inspection and testing, methods of component classification, inspection and testing techniques, acceptance/rejection criteria, and the responsibilities. The paper focuses on ISI planning requirements for welded systems such as vessels, piping, valve bodies, pump casings, and control rod-housing parts. The weld in integral attachments for piping, pumps, and valves are considered too. These are taken in consideration of safety class (1, 2, 3, etc.), reactor age, and weld type. The parts involve in the frequency of inspection, the examination requirements for each inspection, the examination method are included. Moreover the flaw evaluation, the plan of repair or replacement, and the qualification of nondestructive examination personnel are considered

Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

International Nuclear Information System (INIS)

Koponen, B.L.; Hampel, V.E.

1982-01-01

This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

Energy Technology Data Exchange (ETDEWEB)

Koponen, B.L.; Hampel, V.E.

1982-10-21

This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

Structural Aging Program approach to providing an improved basis for aging management of safety-related concrete structures

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.

1993-01-01

The Structural Aging (SAG) Program is being conducted at the Oak Ridge National Laboratory (ORNL) for the United States Nuclear Regulatory Commission (USNRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into four tasks: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for Continued Service Determinations. Objectives and a summary of recent accomplishments under each of these tasks are presented

Correlation of horizontal and vertical components of strong ground motion for response-history analysis of safety-related nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Huang, Yin-Nan, E-mail: ynhuang@ntu.edu.tw [Dept. of Civil Engineering, National Taiwan University, No. 1, Sec. 4, Roosevelt Rd., Taipei 10617, Taiwan (China); Yen, Wen-Yi, E-mail: b01501059@ntu.edu.tw [Dept. of Civil Engineering, National Taiwan University, No. 1, Sec. 4, Roosevelt Rd., Taipei 10617, Taiwan (China); Whittaker, Andrew S., E-mail: awhittak@buffalo.edu [Dept. of Civil, Structural and Environmental Engineering, MCEER, State University of New York at Buffalo, Buffalo, NY 14260 (United States)

2016-12-15

Highlights: • The correlation of components of ground motion is studied using 1689 sets of records. • The data support an upper bound of 0.3 on the correlation coefficient. • The data support the related requirement in the upcoming edition of ASCE Standard 4. - Abstract: Design standards for safety-related nuclear facilities such as ASCE Standard 4-98 and ASCE Standard 43-05 require the correlation coefficient for two orthogonal components of ground motions for response-history analysis to be less than 0.3. The technical basis of this requirement was developed by Hadjian three decades ago using 50 pairs of recorded ground motions that were available at that time. In this study, correlation coefficients for (1) two horizontal components, and (2) the vertical component and one horizontal component, of a set of ground motions are computed using records from a ground-motion database compiled recently for large-magnitude shallow crustal earthquakes. The impact of the orientation of the orthogonal horizontal components on the correlation coefficient of ground motions is discussed. The rules in the forthcoming edition of ASCE Standard 4 for the correlation of components in a set of ground motions are shown to be reasonable.

Safety-related control air systems - approved 1977

International Nuclear Information System (INIS)

Anon.

1978-01-01

This standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements

The qualification of electrical components and instrumentations relevant to safety

International Nuclear Information System (INIS)

Zambardi, F.

1989-03-01

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and peculiar aspects are discussed and the qualification by testing is taken into special account. Technical and organizational aspects related to a plant qualification program are also focused. The report ends with a look to the most significant research and development activities. (author)

On safety classification of instrumentation and control systems and their components

International Nuclear Information System (INIS)

Yastrebenetskij, M.A.; Rozen, Yu.V.

2004-01-01

Safety classification of instrumentation and control systems (I and C) and their components (hardware, software, software-hardware complexes) is described: - evaluation of classification principles and criteria in Ukrainian standards and rules; comparison between Ukrainian and international principles and criteria; possibility and ways of coordination of Ukrainian and international standards related to (I and C) safety classification

Assessment and management of aging of nuclear power plant safety-related structures

International Nuclear Information System (INIS)

Naus, D.J.; Graves, H.L. III; Ellingwood, B.R.

2003-01-01

Background information and data have been developed for improving existing and developing new methods to assist in quantifying the effects of age-related degradation on the performance of nuclear power plant (NPP) safety-related structures. Factors that can lead to age-related degradation of safety-related structures are identified and their manifestations described. Current regulatory testing and inspection requirements are reviewed and a summary of degradation experience presented. Techniques commonly used to inspect NPP concrete structures to assess and quantify age-related degradation are summarized. An approach for conduct of condition assessments of structures in NPPs is presented. Criteria, based primarily on visual indications, are provided for use in classification and assessment of concrete degradation. Materials and techniques for repair of degraded structures are noted and guidance provided on repair options available for various forms of degradation. A probabilistic methodology for condition assessment and reliability-based life prediction has been developed and applied to structures subject to combinations of structural load processes and to structural systems. The methodology has also been used to investigate optimization of in-service inspection and maintenance strategies to maintain failure probability below a specified target value as well as to minimize costs. Fragility assessments involving analytical solutions and finite-element methods have been utilized to predict the effect of aging degradation on structural component performance. (author)

Safety prediction for basic components of safety-critical software based on static testing

International Nuclear Information System (INIS)

Son, H.S.; Seong, P.H.

2000-01-01

The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

Safety surveillance of activities on nuclear pressure components in China

International Nuclear Information System (INIS)

Li Ganjie; Li Tianshu; Yan Tianwen

2005-01-01

The nuclear pressure components, which perform the nuclear safety functions, are one of the key physical barriers for nuclear safety. For the national strategy on further development of nuclear power and localization of nuclear pressure components, there still exist some problems in preparedness on the localization. As for the technical basis, what can not be overlooked is the management. Aiming at the current problems, National Nuclear Safety Administration (NNSA) has taken measures to strengthen the propagation and popularization of nuclear safety culture, adjust the review and approval policies for nuclear pressure components qualification license, establish more stringent management requirements, and enhance the surveillance of activities on nuclear pressure equipment. Meanwhile, NNSA has improved the internal management and the regulation efficiency on nuclear pressure components. At the same time, with the development and implementation of 'Rules on the Safety Regulation for Nuclear Safety Important Components' to be promulgated by the State Council of China, NNSA will complete and improve the regulation on nuclear pressure components and other nuclear equipment. (authors)

Safety prediction for basic components of safety critical software based on static testing

International Nuclear Information System (INIS)

Son, H.S.; Seong, P.H.

2001-01-01

The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, both of which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

78 FR 60290 - Availability of Masked and De-identified Non-Summary Safety and Efficacy Data; Reopening of...

Science.gov (United States)

2013-10-01

..., 2013, on the following topics: (1) What factors should be considered in masking study data (e.g., data...] Availability of Masked and De-identified Non-Summary Safety and Efficacy Data; Reopening of Comment Period... of Masked and De- identified Non-Summary Safety and Efficacy Data; Request for Comments,'' which...

Replacement cross-site transfer system project W-058 safety class upgrade summary report

International Nuclear Information System (INIS)

Schlosser, R.L.

1998-01-01

This report evaluates the design of the replacement cross-site transfer system structures, systems, and components for safety related applications as defined in the Tank Waste Remediation Systems Basis for Interim Operations

Development of safety related technology and infrastructure for safety assessment

International Nuclear Information System (INIS)

Venkat Raj, V.

1997-01-01

Development and optimum utilisation of any technology calls for the building up of the necessary infrastructure and backup facilities. This is particularly true for a developing country like India and more so for an advanced technology like nuclear technology. Right from the inception of its nuclear power programme, the Indian approach has been to develop adequate infrastructure in various areas such as design, construction, manufacture, installation, commissioning and safety assessment of nuclear plants. This paper deals with the development of safety related technology and the relevant infrastructure for safety assessment. A number of computer codes for safety assessment have been developed or adapted in the areas of thermal hydraulics, structural dynamics etc. These codes have undergone extensive validation through data generated in the experimental facilities set up in India as well as participation in international standard problem exercises. Side by side with the development of the tools for safety assessment, the development of safety related technology was also given equal importance. Many of the technologies required for the inspection, ageing assessment and estimation of the residual life of various components and equipment, particularly those having a bearing on safety, were developed. This paper highlights, briefly, the work carried out in some of the areas mentioned above. (author)

Safety classification of nuclear power plant systems, structures and components

International Nuclear Information System (INIS)

1992-01-01

The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given

Selected component failure rate values from fusion safety assessment tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, L.C.

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected Component Failure Rate Values from Fusion Safety Assessment Tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee Charles

1998-09-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

Selected component failure rate values from fusion safety assessment tasks

International Nuclear Information System (INIS)

Cadwallader, L.C.

1998-01-01

This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers

78 FR 17415 - Medical Devices; Availability of Safety and Effectiveness Summaries for Premarket Approval...

Science.gov (United States)

2013-03-21

... its decision. The regulations provide that FDA publish a quarterly list of available safety and..., FDA-2012-M-1183, and FDA-2012-M-1184] Medical Devices; Availability of Safety and Effectiveness.... SUMMARY: The Food and Drug Administration (FDA) is publishing a list of premarket approval applications...

Nuclear power plant systems, structures and components and their safety classification

International Nuclear Information System (INIS)

2000-01-01

The assurance of a nuclear power plant's safety is based on the reliable functioning of the plant as well as on its appropriate maintenance and operation. To ensure the reliability of operation, special attention shall be paid to the design, manufacturing, commissioning and operation of the plant and its components. To control these functions the nuclear power plant is divided into structural and functional entities, i.e. systems. A systems safety class is determined by its safety significance. Safety class specifies the procedures to be employed in plant design, construction, monitoring and operation. The classification document contains all documentation related to the classification of the nuclear power plant. The principles of safety classification and the procedures pertaining to the classification document are presented in this guide. In the Appendix of the guide, examples of systems most typical of each safety class are given to clarify the safety classification principles

Assessment and management of ageing of major nuclear power plant components important to safety: Metal components of BWR containment systems

International Nuclear Information System (INIS)

2000-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed toward technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific

Study on some safety-related aspects of tyre use

NARCIS (Netherlands)

Jansen, S.T.H.; Schmeitz, A.J.C.; Maas, S.; Rodarius, C.; Akkermans, L.

2014-01-01

The tyre is a key component that affects road safety. The European commission has posted a tender aimed to study what measures on a European level can be taken in relation to the use of tyres to improve road safety. The results of this study, supported by a cost benefit analyses and carried out by

Safety design guides for environmental qualification for CANDU 9

International Nuclear Information System (INIS)

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

1996-03-01

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

Ecological Safety Evaluation of Land Use in Ji’an City Based on the Principal Component Analysis

Institute of Scientific and Technical Information of China (English)

2010-01-01

According to the ecological safety evaluation index data of land-use change in Ji’an City from 1999 to 2008,positive treatment on selected reverse indices is conducted by Reciprocal Method.Meanwhile,Index Method is used to standardize the selected indices,and Principal Component Analysis is applied by using year as a unit.FB is obtained,which is related with the ecological safety of land-use change from 1999 to 2008.According to the scientific,integrative,hierarchical,practical and dynamic principles,ecological safety evaluation index system of land-use change in Ji’an City is established.Principal Component Analysis and evaluation model are used to calculate four parameters,including the natural resources safety index of land use,the socio-economic safety indicators of land use,the eco-environmental safety index of land use,and the ecological safety degree of land use in Ji’an City.Result indicates that the ecological safety degree of land use in Ji’an City shows a slow upward trend as a whole.At the same time,ecological safety degree of land-use change is relatively low in Ji’an City with the safety value of 0.645,which is at a weak safety zone and needs further monitoring and maintenance.

Modelling safety of multistate systems with ageing components

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-08

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

Modelling safety of multistate systems with ageing components

International Nuclear Information System (INIS)

Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna

2016-01-01

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

Questionnaire: involved actors in large disused components management - Summary Of Responses To The Questionnaire

International Nuclear Information System (INIS)

2012-01-01

The aim of the Questionnaire is to establish an overview of the various bodies [Actors] that have responsibilities or input to the issue of large component decommissioning. In answering the intent is to cover the overall organisation and those bits that have most relevance to large components. The answers should reflect the areas from site operations to decommissioning as well as the wider issue of disposal at another location. The Questionnaire covers the following points: 1 - What is the country (institutional) structure for decommissioning? 2 - who does what and where lie the responsibilities? 3 - Which bodies have responsibility for onsite safety regulation, discharges and disposal? 4 - Which body(s) owns the facilities? 5 - Describe the responsibilities for funding of the decommissioning plan and disposal plan. Are they one and the same body? Whilst there are differences between countries there are some common threads. Regulation is through the state though the number of regulators involved may vary. In summary, the IAEA principles concerning independence of the regulatory body are followed. Funding arrangements vary but there are plans. Similarly, ownership of facilities is a mix of state and private. Some systems require a separate decommissioning license with Spain having the clearest demarcation of responsibilities for the decommissioning phase and waste management responsibilities

Final summary report of the Nordic Nuclear Safety Research Program 1998-2001

International Nuclear Information System (INIS)

Bennerstedt, T.

2002-11-01

The results of the 1998 - 2001 NKS program are presented in the form of executive summaries, highlighting the conclusions, recommendations and other findings and results of the six projects carried out during that period. The titles of the six projects are: Risk assessment and strategies for safety (NKS/SOS-1); Reactor safety (NKS/SOS-2); Radioactive waste (NKS/SOS-3); Nuclear Emergency preparedness (NKS/BOK-1); Radiological and environmental consequences (NKS/BOK-2); Nuclear threats from Nordic surroundings (NKS/SBA-1) (ln)

Common cause failure data collection and analysis for safety-related components of TRIGA SSR-14MW Pitesti, Romania

International Nuclear Information System (INIS)

Radu, G.; Mladin, D.

2003-01-01

This paper presents a study performed on the set of common cause failures (CCF) of safety-related components of the research reactor TRIGA SSR-14 MW Pitesti. The data collected cover a period of 20 years, from 1979 to 2000. The sources of data are Shift Supervisor Reports, Work Authorizations, and Reactor Log Books. Events collected are analyzed by failure mode and degrees of failure. Qualitative analysis of root causes, coupling factors and corrective actions and quantitative analysis of CCF events are studied. The objective of this work is to develop qualitative insights in the nature of the reported events and to build a site-specific common cause events database. (author)

Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

International Nuclear Information System (INIS)

1989-04-01

This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

Summary of Tiger Team Assessment and Technical Safety Appraisal recurring concerns in the Operations Area

International Nuclear Information System (INIS)

1993-01-01

Fourteen Tiger Team Assessment and eight Technical Safety Appraisal (TSA) final reports have been received and reviewed by the DOE Training Coordination Program during Fiscal Year 1992. These assessments and appraisals included both reactor and non-reactor nuclear facilities in their reports. The Tiger Team Assessments and TSA reports both used TSA performance objectives, and list ''concerns'' as a result of their findings. However, the TSA reports categorized concerns into the following functional areas: (1) Organization and Administration, (2) Radiation Protection, (3) Nuclear Criticality Safety, (4) Occupational Safety, (5) Engineering/Technical Support, (6) Emergency Preparedness, (7) Safety Assessments, (8) Quality Verification, (9) Fire Protection, (10) Environmental Protection, and (11) Energetic Materials Safety. Although these functional areas match most of the TSA performance objectives, not all of the TSA performance objectives are addressed. For example, the TSA reports did not include Training, Maintenance, and Operations as functional areas. Rather, they included concerns that related to these topics throughout the 11 functional areas identified above. For consistency, the Operations concerns that were identified in each of the TSA report functional areas have been included in this summary with the corresponding TSA performance objective

IR-360 nuclear power plant safety functions and component classification

International Nuclear Information System (INIS)

Yousefpour, F.; Shokri, F.; Soltani, H.

2010-01-01

The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

IR-360 nuclear power plant safety functions and component classification

Energy Technology Data Exchange (ETDEWEB)

Yousefpour, F., E-mail: fyousefpour@snira.co [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of); Shokri, F.; Soltani, H. [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of)

2010-10-15

The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

Comparative analysis of quality assurance systems which effectively control, review and verify the quality of components manufactured for liquid metal cooled fast breeder reactors within the EEC

International Nuclear Information System (INIS)

Benn, L.A.

1985-01-01

Comparative analyses are made of Quality Assurance Systems, by techniques and the methodology used, for the manufacture of component parts for the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) within the EEC. Two differing alternative systems are presented in the analysis. First, a tabulated analytical treatment which analyses 14 codes and standards relating to Quality Assurance which can be applied to LMFBR's. The comparison equates equivalent clauses between codes and standards followed by an analysis of individual clauses in tabular form, the International Standard ISO 6215. A statistical summary and recommendations conclude this analysis. The second alternative system used in the comparison is a descriptive analytical method applied to 9 selected codes and standards relating to Quality Assurance based on the 13 criteria of the International IAEA Code of Practice no. 50 C.QA entitled ''Quality Assurance for Safety in Nuclear Power Plants''. An investigation is then made of the state of the art on the subject of classification of component parts bearing generally on Quality Assurance. The method of classification is segregated into General, Safety and Inspection categories. A summary of destructive and non destructive controls that may be applied during the manufacture of LMFBR components is given, together with tests that may be applied to selected components, namely Primary Tank, Secondary Sodium Pump and the Primary Cold Trap allocated to Safety Classes, 1, 2 and 3 respectively. The report concludes with a summary of typical records produced at the delivery of a component

Developing an OMERACT Core Outcome Set for Assessing Safety Components in Rheumatology Trials: The OMERACT Safety Working Group.

Science.gov (United States)

Klokker, Louise; Tugwell, Peter; Furst, Daniel E; Devoe, Dan; Williamson, Paula; Terwee, Caroline B; Suarez-Almazor, Maria E; Strand, Vibeke; Woodworth, Thasia; Leong, Amye L; Goel, Niti; Boers, Maarten; Brooks, Peter M; Simon, Lee S; Christensen, Robin

2017-12-01

Failure to report harmful outcomes in clinical research can introduce bias favoring a potentially harmful intervention. While core outcome sets (COS) are available for benefits in randomized controlled trials in many rheumatic conditions, less attention has been paid to safety in such COS. The Outcome Measures in Rheumatology (OMERACT) Filter 2.0 emphasizes the importance of measuring harms. The Safety Working Group was reestablished at the OMERACT 2016 with the objective to develop a COS for assessing safety components in trials across rheumatologic conditions. The safety issue has previously been discussed at OMERACT, but without a consistent approach to ensure harms were included in COS. Our methods include (1) identifying harmful outcomes in trials of interventions studied in patients with rheumatic diseases by a systematic literature review, (2) identifying components of safety that should be measured in such trials by use of a patient-driven approach including qualitative data collection and statistical organization of data, and (3) developing a COS through consensus processes including everyone involved. Members of OMERACT including patients, clinicians, researchers, methodologists, and industry representatives reached consensus on the need to continue the efforts on developing a COS for safety in rheumatology trials. There was a general agreement about the need to identify safety-related outcomes that are meaningful to patients, framed in terms that patients consider relevant so that they will be able to make informed decisions. The OMERACT Safety Working Group will advance the work previously done within OMERACT using a new patient-driven approach.

Fast reactor safety and related physics. Volume I. Invited papers; panels; summary

Energy Technology Data Exchange (ETDEWEB)

1976-01-01

Separate abstracts were prepared for each of the twenty invited papers included. The papers covered sessions on licensing aspects of safety design bases, safety of demonstration plants, safety aspects of large commercial fast breeders, and safety test facilities.

Contribution of NDT to the safety of pressurized components in power stations

International Nuclear Information System (INIS)

Mletzko, U.; Maier, H.J.

1994-01-01

In the eyes on the MPA Stuttgart, the nondestructive testing has a very high weight relating to the safety of pressure components in power stations (concept of basis safety). In this connection, the performance verification by NDT has a special significance. A qualification of NDT-techniques can be, indeed, executed in the initial stage at test bodies with artificial faults, known with respect to position, size and type. Even theoretical (modelling) considerations can be integrated. For a performance verification in a closer sense however, this is not sufficient. The performance verification should be effected for the overall system, composed of hardware, software and examination personnel at components having the scale of 1:1 (Full Scale) under realistic boundary conditions and given times. The components must have natural or quasi-natural faults in a certain quality. The informative value of performance verifications is considerably increased, when executed as authenitic dry runs, and when the fault state is subsequently verified by destructive (metallographical) methods. (orig.) [de

Summary of LWR safety research in the USA

International Nuclear Information System (INIS)

Murley, T.E.; Tong, L.S.; Bennett, G.L.

1977-01-01

The U.S. Nuclear Regulatory Commission's water reactor safety research program is described and the basic results are presented. The USNRC water reactor safety research program consists of five basic research areas: integrity of vessel and piping, thermal-hydraulic test, fuel rod behaviour, code development and verification, and reactor operational safety. Results from the vessel and piping integrity research have demonstrated the high safety margins in scaled vessels and the analytical procedures for calculating vessel behaviour under pressure. Non-destructive examination techniques are being improved. Work is also proceeding to define the material constituents to reduce the susceptibility of irradiation embrittlement and stress corrosion cracking. The thermal-hydraulic tests have covered the various phases of a hypothetical loss of coolant accident (LOCA) and activation of the emergency core cooling system (ECCS). These tests have led to the development of engineering correlations to describe the phenomena to further quantify the safety margins in commercial nuclear power plants. Specifically, this paper presents selected experimental data and analytical predictions from the initial tests in LOFT and SEMISCALE. Comparisons and evaluations are made between the data and analytical predictions. Significant results and conclusions are presented regarding the behaviour of emergency core cooling systems in a LOCA environment: the ability to predict LOCA-type experiments over a scaling range of thirty and the thermal-hydraulic behaviour of components such as pumps in an integral system LOCA environment. The fuel behaviour research has provided valuable information on decay heat, cladding oxidation, fuel rod behaviour and fuel metling. Both the decay heat and the cladding oxidation have been shown to be lower than assumed in the licensing evaluations. The fuel behaviour and thermo-hydraulic research is being integrated into computer codes to be used to provide additional

The qualification of electrical components and instrumentations relevant to safety

CERN Document Server

Zambardi, F

1989-01-01

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and ...

Categorization of safety related motor operated valve safety significance for Ulchin Unit 3

International Nuclear Information System (INIS)

Kang, D. I.; Kim, K. Y.

2002-03-01

We performed a categorization of safety related Motor Operated Valve (MOV) safety significance for Ulchin Unit 3. The safety evaluation of MOV of domestic nuclear power plants affects the generic data used for the quantification of MOV common cause failure ( CCF) events in Ulchin Units 3 PSA. Therefore, in this study, we re-estimated the MGL(Multiple Greek Letter) parameter used for the evaluation of MOV CCF probabilities in Ulchin Units 3 Probabilistic Safety Assessment (PSA) and performed a classification of the MOV safety significance. The re-estimation results of the MGL parameter show that its value is decreased by 30% compared with the current value in Ulchin Unit 3 PSA. The categorization results of MOV safety significance using the changed value of MGL parameter shows that the number of HSSCs(High Safety Significant Components) is decreased by 54.5% compared with those using the current value of it in Ulchin Units 3 PSA

Concrete component aging and its significance relative to life extension of nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.

1986-09-01

The objectives of this study are to (1) expand upon the work that was initiated in the first two Electric Power Research Institute studies relative to longevity and life extension considerations of safety-related concrete components in light-water reactor (LWR) facilities and (2) provide background that will logically lead to subsequent development of a methodology for assessing and predicting the effects of aging on the performance of concrete-based materials and components. These objectives are consistent with Nuclear Plant Aging Research (NPAR) Program goals: (1) to identify and characterize aging and service wear effects that, if unchecked, could cause degradation of structures, components, and systems and, thereby, impair plant safety; (2) to identify methods of inspection, surveillance, and monitoring or of evaluating residual life of structures, components, and systems that will ensure timely detection of significant aging effects before loss of safety function; and (3) to evaluate the effectiveness of storage, maintenance, repair, and replacement practices in mitigating the rate and extent of degradation caused by aging and service wear

Safety concerns related to modular/prefabricated building construction.

Science.gov (United States)

Fard, Maryam Mirhadi; Terouhid, Seyyed Amin; Kibert, Charles J; Hakim, Hamed

2017-03-01

The US construction industry annually experiences a relatively high rate of fatalities and injuries; therefore, improving safety practices should be considered a top priority for this industry. Modular/prefabricated building construction is a construction strategy that involves manufacturing of the whole building or some of its components off-site. This research focuses on the safety performance of the modular/prefabricated building construction sector during both manufacturing and on-site processes. This safety evaluation can serve as the starting point for improving the safety performance of this sector. Research was conducted based on Occupational Safety and Health Administration investigated accidents. The study found 125 accidents related to modular/prefabricated building construction. The details of each accident were closely examined to identify the types of injury and underlying causes. Out of 125 accidents, there were 48 fatalities (38.4%), 63 hospitalized injuries (50.4%), and 14 non-hospitalized injuries (11.2%). It was found that, the most common type of injury in modular/prefabricated construction was 'fracture', and the most common cause of accidents was 'fall'. The most frequent cause of cause (underlying and root cause) was 'unstable structure'. In this research, the accidents were also examined in terms of corresponding location, occupation, equipment as well as activities during which the accidents occurred. For improving safety records of the modular/prefabricated construction sector, this study recommends that future research be conducted on stabilizing structures during their lifting, storing, and permanent installation, securing fall protection systems during on-site assembly of components while working from heights, and developing training programmes and standards focused on modular/prefabricated construction.

Processing and storage of blood components: strategies to improve patient safety

Directory of Open Access Journals (Sweden)

Pietersz RNI

2015-08-01

Full Text Available Ruby NI Pietersz, Pieter F van der Meer Department of Product and Process Development, Sanquin Blood Bank, Amsterdam, the Netherlands Abstract: This review focuses on safety improvements of blood processing of various blood components and their respective storage. A solid quality system to ensure safe and effective blood components that are traceable from a donor to the patient is the foundation of a safe blood supply. To stimulate and guide this process, National Health Authorities should develop guidelines for blood transfusion, including establishment of a quality system. Blood component therapy enabled treatment of patients with blood constituents that were missing, only thus preventing reactions to unnecessarily transfused elements. Leukoreduction prevents many adverse reactions and also improves the quality of the blood components during storage. The safety of red cells and platelets is improved by replacement of plasma with preservative solutions, which results in the reduction of isoantibodies and plasma proteins. Automation of blood collection, separation of whole blood into components, and consecutive processing steps, such as preparation of platelet concentrate from multiple donations, improves the consistent composition of blood components. Physicians can better prescribe the number of transfusions and therewith reduce donor exposure and/or the risk of pathogen transmission. Pathogen reduction in cellular blood components is the latest development in improving the safety of blood transfusions for patients. Keywords: blood components, red cell concentrates, platelet concentrates, plasma, transfusion, safety 

Safety aspects of nuclear power plant component aging

International Nuclear Information System (INIS)

Conte, M.; Deletre, G.; Henry, J.Y.

1988-01-01

The safety of nuclear plants depends on the capacity of the systems they are composed to perform the functions they were designed for. The identification and understanding of phenomena liable to degrade this operational capacity thus constitute one of the safety problems for which allowance must be made at the earliest stage of a project. Aging, a natural and hence unavoidable process affecting all the components of an installation, was identified at a very early stage as being one of these phenomena. The investigation and implementation of solutions to the safety problems associated to aging make it necessary to: defining the domain in which the consequences of aging are to be evaluated, identifying the parameters involved, identifying the components sensitive to these parameters, understanding the mechanisms which govern its evolution. The results of qualification tests, and of tests and checks carried out at different stages of construction and operation, as well as allowance for operating experience, constitute the necessary basis for establishing or improving the regulatory requirements. The procedures for validating components and systems of the installation are also drawn up on the basis of these tests. Finally, the actions initiated within the scope of research and development programmes supply the additional data necessary for such validation, and provide the indispensable support for knowledge improvement

Evaluation of systems interactions in nuclear power plants: Technical findings related to Unresolved Safety Issue A-17

International Nuclear Information System (INIS)

Thatcher, D.

1989-05-01

This report presents a summary of the activities related to Unresolved Safety Issue (USI)A-17, ''Systems Interactions in Nuclear Power Plants,'' and also includes the NRC staff's conclusions based on those activities. The staff's technical findings provide the framework for the final resolution of this unresolved safety issue. The final resolution will be published later as NUREG-1229. 52 refs., 4 tabs

Seismic safety margin assessment program (Annual safety research report, JFY 2010)

International Nuclear Information System (INIS)

Suzuki, Kenichi; Iijima, Toru; Inagaki, Masakatsu; Taoka, Hideto; Hidaka, Shinjiro

2011-01-01

Seismic capacity test data, analysis method and evaluation code provided by Seismic Safety Margin Assessment Program have been utilized for the support of seismic back-check evaluation of existing plants. The summary of the program in 2010 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. Many seismic capacity tests of various snubbers were conducted and quantitative seismic capacities were evaluated. One of the emergency diesel generator partial-model seismic capacity tests was conducted and quantitative seismic capacity was evaluated. Some of the analytical evaluations of piping-system seismic capacities were conducted. 2. Analysis method for minute evaluation of component seismic response. The difference of seismic response of large components such as primary containment vessel and reactor pressure vessel when they were coupled with 3-dimensional FEM building model or 1-dimensional lumped mass building model, was quantitatively evaluated. 3. Evaluation code for quantitative evaluation of seismic safety margin of systems, structures and components. As the example, quantitative evaluation of seismic safety margin of systems, structures and components were conducted for the reference plant. (author)

Safety analysis of water cooled components inside the JET thermonuclear fusion tokamak

International Nuclear Information System (INIS)

Ageladarakis, P.; O'Dowd, N.; Papastergiou, S.

1998-04-01

The transient thermal behaviour of a number of components, installed in the vessel of the world's largest Fusion Tokamak (JET) has been examined with a theoretical model, which simulated normal operational conditions and abnormal scenarios namely: Loss of Coolant Flow; Loss of Torus Vacuum; and combinations. A number of theoretical results related to water and cryogenically cooled devices have been validated by a comprehensive experimental campaign conducted both inside the JET plasma chamber and in a test rig. The performance of water cooled components which may be subjected to boiling or freeze-up risks in case of a Loss of Water Flow event has also been analysed. Time constants of transient temperature changes were determined by the model while protective actions were prescribed in order to safeguard the equipment against associated risks. A completely automatic safety protection system has been designed on the basis of these analyses and implemented in the routine JET operation. During operation of JET the safety code reacted several times within the specified time limits and protected the relevant components during real off-normal events. (author)

Model summary report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Vahlund, Fredrik

2006-10-15

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

Model summary report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Vahlund, Fredrik

2006-10-01

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

Safety assessment of primary system components at the USNRC

Energy Technology Data Exchange (ETDEWEB)

Serpan, C Z; Chen, C Y; Taboada, A

1988-12-31

This document deals with the safety assessment in nuclear reactor components at the USNRC. The USNRC regulations and requirements concerning nuclear reactor design and operations are presented, together with guides and standards which describe how the actions should be implemented. The safety assessment relies on fracture analysis and Non Destructive Examination (NDE). (TEC).

Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

International Nuclear Information System (INIS)

Andermo, L.

1977-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

Structural evaluation of safety class components to natural phenomena loadings

International Nuclear Information System (INIS)

Conrads, T.J.

1989-01-01

This paper addresses the efforts completed at the US Department of Energy Hanford Site near Richland, Washington, to qualify structurally a number of existing safety class components in the Plutonium Finishing Plant complex. Design, fabrication, and installation of the facility occurred in the 1950s and 1960s and were based on the Uniform Building Code criteria for wind and earthquake loads. Recently the buildings were qualified to site-specific wind and seismic hazards. The methodology employed to qualify seismically the safety class components is discussed

Structural mechanics research and development for main components of chinese 300 MWe PWR NPPs: from design to life management

International Nuclear Information System (INIS)

Yao Weida; Dou Yikang; Xie Yongcheng; He Yinbiao; Zhang Ming; Liang Xingyun

2005-01-01

Qinshan Nuclear Power Plant (Unit I), is a 300 MWe prototype PWR independently developed by Chinese own efforts, from design, manufacture, construction, installation, commissioning, to operation, inspection, maintenance, ageing management and lifetime assessment. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has taken up with and involved in deeply the R and D to tackle problems of this type of reactor since very beginning in early 1970s. Structural mechanics is one of the important aspects to ensure the safety and reliability for NPP components. This paper makes a summary on role of structural mechanics for component safety and reliability assessment in different stages of design, commissioning, operation, as well as lifetime assessment on this type PWR NPPs, including Qinshan-I and Chashma-I, a sister plant in Pakistan designed by SNERDI. The main contents of the paper cover design by analysis for key components of NSSS; mechanical problems relating to safety analysis; special problems relating to pressure retaining components, such as fracture mechanics, sealing analysis and its test verifications, etc.; experimental research on flow-induced vibration; seismic qualification for components; component failure diagnosis and root cause analysis; vibration qualification and diagnosis technique; component online monitoring technique; development of defect assessment; methodology of aging management and lifetime assessment for key components of NPPs, etc. (authors)

Assessment and management of ageing of major nuclear power plant components important to safety: CANDU reactor assemblies

International Nuclear Information System (INIS)

2001-02-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring, and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs) including the Soviet designed water moderated and water cooled energy reactors (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which

Assessment and Management of ageing of major nuclear power plant components important to safety: PWR pressure vessels

International Nuclear Information System (INIS)

1999-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g., caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs; and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. Since the reports are written from a safety perspective, they do not address life or life-cycle management of the plant components, which involves the integration of

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

International Nuclear Information System (INIS)

2005-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and wear out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals

International Nuclear Information System (INIS)

1999-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant

IEEE Std 382-1980: IEEE standard for qualification of safety-related valve actuators

International Nuclear Information System (INIS)

Anon.

1992-01-01

This standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. This standard may also be used to separately qualify actuator components. This standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

School Climate: An Essential Component of a Comprehensive School Safety Plan

Science.gov (United States)

Stark, Heidi

2017-01-01

The intentional assessment and management of school climate is an essential component of a comprehensive school safety plan. The value of this preventive aspect of school safety is often diminished as schools invest resources in physical security measures as a narrowly focused effort to increase school safety (Addington, 2009). This dissertation…

Time-independent and time-dependent contributions to the unavailability of standby safety system components

International Nuclear Information System (INIS)

Lofgren, E.V.; Uryasev, S.; Samanta, P.

1997-01-01

The unavailability of standby safety system components due to failures in nuclear power plants is considered to involve a time-independent and a time-dependent part. The former relates to the component's unavailability from demand stresses due to usage, and the latter represents the component's unavailability due to standby-time stresses related to the environment. In this paper, data from the nuclear plant reliability data system (NPRDS) were used to partition the component's unavailability into the contributions from standby-time stress (i.e., due to environmental factors) and demand stress (i.e., due to usage). Analyses are presented of motor-operated valves (MOVs), motor-driven pumps (MDPs), and turbine-driven pumps (TDPs). MOVs fail predominantly (approx. 78 %) from environmental factors (standby-time stress failures). MDPs fail slightly more frequently from demand stresses (approx. 63 %) than standby-time stresses, while TDPs fail predominantly from standby-time stresses (approx. 78 %). Such partitions of component unavailability have many uses in risk-informed and performance-based regulation relating to modifications to Technical Specification, in-service testing, precise determination of dominant accident sequences, and implementation of maintenance rules

Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

International Nuclear Information System (INIS)

Olsson, S.; Andermo, L.

1980-01-01

This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

An analysis of electronic health record-related patient safety concerns

Science.gov (United States)

Meeks, Derek W; Smith, Michael W; Taylor, Lesley; Sittig, Dean F; Scott, Jean M; Singh, Hardeep

2014-01-01

Objective A recent Institute of Medicine report called for attention to safety issues related to electronic health records (EHRs). We analyzed EHR-related safety concerns reported within a large, integrated healthcare system. Methods The Informatics Patient Safety Office of the Veterans Health Administration (VA) maintains a non-punitive, voluntary reporting system to collect and investigate EHR-related safety concerns (ie, adverse events, potential events, and near misses). We analyzed completed investigations using an eight-dimension sociotechnical conceptual model that accounted for both technical and non-technical dimensions of safety. Using the framework analysis approach to qualitative data, we identified emergent and recurring safety concerns common to multiple reports. Results We extracted 100 consecutive, unique, closed investigations between August 2009 and May 2013 from 344 reported incidents. Seventy-four involved unsafe technology and 25 involved unsafe use of technology. A majority (70%) involved two or more model dimensions. Most often, non-technical dimensions such as workflow, policies, and personnel interacted in a complex fashion with technical dimensions such as software/hardware, content, and user interface to produce safety concerns. Most (94%) safety concerns related to either unmet data-display needs in the EHR (ie, displayed information available to the end user failed to reduce uncertainty or led to increased potential for patient harm), software upgrades or modifications, data transmission between components of the EHR, or ‘hidden dependencies’ within the EHR. Discussion EHR-related safety concerns involving both unsafe technology and unsafe use of technology persist long after ‘go-live’ and despite the sophisticated EHR infrastructure represented in our data source. Currently, few healthcare institutions have reporting and analysis capabilities similar to the VA. Conclusions Because EHR-related safety concerns have complex

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

International Nuclear Information System (INIS)

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

2004-01-01

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

RIBA Project - Risk-Informed approach for In-Service Inspection of Nuclear Power Plant Components. Project summary

International Nuclear Information System (INIS)

Lidbury, D.; Smith, G.

2001-12-01

The need for a European review of a Risk-Informed Approach for In-Service Inspection of Nuclear Power Plant Components (RIBA) was identified in 1998. This was as a priority item in the programme of activities conducted in the framework of the Council Resolutions of 22 July 1975 and of 18 June 1992 on the Technological Problems of Nuclear Safety. The RIBA Project was established in November 1999 as a 24-month Study Contract funded by the European Commission within the frame of the former DG XI WGCS (Working Group on Codes and Standards). The Study Contract was subsequently managed for the EC by DG TREN. The participants in RIBA were Serco Assurance (project coordinator), Ringhals AB, EDF, Tecnatom SA and Westinghouse Electric Europe. The work is presented in a summary report with the detailed results contained in three companion reports as follows: main conclusions and recommendations, Review of Existing Risk-Informed Methodologies, A Comparative Study of Risk-Informed In-Service Inspection Applications, Conclusions and Recommendations for Risk-Informed in-service inspection methodology applied to Nuclear Power Plants in Europe. (author)

Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

International Nuclear Information System (INIS)

2005-10-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and ware out of components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of heavy water moderated reactors (HWRs), boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues

CSNI Status summary on utilization of best-estimate methodology in safety analysis and licensing

International Nuclear Information System (INIS)

1996-10-01

The PWG 2 Task Group on Thermal Hydraulic System Behavior has discussed the subject of the use of best-estimate codes in the licensing process (codes that model thermal hydraulic processes are important to assessing safety system performance). The Task Group set out to determine the prevailing practices in member countries, concerning safety assessment and safety review of transients affecting the reactor coolant system. A summary of information provided by member countries in response to eleven questions is given: Who is Responsible for Safety Analysis? Who is Responsible for Review and Evaluation of Safety Analysis? Do the Regulations Permit the use of Best-Estimate Codes? What are the Requirements for What Constitutes a Best Estimate Code? What are the Requirements Concerning Code Documentation? What are the Requirements for Review of Code Models and Correlations? What are the Requirements Concerning Code Assessment? What are the Requirements Concerning Initial and Boundary Conditions? What are the Requirements Concerning Operability of Active Equipment? What are the Requirements Concerning Operator Actions?

Seismic and tsunami safety margin assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

Seismic and tsunami safety margin assessment

International Nuclear Information System (INIS)

2013-01-01

Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

Summary of tank information relating salt well pumping to flammable gas safety issues

International Nuclear Information System (INIS)

Caley, S.M.; Mahoney, L.A.; Gauglitz, P.A.

1996-09-01

The Hanford Site has 149 single-shell tanks (SSTs) containing radioactive wastes that are complex mixes of radioactive and chemical products. Active use of these SSTs was phased out completely by November 1980, and the first step toward final disposal of the waste in the SSTs is interim stabilization, which involves removing essentially all of the drainable liquid from the tank. Stabilization can be achieved administratively, by jet pumping to remove drainable interstitial liquid, or by supernatant pumping. To date, 116 tanks have been declared interim stabilized; 44 SSTs have had drainable liquid removed by salt well jet pumping. Of the 149 SSTs, 19 are on the Flammable Gas Watch List (FGWL) because the waste in these tanks is known or suspected, in all but one case, to generate and retain mixtures of flammable gases, including; hydrogen, nitrous oxide, and ammonia. Salt well pumping to remove the drainable interstitial liquid from these SSTs is expected to cause the release of much of the retained gas, posing a number of safety concerns. The scope of this work is to collect and summarize information, primarily tank data and observations, that relate salt well pumping to flammable gas safety issues. While the waste within FGWL SSTs is suspected offering flammable gases, the effect of salt well pumping on the waste behavior is not well understood. This study is being conducted for the Westinghouse Hanford Company as part of the Flammable Gas Project at the Pacific Northwest National Laboratory (PNNL). Understanding the historical tank behavior during and following salt well pumping will help to resolve the associated safety issues

Development of safety factors to be used for evaluation of cracked nuclear components

International Nuclear Information System (INIS)

Brickstad, B.; Bergman, M.

1996-10-01

A modified concept for safety evaluation is introduced which separately accounts for the failure mechanisms fracture and plastic collapse. For application on nuclear components a set of safety factors are also proposed that retain the safety margins expressed in ASME, section III and XI. By performing comparative studies of the acceptance levels for surface cracks in pipes and a pressure vessel, it is shown that some of the anomalies connected with the old safety procedures are removed. It is the authors belief that the outlined safety evaluation procedure has the capability of treating cracks in a consistent way and that the procedure together with the proposed safety factors fulfill the basic safety requirements for nuclear components. Hopefully, it is possible in the near future to develop a probabilistic safety assessment procedure in Sweden, which enables a systematic treatment of uncertainties in the involved data. 14 refs

Safety and reliability of pressure components with special emphasis on advanced methods of NDT. Vol. 2

International Nuclear Information System (INIS)

1986-01-01

The 12 papers discuss topics of strength and safety in the field of materials technology and engineering. Conclusions for NPP component safety and materials are drawn. Measurements and studies relate to fracture mechanics methods (oscillation, burst, material strength, characteristics). The dynamic analysis of the behaviour of large test specimens, the influence of load velocity on crack resistance curve and the development of forged parts from austenitic steel for fast breeder reactors are presented. (DG) [de

Aspects of safety assessments for package with additional equipment components

International Nuclear Information System (INIS)

Reiche, I.; Boerst, F.M.; Krietsch, T.

2004-01-01

Many paragraphs in TS-R-1 contain the terms ''package'' or ''packaging''. These terms are defined in TS-R-1 paras 230 and 231 and explained in TS-G-1.1 paras 230.1 - 230.6. The importance of a consistent understanding of these definitions has been shown by recent discussions during the assessment of applications for package design approval. There was disagreement, if equipment components attached to the container body during transport, e.g. a transport frame, should be considered part of the package and taken into account in the safety assessment for the package. Discussions were also caused by the way inner design components are treated in the safety assessment of the package. This paper summarises the regulatory requirements to such additional equipment components and presents the way of their inclusion into the package design approval process in Germany

Inventory of Federal energy-related environment and safety research for FY 1978. Volume II. Project listings and indexes

Energy Technology Data Exchange (ETDEWEB)

None

1979-12-01

This volume contains summaries of FY-1978 government-sponsored environment and safety research related to energy. Project summaries were collected by Aerospace Corporation under contract with the Department of Energy, Office of Program Coordination, under the Assistant Secretary for Environment. Summaries are arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each agency. Information about the projects is included in the summary listings. This includes the project title, principal investigators, research organization, project number, contract number, supporting organization, funding level if known, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in Volume IV.

Inventory of Federal energy-related environment and safety research for FY 1978. Volume II. Project listings and indexes

International Nuclear Information System (INIS)

1979-12-01

This volume contains summaries of FY-1978 government-sponsored environment and safety research related to energy. Project summaries were collected by Aerospace Corporation under contract with the Department of Energy, Office of Program Coordination, under the Assistant Secretary for Environment. Summaries are arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each agency. Information about the projects is included in the summary listings. This includes the project title, principal investigators, research organization, project number, contract number, supporting organization, funding level if known, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in Volume IV

THE MAIN COMPONENTS OF SAFETY CULTURE IN AVIATION

OpenAIRE

Шостак, Оксана Григорівна; Пришупа, Юлія Юріївна

2012-01-01

The purpose of the article is to summarize, analyse and integrate the numerous reports and studies that have been conducted to define and assess safety culture, as well as the highly related concept of safety climate. This article will enable researchers and safety professionals to better understand and assess safety culture and that it will facilitate the sharing of information and strategies for improving safety culture across organizations and industries.

Safety-related decision making at a nuclear power plant

International Nuclear Information System (INIS)

Vaurio, J.K.

1998-01-01

The decision making environment of an operating nuclear power plant is presented. The organizations involved, their roles and interactions as well as the main influencing factors and decision criteria are described. The focus is on safety-related decisions, and the framework is based on the situation at Loviisa power station. The role of probabilistic safety assessment (PSA) is illustrated with decisions concerning plant modifications, optimization, acceptance of temporary configurations and extended repair times. Suggestions are made for rational and flexible risk-based control of allowed times to operate the plant with some components out of service. (orig.)

Research program Integrity of Components (FKS). A substantial contribution to component safety

International Nuclear Information System (INIS)

Kussmaul, K.; Roos, E.; Foehl, J.

1998-01-01

The main objectives pursued are: (a) verify the quality of reactor pressure vessels in existing LWR-type reactors, and (b) quantify the safety margin using both specified and non-specified materials and welds. On the basis of knowledge obtained through earlier programmes, the research project was to examine in particular deviations from the specified materials properties, for more exact quantification of the safety margin before RPV failure. There are three major factors influencing the component performance until failure, which are aggregate material fatigue, flaws, loading conditions, and the research work was to focus on the materials properties. An item of main interest was to assess the impact of long service life on the materials properties, assuming particularly unfavourable boundary conditions for materials properties and operational loads. (orig./CB) [de

Assessment and management of ageing of major nuclear power plant components important to safety. Primary piping in PWRs

International Nuclear Information System (INIS)

2003-07-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of the major NPP components (caused for instance by unanticipated phenomena and by operating, maintenance or manufacturing errors) can jeopardize plant safety and also plant life. Ageing in these NPPs must therefore be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and wear out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. Since the reports are written from a safety perspective, they do not address life or life cycle management of plant components, which involves economic considerations. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), and water moderated, water cooled energy reactors (WWERs) are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues. The

Hanford Site Environment Safety and Health (ES and H) FY 1999 and FY 2000 Execution Commitment Summary

Energy Technology Data Exchange (ETDEWEB)

REEP, I.E.

1999-12-01

All sites in the U.S. Department of Energy (DOE) Complex prepare this report annually for the DOE Office of Environment, Safety and Health (EH). The purpose of this report is to provide a summary of the previous and current year's Environment, Safety and Health (ES&H) execution commitments and the S&H resources that support these activities. The fiscal year (FY) 1999 and 2000 information (Sieracki 1999) and data contained in the ''Hanford Site Environment, Safety and Health Fiscal Year 2001 Budget-Risk Management Summary'' (RL 1999) were the basis for preparing this report. Fiscal year 2000 finding of Office of Environmental Management (EM) and Office of Nuclear Energy, Science and Technology (NE) activities is based on the President's budget of $1,065.1 million and $28.0 million, plus $2.7 million carryover finding, respectively, as of October 31, 1999. Any funding changes as a result of the Congressional appropriation process will be reflected in the Fiscal Year 2002 ES&H Budget-Risk Management Summary to be issued in May 2000. This report provides the end-of-year status of FY 1999 ES&H execution commitments, including actual S&H expenditures, and describes planned FY 2000 ES&H execution commitments and the S&H resources needed to support those activities. This requirement is included in the ES&H ''Guidance for FY200l Budget Formulations and Execution'' (DOE 1999).

Application of the Safety Classification of Structures, Systems and Components in Nuclear Power Plants

International Nuclear Information System (INIS)

2016-04-01

This publication describes how to complete tasks associated with every step of the classification methodology set out in IAEA Safety Standards Series No. SSG-30, Safety Classification of Structures, Systems and Components in Nuclear Power Plants. In particular, how to capture all the structures, systems and components (SSCs) of a nuclear power plant to be safety classified. Emphasis is placed on the SSCs that are necessary to limit radiological releases to the public and occupational doses to workers in operational conditions This publication provides information for organizations establishing a comprehensive safety classification of SSCs compliant with IAEA recommendations, and to support regulators in reviewing safety classification submitted by licensees

Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

International Nuclear Information System (INIS)

Andermo, L.; Sundman, B.

1974-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

Safety culture in nuclear installations: Summary of an international topical meeting

International Nuclear Information System (INIS)

Carnino, A.; Derrough, M.; Weimann, G.

1996-01-01

An international topical meeting, Safety Culture in Nuclear Installations, was organized by the American Nuclear Society (ANS) Austria Local Section, cosponsored by the ANS Nuclear Reactor Safety and Human Factors Divisions in cooperation with the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (NEA/OECD) and held in Vienna April 24-28, 1995. Some 250 experts from 30 different countries and organizations took part in the 85 paper presentations and two workshops. The concept of safety culture was initially used in the first International Nuclear Safety Advisory Group (INSAG) report on the Chernobyl accident analysis report in 1986. Although some elements of safety culture have been used over the years in nuclear safety activities, the new phrase safety culture and the concept were found interesting as highlighting the 'soft' aspects of safety and as encompassing more than human errors. Unfortunately, for many years it was used more in the way of identifying lack of safety culture. Conscious of this application, INSAG further developed the safety culture concept in the INSAG 4 report: The report contains a definition, the universal aspects of safety culture, the two main components of safety culture management and individual behaviour, and performance indicators of a good safety culture. This report is now quite famous and adopted with some additions or complementary definitions by many institutes and organizations for their daily activities

Summary report on the Seismic Safety Margins Research Program

International Nuclear Information System (INIS)

Cummings, G.E.

1986-01-01

The Seismic Safety Margins Research Program (SSMRP) was a program to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. The SSMRP was the first effort to trace seismically induced failure modes in a reactor system down to the individual component level, and to take into account common-cause earthquake-induced failures at the component level. This report summarizes methods and results generated by SSMRP. The SSMRP method makes use of three computer codes, HAZARD, SMACS and SEISIM to calculate ground motion acceleration time histories, structure and component responses and failure, and radioactive release probabilities. To demonstrate the methodology, an analysis was done of the Zion Nuclear Power Plant. The median frequency of core melt was computed to be 3E-5 per year, with upper (90%) and lower (10%) bounds of 8E-4 and 6E-7 per year. The main contribution to risk came from earthquakes about 2 through 4 times the design basis earthquake level. Risk was dominated by structural and inter-building piping failures and loss of off-site power. Sensitivity studies were undertaken to test assumptions and modeling procedures relative to soil-structure interaction effects, feed-and-bleed cooling, and structural failures. Assumptions made could have an order-of-magnitude effect on core melt frequency. Also, guidelines were developed for simplifying the SSMRP method, and importance rankings were generated based on the Zion analysis. 56 refs., 6 figs

Investigation of potential fire-related damage to safety-related equipment in nuclear power plants

International Nuclear Information System (INIS)

Wanless, J.

1985-11-01

Based on a review of vendor information, fire damage reports, equipment qualification and hydrogen burn test results, and material properties, thirty-three types of equipment found in nuclear power plants were ranked in terms of their potential sensitivity to fire environments. The ranking considered both the functional requirements and damage proneness of each component. A further review of the seven top-ranked components was performed, considering the relative prevalence and potential safety significance of each. From this, relays and hand switches dominate as first choices for fire damage testing with logic equipment, power supplies, transmitters, and motor control centers as future candidates

Model summary report for the safety assessment SFR 1 SAR-08

Energy Technology Data Exchange (ETDEWEB)

2008-03-15

This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

Model summary report for the safety assessment SFR 1 SAR-08

International Nuclear Information System (INIS)

2008-03-01

This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

Current trends in degradation assesment on metallic materials of industrial components

International Nuclear Information System (INIS)

Herrera Palma, Victoria

2007-01-01

To needs to assess objectively a structural integrity analysis in nuclear and termal power-, oil- and chemical- industry system, represents a large challenge for engineer and researches related to Materials Science, equipment manufactures or users. These systems share many of their problems with regards to aging mechanism of components metallic materials, high replacement costs and increasing requirements on efficiency and safety. This paper makes an attempt to give an overview of the current trends on material damage and residual life assessment for installation of power-, oil- and chemical industry. Some of the currently existing ideas on components inspection, as an activity for damage detection are shown. A summary on mechanism of material damage and experimental techniques for their characterization is also presented. Finally, some analytical methods with wide appliance in materials damage evaluation and residual life assesment of components are described

New design of engineered safety features-component control system to improve performance and reliability

International Nuclear Information System (INIS)

Kim, S.T.; Jung, H.W.; Lee, S.J.; Cho, C.H.; Kim, D.H.; Kim, H.

2006-01-01

Full text: Full text: The Engineered Safety Features-Component Control System (ESF-CCS) controls the engineered safety features of a Nuclear Power Plant such as Solenoid Operated Valves (SOV), Motor Operated Valves (MOV), pumps, dampers, etc. to mitigate the effects of a Design Basis Accident (DBA) or an abnormal operation. ESF-CCS serves as an interface system between the Plant Protection System (PPS) and remote actuation devices. ESF-CCS is composed of fault tolerant Group Controllers GC, Loop Controllers (LC), ESF-CCS Test and Interface Processor (ETIP) and Cabinet Operator Module (COM) and Control Channel Gateway (CCG) etc. GCs in each division are designed to be fully independent triple configuration, which perform system level NSSS and BOP ESFAS logic (2-out-of-4 logic and l-out-of-2 logic, respectively) making it possible to test each GC individually during normal operation. In the existing configuration, the safety-related plant component control is part of the Plant Control System (PCS) non-safety system. For increased safety and reliability, this design change incorporates this part into the LCs, and is therefore designed according to the safety-critical system procedures. The test and diagnosis capabilities of ETIP and COM are reinforced. By means of an automatic periodic test for all main functions of the system, it is possible to quickly determine an abnormal status of the system, and to decrease the elapsed time for tests, thus effectively increasing availability. ESF-CCS consists of four independent divisions (A, B, C, and D) in the Advanced Power Reactor 1400 (APR1400). One prototype division is being manufactured and will be tested

SAFE-KBS, Substantiating the safety of systems containing knowledge-based components

International Nuclear Information System (INIS)

Mesa, E.; Jimenez, A.

1998-01-01

The overall objective of the Safe-KBS project is to develop generic development and certification methodologies that allow the introduction of knowledge-based components in safety-related applications. The expert system technology presents a set of features, such as the capability to provide the rationale for its conclusions, that may significantly contribute to the new operation support systems. Nevertheless, the use of this technology in safety-related applications is limited by the lack of recognised methodologies and standards that allow a formal demonstration of the quality and reliability of these systems, as required for obtaining the approval for their use at nuclear power plants. The development methodology is structured in three hierarchical levels: life cycle model, i.e., processes and activities constituting the life cycle, life cycle plans, i.e., tasks, and support packages, i.e., set of techniques and methods to perform certain activities or tasks. The certification methodology consists of a set of certification requirements and a certification scheme for demonstrating the compliance with these requirements. This project was developed within the European framework ESPRIT, with the collaboration of Sextant, Cise, Qualience, Ilog, Computes, DNV and Uninfo. (Author)

Safety classification of systems, structures, and components for pool-type research reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-08-15

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

Summary final report: Contract between the Japan atomic power company and the U.S. Department of Energy Improvement of core safety - study on GEM (III)

International Nuclear Information System (INIS)

Burke, T.M.; Lucoff, D.M.

1997-01-01

This report provides a summary of activities associated with the technical exchange between representatives of the Japan Atomic Power Company (JAPC) and the United States Department of Energy (DOE) regarding the development and testing of Gas Expansion Modules (GEM) at the Fast Flux Test Facility (FFTF). Issuance of this report completes the scope of work defined in the original contract between JAPC and DOE titled ''Study on Improvement of Core Safety - Study on GEM (III).'' Negotiations related to potential modification of the contract are in progress. Under the proposed contract modification, DOE would provide an additional report documenting FFTF pump start tests with GEMs and answer additional JAPC questions related to core safety with and without GEMs

ITER plasma facing components

International Nuclear Information System (INIS)

Kuroda, T.; Vieider, G.; Akiba, M.

1991-01-01

This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

Task-related component analysis for functional neuroimaging and application to near-infrared spectroscopy data.

Science.gov (United States)

Tanaka, Hirokazu; Katura, Takusige; Sato, Hiroki

2013-01-01

Reproducibility of experimental results lies at the heart of scientific disciplines. Here we propose a signal processing method that extracts task-related components by maximizing the reproducibility during task periods from neuroimaging data. Unlike hypothesis-driven methods such as general linear models, no specific time courses are presumed, and unlike data-driven approaches such as independent component analysis, no arbitrary interpretation of components is needed. Task-related components are constructed by a linear, weighted sum of multiple time courses, and its weights are optimized so as to maximize inter-block correlations (CorrMax) or covariances (CovMax). Our analysis method is referred to as task-related component analysis (TRCA). The covariance maximization is formulated as a Rayleigh-Ritz eigenvalue problem, and corresponding eigenvectors give candidates of task-related components. In addition, a systematic statistical test based on eigenvalues is proposed, so task-related and -unrelated components are classified objectively and automatically. The proposed test of statistical significance is found to be independent of the degree of autocorrelation in data if the task duration is sufficiently longer than the temporal scale of autocorrelation, so TRCA can be applied to data with autocorrelation without any modification. We demonstrate that simple extensions of TRCA can provide most distinctive signals for two tasks and can integrate multiple modalities of information to remove task-unrelated artifacts. TRCA was successfully applied to synthetic data as well as near-infrared spectroscopy (NIRS) data of finger tapping. There were two statistically significant task-related components; one was a hemodynamic response, and another was a piece-wise linear time course. In summary, we conclude that TRCA has a wide range of applications in multi-channel biophysical and behavioral measurements. Copyright © 2012 Elsevier Inc. All rights reserved.

Safety related terms for advanced nuclear plants; Terminos relacionados con la seguridad para centrales nucleares avanzadas

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety.

Usage of Commercial Grade Programmable Digital Systems in Safety Related Applications

International Nuclear Information System (INIS)

Mandic, D.

2006-01-01

This paper explains methods and conditions, which if completely and correctly fulfilled, enable an operating NPP (Nuclear Power Plant) licensed and operating in accordance with the US codes and US regulatory requirements to use a commercial grade programmable digital device (PLC - Programmable Digital Controller, digital controller, digital computer or process computer) in a safety related application in a NPP. In mid 80's, when an intensive construction cycle of the new NPPs in the U.S.A. was completed, many equipment manufacturers either disappeared from the market or they abandoned their product lines that were designed and manufactured under 10 CFR Part 50 Appendix B quality assurance program. The quality assurance as defined by 10 CFR Part 50 Appendix B comprises all those planned and systematic actions necessary to provide adequate confidence that a Structure, System or Component (SSC) will perform satisfactorily in service . The operating NPPs faced the problem related to the availability of qualified equipment, components and spare parts. The US NRC (Nuclear Regulatory Commission) recognized that problem timely (Oct. 1978 revision of 10CFR21) and required a commercial grade item to be dedicated before it could be used as a basic component. A special process named Dedication of CGI - Commercial Grade Items if conducted properly, provides reasonable assurance that a commercial grade item to be used as a basic component will perform its intended safety related function and, in this respect, is deemed equivalent to an item designed and manufactured under 10 CFR Part 50 Appendix B. After that, the Dedication of CGI has been widely used mostly for relatively simple mechanical, electrical, and IandC components and spare parts. In order to provide guidance to the dedication process, EPRI has issued two documents (EPRI NP-5652 and Supplemental Guidance for EPRI NP-5652). All nuclear power plants, which comply with the US nuclear regulatory requirements, hindered as

Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Steam Generators. 2011 Update

International Nuclear Information System (INIS)

2011-11-01

At present there are over four hundred forty operational nuclear power plants (NPPs) in IAEA Member States. Ageing degradation of the systems, structures of components during their operational life must be effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling, within acceptable limits, the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This IAEA-TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations. The current practices for the assessment of safety margins (fitness for service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuteriumuranium (CANDU) reactor, boiling water reactor (BWR), pressurized water reactor (PWR), and water moderated, water cooled energy reactor (WWER) plants are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. Since the reports are written from a safety perspective, they do not address life or life cycle management of the plant components, which involves the integration of ageing management and economic planning. The target audience of the reports consists of technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. The component addressed in the present publication is the steam

Critical Propulsion Components. Volume 1; Summary, Introduction, and Propulsion Systems Studies

Science.gov (United States)

2005-01-01

Several studies have concluded that a supersonic aircraft, if environmentally acceptable and economically viable, could successfully compete in the 21st century marketplace. However, before industry can commit to what is estimated as a 15 to 20 billion dollar investment, several barrier issues must be resolved. In an effort to address these barrier issues, NASA and Industry teamed to form the High-Speed Research (HSR) program. As part of this program, the Critical Propulsion Components (CPC) element was created and assigned the task of developing those propulsion component technologies necessary to: (1) reduce cruise emissions by a factor of 10 and (2) meet the ever-increasing airport noise restrictions with an economically viable propulsion system. The CPC-identified critical components were ultra-low emission combustors, low-noise/high-performance exhaust nozzles, low-noise fans, and stable/high-performance inlets. Propulsion cycle studies (coordinated with NASA Langley Research Center sponsored airplane studies) were conducted throughout this CPC program to help evaluate candidate components and select the best concepts for the more complex and larger scale research efforts. The propulsion cycle and components ultimately selected were a mixed-flow turbofan (MFTF) engine employing a lean, premixed, prevaporized (LPP) combustor coupled to a two-dimensional mixed compression inlet and a two-dimensional mixer/ejector nozzle. Due to the large amount of material presented in this report, it was prepared in four volumes; Volume 1: Summary, Introduction, and Propulsion System Studies, Volume 2: Combustor, Volume 3: Exhaust Nozzle, and Volume 4: Inlet and Fan/ Inlet Acoustic Team.

Safety-related parameters for the MAPLE research reactor and a comparison with the IAEA generic 10-MW research reactor

International Nuclear Information System (INIS)

Carlson, P.A.; Lee, A.G.; Smith, H.J.; Ellis, R.J.

1989-07-01

A summary is presented of some of the principle safety-related physics parameters for the MAPLE Research Reactor, and a comparison with the IAEA Generic 10-MW Reactor is given. This provides a means to assess the operating conditions and fuelling requirements for safe operation of the MAPLE Research Reactor under accepted standards

Summary report on safety objectives in nuclear power plants

International Nuclear Information System (INIS)

1989-01-01

The special Task Force on Safety Objectives of the Commission of the European Communities (CEC) Working Group on the Safety of Light Water Reactors reported in May 1983 on its review of existing overall safety objectives in nuclear power plants. Since then much relevant worlwide activity has taken place. This report reviews those activities that have taken place since 1983 in European Community Member States, including more recent Members, as well as in Sweden and Finland. The report confines itself to issues related to probabilistic safety objectives, and concludes that significant progress has been made in many areas. Mutual understanding of safety objectives is leading to a convergence of views and approaches, but it is noted that much work remains to be completed

Inventory of Federal energy-related environment and safety research for FY 1979. Volume II. Project listings and indexes

Energy Technology Data Exchange (ETDEWEB)

None

1980-12-01

This volume contains summaries of FY 1979 government-sponsored environment and safety research related to energy arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each contributing agency. Information elements included in the summary listings are project title, principal investigators, research organization, project number, contract number, supporting organization, funding level, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in the back of this volume.

Inventory of Federal energy-related environment and safety research for FY 1979. Volume II. Project listings and indexes

International Nuclear Information System (INIS)

1980-12-01

This volume contains summaries of FY 1979 government-sponsored environment and safety research related to energy arranged by log number, which groups the projects by reporting agency. The log number is a unique number assigned to each project from a block of numbers set aside for each contributing agency. Information elements included in the summary listings are project title, principal investigators, research organization, project number, contract number, supporting organization, funding level, related energy sources with numbers indicating percentages of effort devoted to each, and R and D categories. A brief description of each project is given, and this is followed by subject index terms that were assigned for computer searching and for generating the printed subject index in the back of this volume

The TRAIN-project: railway safety and the train driver information environment and work situation. A summary of the main results

Energy Technology Data Exchange (ETDEWEB)

Kecklund, L. [MTO Psychology and Swedish National Rail Administration (Sweden); Ingre, M.; Kecklund, G.; Soederstroem, M.; Aakerstedt, T. [National Inst. for Psychosocial Factors and Health (Sweden); Lindberg, E. [Swedish National Rail Administration (Sweden); Jansson, A.; Olsson, E.; Sandblad, B. [Uppsala Univ. (Sweden). Dept. of Human-Computer Interaction; Almqvist, P. [Swedish State Railways (Sweden)

2001-07-01

The TRAIN project investigates traffic safety related risks, focusing in particular on the train driver work situation, use of information but also on the supporting safety organisation. It is an on-going project funded and managed by Swedish National Rail Administration and carried out by independent researchers. The project provides a multi-disciplinary investigation by use of a man-technology-organisation (MTO) perspective. Activities performed are task analysis, evaluation of the drivers use of information and interaction with the ATP system as well as analyses of stress, mental workload and work hours. Several methods are being used such as interviews, questionnaires, diaries, activity monitoring and videotapes. This paper gives an overview of the project as well as a short summary of the main results. Detailed results are presented in separate reports as started in the reference list. Some of the main results are that the drivers report severe problems concerning sleepiness on early morning shifts, problems with maintenance on vehicles, lack of information supporting the planning task as well as problems in understanding ATP functions. Two groups of drivers having a feed-back related as opposed to a feed-forward driving style could be identified. In conclusion there is a great need to perform more scientific studies of human factors and railway safety as well as to implement safety management programs including professional human factors competence in the railway industries. (orig.)

Conference summaries

International Nuclear Information System (INIS)

1987-01-01

This volume contains summaries of 28 papers presented at the 27. conference of the Canadian Nuclear Association. These papers discuss the general situation of the Canadian nuclear industry and the CANDU reactor; dialogue with the public; the International Atomic Energy Agency; and economic goals and operating lessons. It also contains summaries of 70 papers presented at the 8. conference of the Canadian Nuclear Society, which discuss plant life extension; safety and the environment; reactor physics; thermalhydraulics; risk assessment; the CANDU spacer location and repositioning project; CANDU operations; safety research after Chernobyl; fuel channels; and nuclear technology developments. The individual papers are also available in INIS-mf--13673 (CNA), and INIS-mf--12909 (CNS). (L.L.)

Summary final report: Contract between the Japan atomic power company and the U.S. Department of Energy Improvement of core safety - study on GEM (III)

Energy Technology Data Exchange (ETDEWEB)

Burke, T.M.; Lucoff, D.M.

1997-03-18

This report provides a summary of activities associated with the technical exchange between representatives of the Japan Atomic Power Company (JAPC) and the United States Department of Energy (DOE) regarding the development and testing of Gas Expansion Modules (GEM) at the Fast Flux Test Facility (FFTF). Issuance of this report completes the scope of work defined in the original contract between JAPC and DOE titled ''Study on Improvement of Core Safety - Study on GEM (III).'' Negotiations related to potential modification of the contract are in progress. Under the proposed contract modification, DOE would provide an additional report documenting FFTF pump start tests with GEMs and answer additional JAPC questions related to core safety with and without GEMs.

Summary

International Nuclear Information System (INIS)

Roehlig, Klaus-Juergen

2014-01-01

and licensing process. This set of issues is by no means complete. For the Regulators' Forum and the IGSC it is now necessary to identify those issues and approaches to their resolutions which are of joint interest in order to address them in their programmes of work. The IGSC will, in accordance with its mandate, focus on topics related to safety case development and to the links to establish between different components of repository development. Subjects which have to be discussed and perhaps addressed in the Programme of Work include: - Operational safety: In the past, IGSC focussed on the relationship of operational and postclosure safety. A move towards questions specific for operational safety and in particular the potential for developing a list of events, incident causes etc. to be accounted for when assessing operational safety ('operational safety FEP list') will be considered. - Further attention will be devoted to establishing the linkage between the construction of engineered components and safety assessment, i.e. to the issue of feasibility to construct components according to the design specifications made by, or used in, safety assessments. - The IGSC will contribute to the EU MoDerN project in order to address issues related to monitoring and its linkage to safety demonstration. - IGSC also will further address organisational issues. (author)

The development of summary components for the Disablement in the Physically Active scale in collegiate athletes.

Science.gov (United States)

Houston, Megan N; Hoch, Johanna M; Van Lunen, Bonnie L; Hoch, Matthew C

2015-11-01

The Disablement in the Physically Active scale (DPA) is a generic patient-reported outcome designed to evaluate constructs of disability in physically active populations. The purpose of this study was to analyze the DPA scale structure for summary components. Four hundred and fifty-six collegiate athletes completed a demographic form and the DPA. A principal component analysis (PCA) was conducted with oblique rotation. Factors with eigenvalues >1 that explained >5 % of the variance were retained. The PCA revealed a two-factor structure consistent with paradigms used to develop the original DPA. Items 1-12 loaded on Factors 1 and Items 13-16 loaded on Factor 2. Items 1-12 pertain to impairment, activity limitations, and participation restrictions. Items 13-16 address psychosocial and emotional well-being. Consideration of item content suggested Factor 1 concerned physical function, while Factor 2 concerned mental well-being. Thus, items clustered around Factor 1 and 2 were identified as physical (DPA-PSC) and mental (DPA-MSC) summary components, respectively. Together, the factors accounted for 65.1 % of the variance. The PCA revealed a two-factor structure for the DPA that resulted in DPA-PSC and DPA-MSC. Analyzing the DPA as separate constructs may provide distinct information that could help to prescribe treatment and rehabilitation strategies.

Parameter Estimation of a Reliability Model of Demand-Caused and Standby-Related Failures of Safety Components Exposed to Degradation by Demand Stress and Ageing That Undergo Imperfect Maintenance

Directory of Open Access Journals (Sweden)

S. Martorell

2017-01-01

Full Text Available One can find many reliability, availability, and maintainability (RAM models proposed in the literature. However, such models become more complex day after day, as there is an attempt to capture equipment performance in a more realistic way, such as, explicitly addressing the effect of component ageing and degradation, surveillance activities, and corrective and preventive maintenance policies. Then, there is a need to fit the best model to real data by estimating the model parameters using an appropriate tool. This problem is not easy to solve in some cases since the number of parameters is large and the available data is scarce. This paper considers two main failure models commonly adopted to represent the probability of failure on demand (PFD of safety equipment: (1 by demand-caused and (2 standby-related failures. It proposes a maximum likelihood estimation (MLE approach for parameter estimation of a reliability model of demand-caused and standby-related failures of safety components exposed to degradation by demand stress and ageing that undergo imperfect maintenance. The case study considers real failure, test, and maintenance data for a typical motor-operated valve in a nuclear power plant. The results of the parameters estimation and the adoption of the best model are discussed.

Stockholm Safety Conference. Analysis of the sessions on radiological protection, licensing and risk assessment

International Nuclear Information System (INIS)

Gea, A.

1981-01-01

A summary of the sessions on radiological protection, licensing and risk assessment in the safety conference of Stockholm is presented. It is considered the new point of view of the nuclear safety, probabilistic analysis, components failures probability and accident analysis. They are included conclusions applicable in many cases to development countries. (author)

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

Energy Technology Data Exchange (ETDEWEB)

REEP, I.E.

2000-12-01

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

The use of probabilistic safety assessment (PSA) based maintenance indicators to increase the availability of safety related systems in nuclear power plants

International Nuclear Information System (INIS)

Kirchsteiger, C.

1991-04-01

This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain nuclear power plant. Plant specific equipment history and maintenance work on data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant components as well as the detection of the dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system-based and maintenance-related performance indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

Environmental qualification - walkdowns: The documentation of configuration information for safety related components, equipment and systems

International Nuclear Information System (INIS)

Melmer, J.; Waters, M.

1995-01-01

Environmental Qualification walkdowns are conducted to collect field data to verify/validate/document configurations of safety related equipment and systems. This paper describes the process for conducting walkdowns and the justification for using an electronic format. The following are described: a) Background; b) Preparing, executing and processing walkdowns; c) Hardware/software; d) Impact of a paperless system on walkdown execution, maintenance and work planning; e) Other applications for the technology

1st IAEA research coordination meeting on tritium retention in fusion reactor plasma facing components. October 5-6, 1995, Vienna, Austria. Summary report

International Nuclear Information System (INIS)

Langley, R.A.

1995-12-01

The proceedings and results of the 1st IAEA research Coordination Meeting on ''Tritium Retention in Fusion Reactor Plasma Facing Components'' held on October 5 and 6, 1995 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of a data survey and needs assessment for the retention, release and removal of tritium from plasma facing components, a summary of data evaluation, and recommendations regarding future work. (author). 4 tabs

Survey Summary

Data.gov (United States)

U.S. Department of Health & Human Services — Nursing home summary information for the Health and Fire Safety Inspections currently listed on Nursing Home Compare, including dates of the three most recent...

77 FR 62248 - Commercial Fishing Safety Advisory Committee; Meeting

Science.gov (United States)

2012-10-12

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [Docket No. USCG-2012-0936] Commercial Fishing Safety.... SUMMARY: The Commercial Fishing Safety Advisory Committee (CFSAC) will meet in Washington, DC to discuss various issues relating to safety in the commercial fishing industry. This meeting will be open to the...

78 FR 44958 - Commercial Fishing Safety Advisory Committee; Meeting

Science.gov (United States)

2013-07-25

... DEPARTMENT OF HOMELAND SECURITY Coast Guard [Docket No. USCG-2013-0625] Commercial Fishing Safety.... SUMMARY: The Commercial Fishing Safety Advisory Committee (CFSAC) will meet in Washington, DC to discuss various issues relating to safety in the commercial fishing industry. This meeting will be open to the...

Summary report for the second TUV-workshop proceedings on living PSA application

International Nuclear Information System (INIS)

1991-01-01

This workshop on living PSA Application was organized to support the OECD/NEA CSNI-Principal Working Group No.5 on Risk Assessment for an international exchange of experience on living PSA application. The first session was devoted to Living PSA Applications and the second session to Tools for Living PSA. Living PSA Applications: Reasons for performing PSA (regulatory requirement, targets; corporate requirement, targets; safety related activity prioritization; other); Logistic of Living PSA Management (Corporate management involvement, Decision making levels and guidance, Plant level involvement, Required personnel commitment, Frequency and extent of re-quantification of PSA, Types of safety/risk parameters to be monitored, Quality assurance on maintaining Living PSA); Examples of Application (Experiences of application, State of Living PSA/e.g. all accident sequences involved, Details of component level involvement). Tools for Living PSA: Data Collection Systems and Codes (Source and type of data collected, Probabilistic parameter quantification, Interface to basic event data, Data code systems). An executive summary of the workshop is given

Summary of Auger-Related Entanglement Incidents Occurring Inside Agricultural Confined Spaces.

Science.gov (United States)

Cheng, Y H; Field, W E

2016-04-01

Entanglements in energized equipment, including augers found in agricultural workplaces, have historically been a significant cause of traumatic injury. Incidents involving augers located inside agricultural confined spaces (primarily grain storage structures and forage silos), although relatively rare events, are a widely recognized problem due to the relative severity of the resulting injuries and the complexities of victim extrication. However, this problem is neither well documented nor elucidated in the research literature, other than anecdotal observations relating to medical treatment of auger-related injuries and citations for non-compliance with federal and state workplace safety regulations. A review of nearly 1,650 cases documented in the Purdue Agricultural Confined Spaces Incident Database from 1964 to 2013 identified 167 incidents involving entanglement in an energized auger that occurred while the victim was working inside an agricultural confined space. These incidents primarily included in-floor unloading augers, sweep augers, stirring augers, and auger components found on silo unloaders. Cases involving portable tube augers used to handle grain outside grain storage structures were not included. Based on analysis of the data, approximately 98% of known victims were male, with the 21-45 age group reporting the largest number of incidents. Nearly one-third (32.3%) of incidents were fatal, and lower limb amputation was the most frequently reported injury type. (It is believed that non-fatal incidents are grossly under-reported in the data set due to a lack of comprehensive reporting requirements, especially for most farms, feedlots, and seed processing operations, which are generally exempt from compliance with OSHA machine guarding, confined-space, and grain-handling standards.) The type of auger identified most frequently as the agent of injury was the exposed in-floor auger (48), which frequently resulted in amputation of one or more lower limbs

Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

International Nuclear Information System (INIS)

Tyacke, M.J.; McConnell, J.W. Jr.; Ayers, A.L. Jr.; O'Connor, S.C.; Jankovich, J.P.

1996-01-01

The Idaho National Engineering Laboratory prepared a technical report for the Office of Nuclear Material Safety and Safeguards of the US Nuclear Regulatory Commission, entitled Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, NUREG/CR-6407. This paper provides the results of that report. It also presents the graded approach for classification of components used in transportation packagings and dry spent fuel storage systems. This approach provides a method for identifying the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements. The general types of transportation packagings and dry storage systems are identified. The methodology used in this paper is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This paper also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems, with a classification category assigned to each component. Several examples concerning the safety importance of components are presented

Safety philosophy and design principles for systems and components of nuclear power plant: external event

International Nuclear Information System (INIS)

Lopes, J.P.G.

1986-01-01

In nuclear power plants, some systems and components are designed to withstand external impacts. Such systems and components are those which have to perform their functions even during and after the occurrences of an earthquake, for example, fulfilling the safety objectives and avoiding the release of radioactive material to the environment. The aim of this report is to introduce the safety philosophy and design principles for systems/components to perform their functions during and after the occurrence of an earthquake, as applied by NUCLEN for Angra 2 and 3. (Author) [pt

Structural safety - Is the safety margin measurable

International Nuclear Information System (INIS)

Rintamaa, R.

1992-01-01

In ensuring the structural safety of the nuclear components one must be aware of the uncertainties related to the material deorientation, loadings and other operational conditions, geometrical dimensions as well as the service environment. Furthermore, the validation of the analysis tools and procedures is of great importance in overall safety assessment of a pressure retaining component. In order to identify and quantify the concerns and risks arising from the uncertainties in the safety related issue intensive research is being carried out all over the world, in particular, on the ageing, plant life extension and management of old nuclear power plants. The presentation includes a general survey of the factors relevant to the assessment of safe and reliable operation of a nuclear component throughout its planned service life. Certain aspects are outlined based on the research work being carried out at the Technical Research Centre of Finland (VTT)(orig.)

Safety recommendation component of mobile information assistant of the tourist

Science.gov (United States)

Savchuk, Valeriya V.; Kunanec, Natalia E.; Pasichnyk, Volodymyr V.; Popiel, Piotr; Weryńska-Bieniasz, RóŻa; Kashaganova, Gulzhamal; Kalizhanova, Aliya

2017-08-01

The goal of article is to introduce and justify the need for the safety system components of Mobile Information of the tourist (MIAT). One of the objectives of the system is to determine the level of risk in a particular tourist destination on the basis of available information in the knowledge base.

Columbia River Component Data Evaluation Summary Report

Energy Technology Data Exchange (ETDEWEB)

C.S. Cearlock

2006-08-02

The purpose of the Columbia River Component Data Compilation and Evaluation task was to compile, review, and evaluate existing information for constituents that may have been released to the Columbia River due to Hanford Site operations. Through this effort an extensive compilation of information pertaining to Hanford Site-related contaminants released to the Columbia River has been completed for almost 965 km of the river.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Safety characteristics of options for plasma-facing components for ITER and beyond

International Nuclear Information System (INIS)

Piet, S.J.; McCarthy, K.A.; Holland, D.F.; Longhurst, G.R.; Merrill, B.J.

1991-01-01

Plasma-facing components (PFC) likely dominate the safety hazards of the International Thermonuclear Experimental Reactor (ITER) and post-ITER machines. To gain regulatory approval and for fusion energy to fulfill its ultimate attractive safety and environmental potential, safety must be considered when selecting among PFC options. This paper summarizes current PFC safety information. PFC safety issues fall into seven areas: disruption tolerance, disruption severity, tritium inventory and permeation, accidental energy release, activation/toxin hazards, cooling disturbances, and system issues. RFC options include current ITER mainline options (Be or W coating, C tiles), variants on current ITER options, and liquid metal (LM) divertors. No PFC option that we have examined is free of critical safety concerns. There are also innovative ideas that may improve any PFC's performance -- super-permeable vacuum ducts, helium self-pumping, and gaseous divertors. We conclude with recommendations and a future strategy. 17 refs., 1 fig., 3 tabs

A Study on the Organizational Components Affecting the Communication-Related Events in Nuclear Power Plant

International Nuclear Information System (INIS)

Lee, Seung Min; Jang, In Seok; Seong, Poong Hyun

2009-01-01

It is important to communicate clearly and effectively in order to achieve and improve team performance, also in the view point of safety, in nuclear power plant (NPP). Researchers have studied on lots of accidents and incidents related to communication and analyzed the elements affecting communication fail in the side of sender-receiver communication process so that they have found which process was failed to communicate each other. But we cannot disregard on human cognition, level of understanding, and individual or team characteristic on the communication process, so we need to analyze the elements of communication-related events in the side of human and team components that we will find why operators could not avoid failing their communication. In this paper we enumerate key organizational components, collect events related to communication in NPP and count the total number of components affecting communication fail. Finally we perform the pairwise-comparison using those values and understand major factors affecting communication-related events

Use of expert systems in the structural safety assessment of of pressurized nuclear components

International Nuclear Information System (INIS)

Jovanovic, A.; Sturm, D.

1990-01-01

The paper describes research currently performed at MPA Stuttgart on development of expert systems and application of artificial intelligence methods and techniques, for structural safety assessment of power plant pressurized components. The research is done as an extension of preceding and existing large research programs of MPA, in the domain of structural safety of components. In this preceding research a waste amount of practical engineering knowledge and experience has been accumulated: development in the direction of AI-based systems is a way to use this knowledge more efficiently in future research and in the nuclear power plant practice. Applications on which the current research is focussed are expert systems applied for the leak-before-break analysis for the structural safety evaluation in high temperature regimes

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.

Criteria for safety-related operator actions

International Nuclear Information System (INIS)

Gray, L.H.; Haas, P.M.

1983-01-01

The Safety-Related Operator Actions (SROA) Program was designed to provide information and data for use by NRC in assessing the performance of nuclear power plant (NPP) control room operators in responding to abnormal/emergency events. The primary effort involved collection and assessment of data from simulator training exercises and from historical records of abnormal/emergency events that have occurred in operating plants (field data). These data can be used to develop criteria for acceptability of the use of manual operator action for safety-related functions. Development of criteria for safety-related operator actions are considered

Optimized maintenance concept of safety relevant valves related to ageing management features in nuclear power plants

International Nuclear Information System (INIS)

Koring, R.

2007-01-01

This paper presents the existing concept in E.ON Kernkraft and its sound application to ageing management issues by focussing on group 2 components such as safety relevant valves. It is demonstrated how the maintenance concept of safety relevant valves is supported by a valve diagnostic system accompanied by an applied procedure to assess the measured results with respect to the required functionality and ageing phenomena. Furthermore this concept has been developed to optimize the existing preventive maintenance of the safety relevant valves by implementing condition oriented aspects derived from the diagnostic results. The main issue of this maintenance concept is to demonstrate the high level of the secured function, reliability and performance of the safety relevant valves within an integrated ageing management. Additionally it offers improvements of all preventive maintenance issues as maintenance periods and the component related volume, spare parts management and costs. (author)

Evaluation of Adverse Events in Total Disc Replacement: A Meta-Analysis of FDA Summary of Safety and Effectiveness Data.

Science.gov (United States)

Anderson, Paul A; Nassr, Ahmad; Currier, Bradford L; Sebastian, Arjun S; Arnold, Paul M; Fehlings, Michael G; Mroz, Thomas E; Riew, K Daniel

2017-04-01

Systematic review and meta-analysis. The safety of new technology such as cervical total disc replacement (TDR) is of paramount importance and is best evaluated in randomized clinical trials (RCT). We compared complication risks of TDR to fusion using data from Investigational Device Exemptions. A systematic review of FDA Summary of Safety and Effectiveness reports of the 8 approved cervical TDRs was performed. These were all randomized controlled trials comparing anterior cervical discectomy and fusion (ACDF) to TDR. Important outcome variables were dysphagia, wound infection, neurologic injuries, heterotopic ossification, death, and secondary surgeries. A random effects model was selected a priori. Data on adverse events was abstracted and analyzed by calculating relative risk of ACDF to TDR by meta-analysis techniques. The study included 3027 patients with 1377 randomized to ACDF and 1652 to TDR. No statistical differences were present between the 2 groups in dysphagia/dysphonia, hardware related, heterotopic ossification, death, and overall neurologic adverse events and incidence of neurologic deterioration. The relative risk of wound-related problems ACDF to TDR was 0.76 (95% confidence interval [CI] = 0.59, 0.98) favoring ACDF, which was statistically significant, but these were minor and never required a second surgical procedure for deep wound infection. The relative risk of ACDF to TDR in surgical-related neurologic events and secondary surgeries was 1.62 (95% CI = 1.04, 2.53) and 1.79 (95% CI = 1.17, 2.74), both favoring TDR. Cervical TDR appears to be as safe as or safer than ACDF at 2-year follow-up.

Conference summaries

International Nuclear Information System (INIS)

1988-01-01

This volume contains conference summaries of the 28. annual conference of the Canadian Nuclear Association, and the 9. annual conference of the Canadian Nuclear Society. Topics of discussion include: power reactors; fuel cycles; nuclear power and public understanding; future trends; applications of nuclear technology; CANDU reactors; operational enhancements; design of small reactors; accident behaviour in fuel channels; fuel storage and waste management; reactor commissioning/decommissioning; nuclear safety experiments and modelling; the next generation reactors; advances in nuclear engineering education in Canada; safety of small reactors; current position and improvements of fuel channels; current issues in nuclear safety; and radiation applications - medical and industrial

Development of flaw assesment methodology for elevated temperature components of FBR plants

International Nuclear Information System (INIS)

Shimakawa, Takashi; Takahashi, Yukio; Miura, Naoki; Nakayama, Yasunari; Sawai, Tatsuaki; Tooya, Yuuji

1999-01-01

Fracture mechanics is applicable for the safety assessment of FBR component if a crack is assumed to exist. Inelastic response should be taken into account due to high temperature operation of FBR components. However, methodology for the application of inelastic fracture mechanics has not been established sufficiently. CRIEPI has been conducted research projects to develop a flaw assessment guideline for FBR components. This guideline consists of evaluation methods for creep-fatigue crack propagation, ductile fracture and sodium leak rate. The summary of evaluation methods on creep-fatigue crack and ductile fracture is presented in this paper. (author)

Radiation safety aspects during nondestructive testing of reactor shielding components by gamma radiometry

International Nuclear Information System (INIS)

Viswanathan, S.; Jose, M.T.; Venkatraman, B.

2016-01-01

In nuclear facilities, effective shielding of radioactive components and structures are essential to ensure radiation protection to operating personnel. The shield structures are made of lead, steel and concrete with varying thickness of up to 1200 mm. It needs to be verified for shielding integrity, presence of voids, blowholes and defects to avoid exposure to workers and to public at large. Radiometry using gamma source serves as excellent tool for non-destructive examination of such structures and components. Gamma sources of high activity up to 50 Curies (gamma camera type) depending on the thickness of component have to be used. During the testing exposure to the operating personnel needs to be minimized, this requires certain safety procedures to be followed. This paper focuses the methodology to be adapted by means of selection of source, effective training of personnel, compliance with safety requirements and maintenance of source devices

Developing an OMERACT Core Outcome Set for Assessing Safety Components in Rheumatology Trials

DEFF Research Database (Denmark)

Klokker, Louise; Tugwell, Peter; Furst, Daniel E

2016-01-01

in such COS. The Outcome Measures in Rheumatology (OMERACT) Filter 2.0 emphasizes the importance of measuring harms. The Safety Working Group was reestablished at the OMERACT 2016 with the objective to develop a COS for assessing safety components in trials across rheumatologic conditions. METHODS: The safety......OBJECTIVE: Failure to report harmful outcomes in clinical research can introduce bias favoring a potentially harmful intervention. While core outcome sets (COS) are available for benefits in randomized controlled trials in many rheumatic conditions, less attention has been paid to safety...... that patients consider relevant so that they will be able to make informed decisions. CONCLUSION: The OMERACT Safety Working Group will advance the work previously done within OMERACT using a new patient-driven approach....

Synergistic behaviour of nuclear radiation, temperature-humidity extremes and LOCA situation on safety and safety-related equipment in Indian nuclear power plants

International Nuclear Information System (INIS)

Kulkarni, R.D.; Bora, J.S.; Prakash, Ravi; Agarwal, Vivek; Sundersingh, V.P.

2002-01-01

Full text: The general philosophy for the instrumentation in nuclear power plants is based on the use of equipment/instruments which are capable of continuous satisfactory operation over a long period of time with minimum attention. Long term reliability under varying service conditions is of prime importance. The reliability of nuclear power plant depends on the reliability of safety and safety-related electronic instruments/ equipment used for performing the crucial tasks. The electrical and electronic systems/ circuits/ components of the equipment used in reactor safety systems (e.g. reactor protection system, emergency core cooling system, etc.) and reactor safety-related systems (e.g. reactor containment isolation and cooling system, reactor shutdown system, etc.) are responsible for safe and reliable operation of a nuclear power plant. The performance of reactor safety and safety-related equipment/instruments viz. pressure and differential pressure transmitter, amplifier for ion chamber, etc. has been evaluated under synergistic atmosphere including LOCA to find out the critical link in the circuits and subsequent modifications are suggested. The mathematical representation of the generated database has been done to estimate the life span of the instruments and accordingly the guidelines has been prepared for the operational staff to avoid the forced outage of the plant. All the details are included and mathematical models are presented to predict the future performances

Safety functions and safety function indicators - key elements in SKB'S methodology for assessing long-term safety of a KBS-3 repository

International Nuclear Information System (INIS)

Hedin, A.

2008-01-01

The application of so called safety function indicators in SKB safety assessment of a KBS-3 repository for spent nuclear fuel is presented. Isolation and retardation are the two main safety functions of the KBS-3 concept. In order to quantitatively evaluate safety on a sub-system level, these functions need to be differentiated, associated with quantitative measures and, where possible, with quantitative criteria relating to the fulfillment of the safety functions. A safety function is defined as a role through which a repository component contributes to safety. A safety function indicator is a measurable or calculable property of a repository component that allows quantitative evaluation of a safety function. A safety function indicator criterion is a quantitative limit such that if the criterion is fulfilled, the corresponding safety function is upheld. The safety functions and their associated indicators and criteria developed for the KBS-3 repository are primarily related to the isolating potential and to physical states of the canister and the clay buffer surrounding the canister. They are thus not directly related to release rates of radionuclides. The paper also describes how the concepts introduced i) aid in focussing the assessment on critical, safety related issues, ii) provide a framework for the accounting of safety throughout the different time frames of the assessment and iii) provide key information in the selection of scenarios for the safety assessment. (author)

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

International Nuclear Information System (INIS)

Belal, Al Momani; Jo, Jong Chull

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Safety Review related to Commercial Grade Digital Equipment in Safety System

International Nuclear Information System (INIS)

Yu, Yeongjin; Park, Hyunshin; Yu, Yeongjin; Lee, Jaeheung

2013-01-01

The upgrades or replacement of I and C systems on safety system typically involve digital equipment developed in accordance with non-nuclear standards. However, the use of commercial grade digital equipment could include the vulnerability for software common-mode failure, electromagnetic interference and unanticipated problems. Although guidelines and standards for dedication methods of commercial grade digital equipment are provided, there are some difficulties to apply the methods to commercial grade digital equipment for safety system. This paper focuses on regulatory guidelines and relevant documents for commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. This paper focuses on KINS regulatory guides and relevant documents for dedication of commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. Dedication including critical characteristics is required to use the commercial grade digital equipment on safety system in accordance with KEPIC ENB 6370 and EPRI TR-106439. The dedication process should be controlled in a configuration management process. Appropriate methods, criteria and evaluation result should be provided to verify acceptability of the commercial digital equipment used for safety function

Statistical summary 1990-91

International Nuclear Information System (INIS)

1991-01-01

The information contained in this statistical summary leaflet summarizes in bar charts or pie charts Nuclear Electric's performance in 1990-91 in the areas of finance, plant and plant operations, safety, commercial operations and manpower. It is intended that the information will provide a basis for comparison in future years. The leaflet also includes a summary of Nuclear Electric's environmental policy statement. (UK)

Manual on maintenance of systems and components important to safety

International Nuclear Information System (INIS)

1986-01-01

The Manual should serve as guidance at the plant management level for the maintenance of systems and components important to safety. It includes a detailed description of management systems, administrative controls and procedures. The Annexes contain examples of documents and practices adopted by Operating Organizations of some Member States. It is not the intention of this Manual to address the technical problem of how to maintain a particular component but rather to cover the programmatic aspects of maintenance. It also contains some aspects of surveillance and verification activities. The Manual makes only general statements about radiation protection provisions in connection with maintenance; detailed guidance can be found in other IAEA documents

Final summary report of the Nordic Nuclear Safety Research Program 1998-2001; Sammanfattning av det nordiska forsknings-programmet for karnsakerhet

Energy Technology Data Exchange (ETDEWEB)

Bennerstedt, T. (ed.)

2002-11-01

The results of the 1998 - 2001 NKS program are presented in the form of executive summaries, highlighting the conclusions, recommendations and other findings and results of the six projects carried out during that period. The titles of the six projects are: Risk assessment and strategies for safety (NKS/SOS-1); Reactor safety (NKS/SOS-2); Radioactive waste (NKS/SOS-3); Nuclear Emergency preparedness (NKS/BOK-1); Radiological and environmental consequences (NKS/BOK-2); Nuclear threats from Nordic surroundings (NKS/SBA-1) (ln)

C-reactive protein as an available biomarker determining mental component of health-related quality of life among individuals with spinal cord injury.

Science.gov (United States)

Sabour, Hadis; Latifi, Sahar; Soltani, Zahra; Shakeri, Hania; Norouzi Javidan, Abbas; Ghodsi, Seyed-Mohammad; Hadian, Mohammad Reza; Emami Razavi, Seyed-Hassan

2017-05-01

C-reactive protein (CRP) has been shown to correlate with health-related quality of life (HRQL) in some chronic medical conditions. However, these associations have not yet described in spinal cord injury (SCI). In this study, we tried to identify biomarkers associated with HRQL in SCI. Cross-sectional. Tertiary rehabilitation center. Referred patients to Brain and Spinal Cord Injury Research Center between November 2010 and April 2013. Blood samples were taken to measure circulatory CRP, leptin, adiponectin, ferritin, parathyroid hormone, calcitonin, thyroid hormones, fasting plasma glucose and lipid profile. All the analyses were performed with adjustment for injury-related confounders (level of injury, injury completeness and time since injury) and demographic characteristics. HRQL was measured with Short Form health survey (SF-36). The initial inverse association between CRP and total score of SF-36 (P: 0.006, r = -0.28) was lost after adjustment for confounders. However, the negative correlation between CRP and Mental Component Summary (MCS) remained significant (P: 0.0005, r = -0.38). Leptin level was inversely correlated with Physical Component Summary (PCS) (P: 0.02, r = -0.30). Although CRP and leptin levels were not related with total scores of SF-36 questionnaire, CRP can be more useful in determining mental component of HRQL whereas leptin can be a determinant of physical component. The combined consideration of these two biomarkers may help to predict HRQL in individuals with SCI.

The qualification of electrical components and instrumentations relevant to safety; La qualificazione dei componenti elettrici e di strumentazione rilevanti per la sicurezza

Energy Technology Data Exchange (ETDEWEB)

Zambardi, F [ENEA - Direzione Sicurezza Nucleare e Protezione Sanitaria, Divisione Sistemi Elettrici e Strumentazione, Rome (Italy)

1989-03-15

Systems and components relevant to safety of nuclear power plants must maintain their functional integrity in order to assure accident prevention and mitigation. Redundancy is utilized against random failures, nevertheless care must be taken to avoid common failures in redundant components. Main sources of degradation and common cause failures consist in the aging effects and in the changes of environmental conditions which occur during the plant life and the postulated accidents. These causes of degradation are expected to be especially significant for instrumentation and electrical equipment, which can have a primary role in safety systems. The qualification is the methodology by which component safety requirements can be met against the above mentioned causes of degradation. In this report the connection between the possible, plant conditions and the resulting degradation effects on components is preliminarily addressed. A general characterization of the qualification is then presented. Basis, methods and peculiar aspects are discussed and the qualification by testing is taken into special account. Technical and organizational aspects related to a plant qualification program are also focused. The report ends with a look to the most significant research and development activities. (author)

Safety guidebook relative to the disposal of radioactive wastes in deep geologic formation

International Nuclear Information System (INIS)

2008-01-01

The French nuclear safety authority (ASN) initiated in 2003 a revision process of the objectives to be considered during the research and work steps of the implementation of a radioactive waste storage facility in deep geologic formations. The purpose of this document is to define the safety objectives that have to be retained at each step of this implementation, from the site characterization to the closure of the facility. This update takes into account the works carried out by the ANDRA (French national agency of radioactive wastes) in the framework of the law from December 30, 1991, and the advices of the permanent experts group about these works. It takes also into consideration the international research works in this domain and the choices defined in the program law no 2006-739 from June 28, 2006 relative to the sustainable management of radioactive materials and wastes. The main modifications concern: the notion of reversibility, the definition of the safety functions of disposal components, the safety goals and the design principles assigned to waste packages, the control of nuclear materials and the monitoring objectives of the facility. The documents treats of the following points: 1 - the objectives of public health and environment protection; 2 - the safety principles and the safety-related design bases of the facility; and 3 - the method used for demonstrating the disposal safety. (J.S.)

Continuing the service of safety-related concrete structures in nuclear power plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Mori, Y.; Arndt, E.G.

1993-01-01

The Structural Aging (SAG) Program is addressing the aging management of safety-related concrete structures in nuclear power plants (NPPs) for the purpose of providing improved technical bases for their continued service. The program consists of three technical tasks: materials property data base, structural component assessment/repair technologies, and quantitative methodologies for continued service determinations. Recent accomplishments under each of these tasks are summarized

Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Staunton, R.H.; Cox, D.F. [Oak Ridge National Lab., TN (United States)

1995-03-01

Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail.

Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

International Nuclear Information System (INIS)

Staunton, R.H.; Cox, D.F.

1995-03-01

Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail

Rational quantitative safety goals: a summary

International Nuclear Information System (INIS)

Unwin, S.D.; Hayns, M.R.

1984-08-01

We introduce the notion of a Rational Quantitative Safety Goal. Such a goal reflects the imprecision and vagueness inherent in any reasonable notion of adequate safety and permits such vagueness to be incorporated into the formal regulatory decision-making process. A quantitative goal of the form, the parameter x, characterizing the safety level of the nuclear plant, shall not exceed the value x 0 , for example, is of a non-rational nature in that it invokes a strict binary logic in which the parameter space underlying x is cut sharply into two portions: that containing those values of x that comply with the goal and that containing those that do not. Here, we utilize an alternative form of logic which, in accordance with any intuitively reasonable notion of safety, permits a smooth transition of a safety determining parameter between the adequately safe and inadequately safe domains. Fuzzy set theory provides a suitable mathematical basis for the formulation of rational quantitative safety goals. The decision-making process proposed here is compatible with current risk assessment techniques and produces results in a transparent and useful format. Our methodology is illustrated with reference to the NUS Corporation risk assessment of the Limerick Generating Station

Manufacturing and testing experience for FFTF major safety related components

International Nuclear Information System (INIS)

Peckinpaugh, C.L.

1976-01-01

Experience with FFTF Heat Transport System components during design, manufacturing, and prototype testing is dscussed. Specifically the special design features and the results of the testing performed to assure that the designs provide for safe operation are outlined. Particular emphasis is placed on the full size prototype testing programs and the valuable experience gained

Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

International Nuclear Information System (INIS)

Sarkulov, M.K.

2015-01-01

There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

Preliminary study on health-related lipid components of bakery products.

Science.gov (United States)

Cercaci, Luisito; Conchillo, Ana; Rodriguez-Estrada, Maria Teresa; Ansorena, Diana; Astiasarán, Iciar; Lercker, Giovanni

2006-06-01

The purpose of this study was to evaluate the presence of health-related lipid components, in particular trans fatty acids and sterol oxidation products, in four bakery products. Both types of components are known for their adverse biological effects, especially the increase of atherogenic risk, and therefore it is advisable to monitor their presence in food products. Trans fatty acids were determined by silver-ion thin-layer chromatography-gas chromatography, whereas sterol oxidation was assessed by gas chromatography and gas chromatography-mass spectrometry determination of 7-keto derivatives (tracers of sterol oxidation). The amount of trans fatty acids (0.02 to 3.13 g/100 g of product), sterols (34.9 to 128.3 mg/100 g of product), and 7-keto derivatives of sterols (1.88 to 3.14 mg/kg of product) varied considerably among samples. The supply of phytosterols (22.5 to 64.0 mg/100 g of product) was not significant, and the extent of oxidation of most phytosterols to its corresponding 7-keto derivative was low (0.29 to 0.84%), except for that of brassicasterol (2.01 to 3.11%). The quality of ingredients and raw materials seems to have greatly influenced the fatty acid profile, stability, safety, and quality of the final product; these ingredients should be chosen with extreme care to decrease their potential negative health effects and to increase safety of these products.

Case study on the use of PSA methods: Determining safety importance of systems and components at nuclear power plants

International Nuclear Information System (INIS)

1991-04-01

This case study emphasizes the step of probabilistic safety assessment (PSA) regarding identification of systems and components important to nuclear plant safety. An importance analysis involves combining information that is both qualitative and probabilistic in nature to generate a numerical ranking to determine the system and/or component failures that dominate the risk. Such a ranking can suggest where hardware, software, human factors and component design changes can be implemented to improve plant safety. Examples of using ranking methodology are described. A qualitative ranking criteria is discussed for components and systems that are not included in a PSA. 18 refs, 7 figs, 18 tabs

Chemical Safety Vulnerability Working Group report. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1994-09-01

The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains the Executive summary; Introduction; Summary of vulnerabilities; Management systems weaknesses; Commendable practices; Summary of management response plan; Conclusions; and a Glossary of chemical terms.

Chemical Safety Vulnerability Working Group report. Volume 1

International Nuclear Information System (INIS)

1994-09-01

The Chemical Safety Vulnerability (CSV) Working Group was established to identify adverse conditions involving hazardous chemicals at DOE facilities that might result in fires or explosions, release of hazardous chemicals to the environment, or exposure of workers or the public to chemicals. A CSV Review was conducted in 148 facilities at 29 sites. Eight generic vulnerabilities were documented related to: abandoned chemicals and chemical residuals; past chemical spills and ground releases; characterization of legacy chemicals and wastes; disposition of legacy chemicals; storage facilities and conditions; condition of facilities and support systems; unanalyzed and unaddressed hazards; and inventory control and tracking. Weaknesses in five programmatic areas were also identified related to: management commitment and planning; chemical safety management programs; aging facilities that continue to operate; nonoperating facilities awaiting deactivation; and resource allocations. Volume 1 contains the Executive summary; Introduction; Summary of vulnerabilities; Management systems weaknesses; Commendable practices; Summary of management response plan; Conclusions; and a Glossary of chemical terms

Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables

International Nuclear Information System (INIS)

Koponen, B.L.; Hampel, V.E.

1982-01-01

This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41

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

International Nuclear Information System (INIS)

1984-12-01

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

Component reliability data for use in probabilistic safety assessment

International Nuclear Information System (INIS)

1988-10-01

Generic component reliability data is indispensable in any probabilistic safety analysis. It is not realistic to assume that all possible component failures and failure modes modeled in a PSA would be available from the operating experience of a specific plant in a statistically meaningful way. The degree that generic data is used in PSAs varies from case to case. Some studies are totally based on generic data while others use generic data as prior information to be specialized by plant specific data. Most studies, however, finally use a combination where data for certain components come from generic data sources and others from Bayesian updating. The IAEA effort to compile a generic component reliability data base aimed at facilitating the use of data available in the literature and at highlighting pitfalls which deserve special consideration. It was also intended to complement the fault tree and event tree package (PSAPACK) and to facilitate its use. Moreover, it should be noted, that the IAEA has recently initiated a Coordinated Research Program in Reliability Data Collection, Retrieval and Analysis. In this framework the issues identified as most affecting the quality of existing data bases would be addressed. This report presents the results of a compilation made from the specialized literature and includes reliability data for components usually considered in PSA

Highway Safety Program Manual: Volume 8: Alcohol in Relation to Highway Safety.

Science.gov (United States)

National Highway Traffic Safety Administration (DOT), Washington, DC.

Volume 8 of the 19-volume Highway Safety Program Manual (which provides guidance to State and local governments on preferred highway safety practices) concentrates on alcohol in relation to highway safety. The purpose and objectives of the alcohol program are outlined. Federal authority in the area of highway safety and general policies regarding…

A summary of the Nordic-group conference on safety management

International Nuclear Information System (INIS)

Salo, I.; Svenson, O.

2005-04-01

The report summarizes the Nordic-group conference on safety management, which took place in Lund, Sweden on October 28-29, 2004. The theme-group was originally created by researchers who had a common interest in cooperation, sharing their results, and discuss topics focusing on safety management and safety culture in nuclear power production, but also in other technologies involving risks. The research has, so far, basically been related to the areas of MTO, partly from a psychological perspective, but also from other perspectives. Today, the group consists primarily of members from Sweden, Finland and Norway. During the last three years the group has gathered twice a year. (au)

Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

Materials and Components Technology Division research summary, 1992

International Nuclear Information System (INIS)

1992-11-01

The Materials and Components Technology Division (MCT) provides a research and development capability for the design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs related to nuclear energy support the development of the Integral Fast Reactor (IFR): life extension and accident analyses for light water reactors (LWRs); fuels development for research and test reactors; fusion reactor first-wall and blanket technology; and safe shipment of hazardous materials. MCT Conservation and Renewables programs include major efforts in high-temperature superconductivity, tribology, nondestructive evaluation (NDE), and thermal sciences. Fossil Energy Programs in MCT include materials development, NDE technology, and Instrumentation design. The division also has a complementary instrumentation effort in support of Arms Control Technology. Individual abstracts have been prepared for the database

Model summary report for the safety assessment SR-Site

International Nuclear Information System (INIS)

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

2010-12-01

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

Model summary report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

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

2010-12-15

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

Summary of the law relating to atomic energy and radioactive substances as at March 1979

International Nuclear Information System (INIS)

Sim, D.F.; Ritchie, K.J.S.

1979-01-01

This summary is intended to be a 'signpost' to the relevant law in the United Kingdom, but does not cover any aspect in detail. It falls under the following headings: common law; legislation (Atomic Energy Act 1946 and subordinate legislation; Radioactive Substances Act 1948 and subordinate legislation; Radioactive Substances Act 1960; Electricity (Amendment) Act 1961; Nuclear Installations Acts 1965 and 1969 and subordinate legislation; the Secretary of State for Trade and Industry Order 1970; Radiological Protection Act 1970 as amended by the Health and Safety at Work etc. Act 1974; Air Navigation (Restriction of Flying)(Atomic Energy Establishments) Regulations 1976; Nuclear Safeguards and Electricity (Finance) Act 1978; legislation relating to the United Kingdom Atomic Energy Authority); regulations under the Factories Act 1961; regulations relating to educational establishments; regulations and orders relating to food and medicines; regulations, rules etc. affecting the transport of radioactive materials; regulations under the Social Security Act 1975; control of import and export; the Euratom Treaty; important non-statutory codes of practice etc.; international conventions, regulations etc. relating to the peaceful use of atomic energy and radioactive substances, in which the United Kingdom is interested; foreign legislation. (U.K.)

Final safety analysis report (FSAR) for waste receiving and processing (WRAP) facility

International Nuclear Information System (INIS)

Weidert, J.R.

1997-01-01

This safety analysis report provides a summary description of the WRAP Facility, focusing on significant safety-related characteristics of the location and facility design. This report demonstrates that adherence to the safety basis wi11 ensure necessary operational safety considerations have been addressed sufficiently and justifies the adequacy of the safety basis in protecting the health and safety of the public, workers, and the environment

Annual report of the Association EURATOM-Cea 2004 (executive summary)

International Nuclear Information System (INIS)

2004-01-01

Progress in fusion technology is constant over the years and this report once again highlights a number of important steps that have been accomplished in this domain. This document is the executive summary of the full annual report, summarizing activities performed by the EURATOM-Cea association. This report has been organized into 10 issues: 1) physics integration, 2) reactor vessel, 3) plasma facing components, 4) remote handling, 5) magnets structures, 6) tritium breeding blankets, 7) structural material, 8) safety and environment, 9) system study, and 10) ITER site preparation

Annual report of the Association EURATOM-Cea 2005 (executive summary)

International Nuclear Information System (INIS)

2005-01-01

Progress in fusion technology is constant over the years and this report once again highlights a number of important steps that have been accomplished in this domain. This document is the executive summary of the full annual report, summarizing activities performed by the EURATOM-Cea association. This report has been organized into 10 issues: 1) physics integration, 2) reactor vessel, 3) plasma facing components, 4) remote handling, 5) magnets structures, 6) tritium breeding blankets, 7) structural material, 8) safety and environment, 9) system study, and 10) ITER site preparation

Storage, handling and movement of fuel and related components at nuclear power plants

International Nuclear Information System (INIS)

1979-01-01

The report describes in general terms the various operations involved in the handling of fresh fuel, irradiated fuel, and core components such as control rods, neutron sources, burnable poisons and removable instruments. It outlines the principal safety problems in these operations and provides the broad safety criteria which must be observed in the design, operation and maintenance of equipment and facilities for handling, transferring, and storing nuclear fuel and core components at nuclear power reactor sites

Design, construction, qualification and reliability of main components, from the safety aspect

International Nuclear Information System (INIS)

Crette, J.P.

1982-01-01

In FRANCE, the design and construction of reliable components, which condition the safe operation and availability of breeder plants, is based on the experience acquired during the operation of RAPSODIE, PHENIX and the various test facilities. The technical progress achieved on all main components is illustrated by examples taken from the CREYS-MALVILLE plant. In parallel with the development of these components, an extensive program covering research, development and the definition of design, construction and inspection rules, together with scheduling and quality assurance methods, prepares the industrialization of this reactor system, in compliance with the rules and recommendations issued by the pertinent safety authorities

Summary of the Semiscale Program, 1965-1986

International Nuclear Information System (INIS)

Loomis, G.G.

1987-07-01

This document summarizes significant results from the Semiscale Program, which examined pressurized water reactor (PWR) safety issues from 1965 to 1986. Most of these issues were related to plant response during loss-of-coolant accidents and operational transients. The Semiscale program utilized a series of non-nuclear, scaled, PWR plant simulators to provide thermal-hydraulic data at prototypical pressures and temperatures for a wide range of nuclear safety issues. Presented are: a historical perspective of the Semiscale Program relative to reactor safety with a catalog of the Semiscale experimental facilities and data bases, the relationship of Semiscale results to 10 CFR Part 50 (Appendix K), the impact of Semiscale results on scaling experimental results to full sized operating plants, a summary of safety issues that were addressed by Semiscale testing as they arose throughout the operational lifetime of Semiscale, the contributions of the Semiscale Program to safety technology, a description of phenomena observed during Semiscale testing, the impact of Semiscale data on code development and assessment, the applicability of the Semiscale data base for code development and assessment, and finally, major conclusions and accomplishments of the Semiscale program

Analysis on relation between safety input and accidents

Institute of Scientific and Technical Information of China (English)

YAO Qing-guo; ZHANG Xue-mu; LI Chun-hui

2007-01-01

The number of safety input directly determines the level of safety, and there exists dialectical and unified relations between safety input and accidents. Based on the field investigation and reliable data, this paper deeply studied the dialectical relationship between safety input and accidents, and acquired the conclusions. The security situation of the coal enterprises was related to the security input rate, being effected little by the security input scale, and build the relationship model between safety input and accidents on this basis, that is the accident model.

Report to NASA Committee on Aircraft Operating Problems Relative to Aviation Safety Engineering and Research Activities

Science.gov (United States)

1963-01-01

The following report highlights some of the work accomplished by the Aviation Safety Engineering and Research Division of the Flight Safety Foundations since the last report to the NASA Committee on Aircraft Operating Problems on 22 May 1963. The information presented is in summary form. Additional details may be provided upon request of the reports themselves may be obtained from AvSER.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

The selection of field component reliability data for use in nuclear safety studies

International Nuclear Information System (INIS)

Coxson, B.A.; Tabaie, Mansour

1990-01-01

The paper reviews the user requirements for field component failure data in nuclear safety studies, and the capability of various data sources to satisfy these requirements. Aspects such as estimating the population of items exposed to failure, incompleteness, and under-reporting problems are discussed. The paper takes as an example the selection of component reliability data for use in the Pre-Operational Safety Report (POSR) for Sizewell 'B' Power Station, where field data has in many cases been derived from equipment other than that to be procured and operated on site. The paper concludes that the main quality sought in the available data sources for such studies is the ability to examine failure narratives in component reliability data systems for equipment performing comparable duties to the intended plant application. The main benefit brought about in the last decade is the interactive access to data systems which are adequately structured with regard to the equipment covered, and also provide a text-searching capability of quality-controlled event narratives. (author)

Better Science Through Safety.

Science.gov (United States)

Gerlovich, Jack A.; Downs, Gary E.

Following a brief description of the major components found effective in school safety programs (safety management, education, and services) and data on school accidents in Iowa, this book addresses various aspects of safety related to science instruction, emphasizing that responsibility for safety must be shared by both teacher and students.…

Optimizing Safety Stock Levels in Modular Production Systems Using Component Commonality and Group Technology Philosophy: A Study Based on Simulation

Directory of Open Access Journals (Sweden)

Kenneth Edgar Hernandez-Ruiz

2016-01-01

Full Text Available Modular production and component commonality are two widely used strategies in the manufacturing industry to meet customers growing needs for customized products. Using these strategies, companies can enhance their performance to achieve optimal safety stock levels. Despite the importance of safety stocks in business competition, little attention has been paid to the way to reduce them without affecting the customer service levels. This paper develops a mathematical model to reduce safety stock levels in organizations that employ modular production. To construct the model, we take advantage of the benefits of aggregate inventories, standardization of components, component commonality, and Group Technology philosophy in regard to stock levels. The model is tested through the simulation of three years of operation of two modular product systems. For each system, we calculated and compared the safety stock levels for two cases: (1 under the only presence of component commonality and (2 under the presence of both component commonality and Group Technology philosophy. The results show a reduction in safety stock levels when we linked the component commonality with the Group Technology philosophy. The paper presents a discussion of the implications of each case, features of the model, and suggestions for future research.

Summary of Tiger Team Assessment and Technical Safety Appraisal recurring concerns in the Maintenance Area

International Nuclear Information System (INIS)

1993-01-01

Tiger Team Assessments and Technical Safety Appraisals (TSA) were reviewed and evaluated for concerns in the Maintenance Area (MA). Two hundred and thirty one (231) maintenance concerns were identified by the Tiger Team Assessments and TSA reports. These recurring concerns appear below. A summary of the Noteworthy Practices that were identified and a compilation of the maintenance concerns for each performance objective that were not considered as recurring are also included. Where the Tiger Team Assessment and TSA identified the operating contractor or facility by name, the concern has been modified to remove the name while retaining the intent of the comment

Applications and limits of application of fracture mechanics methods in assessing the safety of components

International Nuclear Information System (INIS)

Stahlberg, R.

1977-01-01

On the basis of fracture mechanics calculations and experimental investigations, it is shown how cracks of different shape and location behave under given static and cyclic loads. In particular, component safety with regard to spontaneous failure and crack growth behaviour in different components are discussed. [de

Summary of the law relating to atomic energy and radioactive substances

International Nuclear Information System (INIS)

Sim, D.F.; Ritchie, K.J.S.

1983-04-01

Intended as a signpost to the relevant law no aspect is covered in detail but a full summary is given. For further details reference has to be made to Acts or regulations themselves. The summary covers the Common Law, the laws in force, regulations under the Factories Act 1961, regulations and orders relating to food and medicines, those concerned with the transport of radioactive materials, regulations under the Social Security Act 1975, Control of Import and Export, the Euratom treaty, important non-statutory codes of practice etc., international conventions, regulations etc. relating to the peaceful use of atomic energy and radioactive substances in which the UK is interested and finally, foreign legislation. The details have been revised as at 31 March 1983. (U.K.)

Research activities of MPA, Stuttgart University, for enhanced safety and reliability of components under complex load

International Nuclear Information System (INIS)

Herter, K.H.; Roos, E.; Schuler, X.; Maile, K.

2004-01-01

MPA research activities focus on fracture prevention and on the development of a generally applicable method of component integrity testing which, independent of the safety relevance of the components involved, is also part of ageing management. (orig.) [de

Research on patient safety: falls and medications.

Science.gov (United States)

Boddice, Sandra Dawn; Kogan, Polina

2009-10-01

Below you will find summaries of published research describing investigations into patient safety issues related to falls and medications. The first summary provides details on the incidence of falls associated with the use of walkers and canes. This is followed by a summary of a fall-prevention intervention study that evaluated the effectiveness of widespread dissemination of evidence-based strategies in a community in Connecticut. The third write up provides information on three classes of medications that are associated with a significant number of emergency room visits. The last summary describes a pharmacist-managed medication reconciliation intervention pilot program. For additional details about the study findings and interventions, we encourage readers to review the original articles.

The complementary roles of fracture mechanics and non-destructive examination in the safety assessment of components

International Nuclear Information System (INIS)

1988-01-01

This document presents the various speeches of the workshop of the Committee on Safety of Nuclear Installations (CSNI) that took place in Wuerenligen, Switzerland, in October 1988. The speeches deal with the roles of Non-Destructive Examination (NDE) and Fracture Mechanics (FM) in the safety assessment of reactor components, such as pressure vessels. Some calibration standards and reference values of defects are presented, and several NDE and FM methods for the assessment of components are described. Separate abstracts were prepared for all the papers in this volume. (TEC)

The complementary roles of fracture mechanics and non-destructive examination in the safety assessment of components

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-31

This document presents the various speeches of the workshop of the Committee on Safety of Nuclear Installations (CSNI) that took place in Wuerenligen, Switzerland, in October 1988. The speeches deal with the roles of Non-Destructive Examination (NDE) and Fracture Mechanics (FM) in the safety assessment of reactor components, such as pressure vessels. Some calibration standards and reference values of defects are presented, and several NDE and FM methods for the assessment of components are described. Separate abstracts were prepared for all the papers in this volume. (TEC).

Components in models of learning: Different operationalisations and relations between components

Directory of Open Access Journals (Sweden)

Mirkov Snežana

2013-01-01

Full Text Available This paper provides the presentation of different operationalisations of components in different models of learning. Special emphasis is on the empirical verifications of relations between components. Starting from the research of congruence between learning motives and strategies, underlying the general model of school learning that comprises different approaches to learning, we have analyzed the empirical verifications of factor structure of instruments containing the scales of motives and learning strategies corresponding to these motives. Considering the problems in the conceptualization of the achievement approach to learning, we have discussed the ways of operational sing the goal orientations and exploring their role in using learning strategies, especially within the model of the regulation of constructive learning processes. This model has served as the basis for researching learning styles that are the combination of a large number of components. Complex relations between the components point to the need for further investigation of the constructs involved in various models. We have discussed the findings and implications of the studies of relations between the components involved in different models, especially between learning motives/goals and learning strategies. We have analyzed the role of regulation in the learning process, whose elaboration, as indicated by empirical findings, can contribute to a more precise operationalisation of certain learning components. [Projekat Ministarstva nauke Republike Srbije, br. 47008: Unapređivanje kvaliteta i dostupnosti obrazovanja u procesima modernizacije Srbije i br. 179034: Od podsticanja inicijative, saradnje i stvaralaštva u obrazovanju do novih uloga i identiteta u društvu

Treatment of Passive Component Reliability in Risk-Informed Safety Margin Characterization FY 2010 Report

Energy Technology Data Exchange (ETDEWEB)

Robert W Youngblood

2010-09-01

The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A technical challenge at the core of this effort is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, is founded on probabilistic characterizations of SSC performance.

Qualification of FPGA-Based Safety-Related PRM System

International Nuclear Information System (INIS)

Miyazaki, Tadashi; Oda, Naotaka; Goto, Yasushi; Hayashi, Toshifumi

2011-01-01

Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system. Considering application to safety-related systems, nonvolatile and non-rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. FPGA is a device which consists only of basic logic circuits, and FPGA performs defined processing which is configured by connecting the basic logic circuit inside the FPGA. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing unit (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. The system which Toshiba developed this time is Power Range Neutron Monitor (PRM). Toshiba is planning to expand application of FPGA-based technology by adopting this development process to the other safety-related systems such as RPS from now on. Toshiba developed a special design process for NRW-FPGA-based safety-related I and C systems. The design process resolves issues for many years regarding testability of the digital system for nuclear safety application. Thus, Toshiba NRW-FPGA-based safety-related I and C systems has much advantage to be a would standard of the digital systems for nuclear safety application. (author)

Significance of BETA and GAMMA dose on environmental qualification of components

International Nuclear Information System (INIS)

Aydogdu, K.M.; Tsang, K.T.

1999-01-01

Safety-related systems and components that are required to perform safety functions during accident conditions must be designed to withstand the harsh environmental conditions that occur as a consequence of the accident. Where these conditions are 'harsh', and equipment operability can potentially be affected by the post-accident environment environmental qualification of the equipment must be conducted to demonstrate that the required safety function can be maintained. It is also understood that non-safety related equipment that affects, or prevents, the satisfactory operation of a safety-related system should also withstand the 'harsh' environmental conditions caused by an appropriate design-basis accident. There are essentially two types of requirements that must be satisfied to qualify equipment or components to withstand radiation damage, namely economic requirements and safety requirements. The general objective of the economic requirement is to reduce maintenance cost and to maximize component life during reactor operation. The general objective of the safety requirement is that the equipment should be qualified to withstand the harsh post-accident environmental conditions and should function properly for the appropriate length of time after a design-basis accident has occurred. To address the economic factors - i.e., to reduce maintenance costs and to maximize component life - the radiation dose rates to equipment are calculated throughout the reactor building and the service building during reactor operation. These are also used for the safety requirement purpose, to assess radiation ageing of safety-related components caused by degradation of material properties with time at radiation exposure. To address the safety requirement, the dose-rate estimates and accumulated doses after a LOCA coincident with loss-of-emergency-core cooling (LOECC) are provided. The harsh post-accident environmental conditions defined for environmental qualification of components

Expansion of passive safety function

International Nuclear Information System (INIS)

Inai, Nobuhiko; Nei, Hiromichi; Kumada, Toshiaki.

1995-01-01

Expansion of the use of passive safety functions is proposed. Two notions are presented. One is that, in the design of passive safety nuclear reactors where aversion of active components is stressed, some active components are purposely introduced, by which a system is built in such a way that it behaves in an apparently passive manner. The second notion is that, instead of using a passive safety function alone, a passive safety function is combined with some active components, relating the passivity in the safety function with enhanced controllability in normal operation. The nondormant system which the authors propose is one example of the first notion. This is a system in which a standby safety system is a portion of the normal operation system. An interpretation of the nondormant system via synergetics is made. As an example of the second notion, a PIUS density lock aided with active components is proposed and is discussed

On Optimum Safety Levels of Breakwaters

DEFF Research Database (Denmark)

Burcharth, Hans F.; Sørensen, John Dalsgaard

2006-01-01

The paper presents results from numerical simulations performed with the objective of identifying optimum design safety levels of conventional rubble mound and caisson breakwaters, corresponding to the lowest costs over the service life of the structures. The work is related to the PIANC Working...... Group 47 on "Selection of type of breakwater structures". The paper summaries results given in Burcharth and Sorensen (2005) related to outer rubble mound breakwaters but focus on optimum safety levels for outer caisson breakwaters on low and high rubble foundations placed on sea beds strong enough...... to resist geotechnical slip failures. Optimum safety levels formulated for use both in deterministic and probabilistic design procedures are given. Results obtained so far indicate that the optimum safety levels for caisson breakwaters are much higher than for rubble mound breakwaters....

SAFIR. The Finnish research programme on nuclear power plant safety 2003-2006. Executive summary

International Nuclear Information System (INIS)

Puska, E.

2006-12-01

Major part of Finnish public research on nuclear power plant safety during the years 2003-2006 has been carried out in the SAFIR programme. The programme has been administrated by the steering group that was nominated by the Ministry of Trade and Industry (KTM). The steering group of SAFIR has consisted of representatives from Radiation and Nuclear Safety Authority (STUK), Ministry of Trade and Industry (KTM), Technical Research Centre of Finland (VTT), Teollisuuden Voima Oy (TVO), Fortum Power and Heat Oy, Fortum Nuclear Services Oy (Fortum), Finnish Funding Agency for Technology and Innovation (Tekes), Helsinki University of Technology (TKK) and Lappeenranta University of Technology (LTY). The key research areas of SAFIR have been (1) reactor fuel and core, (2) reactor circuit and structural safety, (3) containment and process safety functions, that was divided in 2005 into (3a) thermal hydraulics and (3b) severe accidents, (4) automation, control room and IT, (5) organisations and safety management and (6) risk-informed safety management. The research programme has included annually from 20 up to 24 research projects, whose volume has varied from a few person months to several person years. The total volume of the programme during the four year period 2003-2006 has been 19.7 million euros and 148 person years. The research in the programme has been carried out primarily by Technical Research Centre of Finland (VTT). Other research units responsible for the projects include Lappeenranta University of Technology, Fortum Nuclear Services Oy, Helsinki University of Technology and RAMSE Consulting Oy. In addition, there have been a few minor subcontractors in some projects. The programme management structure has consisted of the steering group, a reference group in each of the seven research areas and a number of ad hoc groups in the various research areas. This report gives a short summary of the results of the SAFIR programme for the period January 2003 - November

A summary description of the flammable gas tank safety program

International Nuclear Information System (INIS)

Johnson, G.D.; Sherwood, D.J.

1994-10-01

Radioactive liquid waste may produce hydrogen as result of the interaction of gamma radiation and water. If the waste contains organic chelating agents, additional hydrogen as well as nitrous oxide and ammonia may be produced by thermal and radiolytic decomposition of these organics. Several high-level radioactive liquid waste storage tanks, located underground at the Hanford Site in Washington State, are on a Flammable Gas Watch List. Some contain waste that produces and retains gases until large quantities of gas are released rapidly to the tank vapor space. Tanks nearly-filled to capacity have relatively little vapor space; therefore if the waste suddenly releases a large amount of hydrogen and nitrous oxide, a flammable gas mixture could result. The most notable example of a Hanford waste tank with a flammable gas problem is tank 241-SY-101. Upon occasion waste stored in this tank has released enough flammable gas to burn if an ignition source had been present inside of the tank. Several, other Hanford waste tanks exhibit similar behavior although to a lesser magnitude. Because this behavior was hot adequately-addressed in safety analysis reports for the Hanford Tank Farms, an unreviewed safety question was declared, and in 1990 the Flammable Gas Tank Safety Program was established to address this problem. The purposes of the program are a follows: (1) Provide safety documents to fill gaps in the safety analysis reports, and (2) Resolve the safety issue by acquiring knowledge about gas retention and release from radioactive liquid waste and developing mitigation technology. This document provides the general logic and work activities required to resolve the unreviewed safety question and the safety issue of flammable gas mixtures in radioactive liquid waste storage tanks

Managing Safety in Small and Medium Enterprises

DEFF Research Database (Denmark)

Stephen, legg; Olsen, Kirsten Bendix; Ian S., laird

2015-01-01

on safety in SMEs, showing how most current policy and legislation on occupational health and safety (OSH) and the work environment is based on large enterprises and that there is a relative paucity of research on OSH in SMEs. In a summary of current knowledge, it is argued that modern OHS legislation......This paper presents a conceptual model for increasing acceptable working environments for SMEs. It also acts as an editorial for the special issue of Safety Science on ‘Managing safety in small and medium enterprises (SMEs)’. It describes how seven of the ten papers in the special issue originate...

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

XUV synchrotron optical components for the Advanced Light Source: Summary of the requirements and the developmental program

International Nuclear Information System (INIS)

McKinney, W.; Irick, S.; Lunt, D.

1992-07-01

We give a brief summary of the requirements for water cooled optical components for the Advanced Light Source (ALS), a third generation synchrotron radiation source under construction at Lawrence Berkeley Laboratory (LBL). Materials choices, surface figure and smoothness specifications, and metrology systems for measuring the plated metal surfaces are discussed. Results from a finished water cooled copper alloy mirror will be used to demonstrate the state of the art in optical metrology with the Takacs Long Trace Profiler (LTP II)

The Agency's Technical Co-operation programme with Uganda, 1982-1992. Country programme summaries

International Nuclear Information System (INIS)

1992-01-01

The country programme summary reported here is one in the series of such studies being undertaken of the Agency's TC programme with Member States. With $1.6 million of Agency support received, Uganda ranks 61st among all recipients of technical assistance in the period 1958 through 1991. Almost half of the assistance received during the past ten years has been provided in the form of equipment (47%), complemented by a strong training component (37%), while expert services have received only a relatively small share (16%). Ninety-eight per cent of the resources were provided by the Technical Assistance and Co-operation Fund, the remaining two per cent were made available through assistance in kind. During the past ten years, project activities - and disbursements - have concentrated exclusively on three major areas: agriculture (64%), nuclear medicine (24%) and nuclear safety (12%)

75 FR 15485 - Pipeline Safety: Workshop on Guidelines for Integrity Assessment of Cased Pipe

Science.gov (United States)

2010-03-29

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket ID...: Pipeline and Hazardous Materials Safety Administration (PHMSA), DOT. ACTION: Notice of workshop. SUMMARY... ``Guidelines for Integrity Assessment of Cased Pipe in Gas Transmission Pipelines'' and related Frequently...

The reactor safety study of experimental multi-purpose VHTR design

International Nuclear Information System (INIS)

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

1981-01-01

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

Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

International Nuclear Information System (INIS)

McConnell, J.W., Jr; Ayers, A.L. Jr; Tyacke, M.J.

1996-02-01

This report provides a graded approach for classification of components used in transportation packaging and dry spent fuel storage systems. This approach provides a method for identifying, the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements The general types of transportation packagings and dry storage systems were identified. Discussions were held with suppliers and fabricators of packagings and storage systems to determine current practices. The methodology used in this report is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This report also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems. The safety importance of each component is discussed, and a classification category is assigned

Transactions of the nineteenth water reactor safety information meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1991-10-01

This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately

Current safety issues related to research reactor operation

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

2000-01-01

The Agency has included activities on research reactor safety in its Programme and Budget (P and B) since its inception in 1957. Since then, these activities have traditionally been oriented to fulfil the Agency's functions and obligations. At the end of the decade of the eighties, the Agency's Research Reactor Safety Programme (RRSP) consisted of a limited number of tasks related to the preparation of safety related publications and the conduct of safety missions to research reactor facilities. It was at the beginning of the nineties when the RRSP was upgraded and expanded as a subprogramme of the Agency's P and B. This subprogramme continued including activities related to the above subjects and started addressing an increasing number of issues related to the current situation of research reactors (in operation and shut down) around the world such as reactor ageing, modifications and decommissioning. The present paper discusses some of the above issues as recognised by various external review or advisory groups (e.g., Peer Review Groups under the Agency's Performance Programme Appraisal System (PPAS) or the standing International Nuclear Safety Advisory Group (INSAG)) and the impact of their recommendations on the preparation and implementation of the part of the Agency's P and B relating to the above subject. (author)

Reliability Analysis on NPP's Safety-Related Control Module with Field Data

International Nuclear Information System (INIS)

Lee, Sang Yong; Jung, Jae Hyun; Kim, Seong Hun

2006-01-01

The automatic control systems used in nuclear power plant (NPP) consists of numerous control modules that can be considered to be a network of components various complex ways. The control modules require relatively high reliability than industrial electronic products. Reliability prediction provides the rational basis of system designs and also provides the safety significance of system operations. The aim of this paper is to minimize the deficiencies of the traditional reliability prediction method calculation using the available field return data. This way is possible to do more realistic reliability assessment. SAMCHANG Enterprise Company (SEC) has established database containing high quality data at the module and component level from module maintenance in NPP. On the basis of these, this paper compares results that add failure record (field data) to Telcordia-SR-332 reliability prediction model with MIL-HDBK-217F prediction results

INC93: international nuclear congress. Technical sessions summaries

International Nuclear Information System (INIS)

Boulton, J.

1993-01-01

The conference dealt mainly with safety, environmental and human aspects of nuclear energy, including waste management. This volume consists of summaries of the technical sessions contained in Volumes 2 and 3. The individual papers have also been abstracted separately for INIS, with the summaries abridged

INC93: international nuclear congress. Technical sessions summaries

Energy Technology Data Exchange (ETDEWEB)

Boulton, J

1994-12-31

The conference dealt mainly with safety, environmental and human aspects of nuclear energy, including waste management. This volume consists of summaries of the technical sessions contained in Volumes 2 and 3. The individual papers have also been abstracted separately for INIS, with the summaries abridged.

76 FR 33773 - Navigation Safety Advisory Council; Vacancies

Science.gov (United States)

2011-06-09

... Council; Vacancies AGENCY: Coast Guard, DHS. ACTION: Request for applications. SUMMARY: The Coast Guard seeks applications for membership on the Navigation Safety Advisory Council (NAVSAC). NAVSAC provides... the U.S. Coast Guard, on matters relating to prevention of maritime collisions, rammings, and...

Transactions of the Twentieth Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Weiss, A.J.

1992-10-01

This report contains summaries of papers on reactor safety research to be presented at the 20th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 21--23, 1992. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included

Aging and service wear of air-operated valves used in safety-related systems at nuclear power plants

International Nuclear Information System (INIS)

Cox, D.F.; McElhaney, K.L.; Staunton, R.H.

1995-05-01

Air-operated valves (AOVs) are used in a variety of safety-related applications at nuclear power plants. They are often used where rapid stroke times are required or precise control of the valve obturator is required. They can be designed to operate automatically upon loss of power, which is often desirable when selecting components for response to design basis conditions. The purpose of this report is to examine the reported failures of AOVs and determine whether there are identifiable trends in the failures related to predictable causes. This report examines the specific components that comprise a typical AOV, how those components fail, when they fail, and how such failures are discovered. It also examines whether current testing frequencies and methods are effective in predicting such failures

Safety Basis Report

International Nuclear Information System (INIS)

R.J. Garrett

2002-01-01

As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities

Safety Basis Report

Energy Technology Data Exchange (ETDEWEB)

R.J. Garrett

2002-01-14

As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities.

Review of current summary of irradiated foods

International Nuclear Information System (INIS)

Dyakova, A.; Tsvetkova, E.; Miteva, D.; Dimov, K.

2005-01-01

The reasons about widely application of food irradiation as a technology for increasing food safety were presenting in this summary. Topics which are discussing are about: purpose of the irradiation, radiological safety, toxicology, microbiological conclusion, alimentary adequacy, packing, labeling, public acceptance, inspection of food irradiated plants

Modelling the effects of road traffic safety measures.

Science.gov (United States)

Lu, Meng

2006-05-01

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

A Case of Pediatric Abdominal Wall Reconstruction: Components Separation within the Austere War Environment

Directory of Open Access Journals (Sweden)

Ian Valerio, MD, MS, MBA, FACS

2014-07-01

Full Text Available Summary: Reconstructive surgeons supporting military operations are required to definitively treat severe pediatric abdominal injuries in austere environments. The safety and efficacy of using a components separation technique to treat large ventral hernias in pediatric patients in this setting remains understudied. Components separation technique was required to achieve definitive closure in a 12-month-old pediatric patient in Kandahar, Afghanistan. Her course was complicated by an anastomotic leak after small bowel resection. Her abdominal was successfully reopened, the leak repaired, and closed primarily without incident on postinjury day 9. Abdominal trauma with a large ventral hernia requiring components separation is extremely rare. A pediatric patient treated with components separation demonstrated minimal complications, avoidance of abdominal compartment syndrome, and no mortality.

IEEE Std 382-1985: IEEE standard for qualification of actuators for power operated valve assemblies with safety-related functions for nuclear power plants

International Nuclear Information System (INIS)

Anon.

1992-01-01

This standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. This standard may also be used to separately qualify actuator components. This standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of power-driven valve actuators with safety-related functions Part I describes the qualification process. Part II describes the standard qualification cases and their environmental parameters for the usual locations of safety-related equipment in a nuclear generating station. Part III describes the qualification tests outlined in 6.3.3

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Summary report

International Nuclear Information System (INIS)

Smith, Paul; Neall, Fiona; Snellman, Margit; Pastina, Barbara; Nordman, Henrik; Johnson, Lawrence; Hjerpe, Thomas

2008-03-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. This safety assessment is summarised in the present report. The scientific basis of the safety assessment includes around 30 years of scientific RandD and technical development in the Swedish and Finnish KBS-3V programmes. Much of this scientific basis is directly applicable to KBS-3H. This has allowed the KBS-3H safety studies to focus on those issues that are unique to this design alternative, identified in a systematic 'difference analysis' of KBS-3H and KBS-3V. This difference analysis has shown that the key differences in the evolution and performance of KBS-3H and KBS-3V relate mainly to the engineered barrier system and to the impact of local variations in the rate of groundwater inflow on buffer saturation along the KBS-3H deposition drifts. No features or processes specific to KBS-3H have been identified that could lead to a loss or substantial degradation of the safety functions of the engineered barriers over a million year time frame. Radionuclide release from the repository near field in the event of

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Summary report

Energy Technology Data Exchange (ETDEWEB)

Smith, Paul; Neall, Fiona; Snellman, Margit; Pastina, Barbara; Nordman, Henrik; Johnson, Lawrence; Hjerpe, Thomas

2008-03-15

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. This safety assessment is summarised in the present report. The scientific basis of the safety assessment includes around 30 years of scientific RandD and technical development in the Swedish and Finnish KBS-3V programmes. Much of this scientific basis is directly applicable to KBS-3H. This has allowed the KBS-3H safety studies to focus on those issues that are unique to this design alternative, identified in a systematic 'difference analysis' of KBS-3H and KBS-3V. This difference analysis has shown that the key differences in the evolution and performance of KBS-3H and KBS-3V relate mainly to the engineered barrier system and to the impact of local variations in the rate of groundwater inflow on buffer saturation along the KBS-3H deposition drifts. No features or processes specific to KBS-3H have been identified that could lead to a loss or substantial degradation of the safety functions of the engineered barriers over a million year time frame. Radionuclide release from the repository near field in the

Reactor safety activities at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN)-Nuclebras

International Nuclear Information System (INIS)

Pinheiro, R.B.

1986-08-01

This report presents the major activities developed at NUCLEBRAS in the area of Reactor Safety. The Thermohydraulics and Reactor Safety program consists of activities in the analytical and experimental area. The more recent activities under way in the analytical area are presented together with some results. They are mainly related to the development and assessment of advanced computer codes to be used in the analysis of the behavior of the plant during operational transients and postulated accidents and in thermohydraulic core design. In the experimental area, the main accomplishments and activities under way are reported they refer to steady state CHF tests in a 9-rod bundle and rewetting experiments in a tubular test section. In particular, the status of the design development of a new facility for Separate Effects LOCA tests (DTL-ES) is described. Concerning the Components Testing program, the basic characteristics of a valve test facility (CTC) are described and the present status of its construction is reported. Reference is also made to existing installations and a summary of tests already performed for Angra-2 components qualification is presented. (Author) [pt

78 FR 40146 - Patient Safety Organizations: Voluntary Relinquishment From Northern Metropolitan Patient Safety...

Science.gov (United States)

2013-07-03

... DEPARTMENT OF HEALTH AND HUMAN SERVICES Agency for Healthcare Research and Quality Patient Safety Organizations: Voluntary Relinquishment From Northern Metropolitan Patient Safety Institute AGENCY: Agency for Healthcare Research and Quality (AHRQ), HHS. ACTION: Notice of Delisting. SUMMARY: The Patient Safety and...

76 FR 7853 - Patient Safety Organizations: Voluntary Delisting From Oregon Patient Safety Commission

Science.gov (United States)

2011-02-11

... DEPARTMENT OF HEALTH AND HUMAN SERVICES Agency for Healthcare Research and Quality Patient Safety Organizations: Voluntary Delisting From Oregon Patient Safety Commission AGENCY: Agency for Healthcare Research and Quality (AHRQ), HHS. ACTION: Notice of delisting. SUMMARY: Oregon Patient Safety Commission: AHRQ...

Aging management and PLEX in Swiss nuclear power plants and prioritization of safety class 2 and 3 components

International Nuclear Information System (INIS)

Fuchs, R.; Stejskal, J.

2000-01-01

In this presentation ageing management of systems and components important to safety of the Swiss nuclear power plants are presented. Status of electrical components, status of mechanical components as well as status of civil structures are reviewed. The scheme of the high pressure core spray system is included

Summary of existing superconducting magnet experience and its relevance to the safety of fusion magnet

International Nuclear Information System (INIS)

Hsieh, S.Y.; Allinger, J.; Danby, G.; Keane, J.; Powell, J.; Prodell, A.

1975-01-01

A comprehensive summary of experience with over twenty superconducting magnet systems has been collected through visits to and discussions about existing facilities including, for example, the bubble chamber magnets at Brookhaven National Laboratory, Argonne National Laboratory and Fermi National Accelerator Laboratory, and the large superconducting spectrometer at Stanford Linear Accelerator Center. This summary includes data relating to parameters of these magnets, magnet protection methods, and operating experiences. The information received is organized and presented in the context of its relevance to the safe operation of future, very large superconducting magnet systems for fusion power plants

K Basin Fuel Characterization Program Technical Baseline Summary

International Nuclear Information System (INIS)

SUYAMA, R.M.

1999-01-01

This document provides a summary of the systematic process used by the SNF Project to characterize K-Basin spent fuel, and to develop and apply the appropriate conservative safety margins to the resulting parameters for technical designs and safety analyses

Summary cortisol reactivity indicators: Interrelations and meaning

Directory of Open Access Journals (Sweden)

Jennifer E. Khoury

2015-01-01

Full Text Available Research on the hypothalamic pituitary adrenal (HPA axis has involved a proliferation of cortisol indices. We surveyed recently published HPA-related articles and identified 15 such indices. We sought to clarify their biometric properties, specifically, how they interrelate and what they mean, because such information is rarely offered in the articles themselves. In the present article, the primary samples consist of community mothers and their infants (N = 297, who participated in two challenges, the Toy Frustration Paradigm and the Strange Situation Procedure. We sought to cross-validate findings from each of these samples against the other, and also against a clinically depressed sample (N = 48 and a sample of healthy older adults (N = 51 who participated in the Trier Social Stress Test. Cortisol was collected from all participants once before and twice after the challenges. These heterogenous samples were chosen to obtain the greatest possible range in cortisol levels and stress response regulation. Using these data, we computed the 15 summary cortisol indices identified in our literature survey. We assessed inter-relations amongst indices and determined their underlying dimensions via principal component analysis (PCA. The PCAs consistently extracted two components, accounting for 79%–93% of the variance. These components represent “total cortisol production” and “change in cortisol levels.” The components were highly congruent across challenge, time, and sample. High variable loadings and explained factor variance suggest that all indices represent their underlying dimensions very well. Thus the abundance of summary cortisol indices currently represented in the literature appears superfluous.

Safety assessment of Olkiluoto NPP units 1 and 2. Decision of the Radiation and Nuclear Safety Authority regarding the periodic safety review of the Olkiluoto NPP

International Nuclear Information System (INIS)

2010-02-01

In this safety assessment the Radiation and Nuclear Safety Authority (STUK) has evaluated the safety of the Olkiluoto Nuclear Power Plant units 1 and 2 in connection with the periodic safety review. This safety assessment provides a summary of the reviews, inspections and continuous oversight carried out by STUK. The issues addressed in the assessment and the related evaluation criteria are set forth in the nuclear energy and radiation safety legislation and the regulations issued thereunder. The provisions of the Nuclear Energy Act concerning the safe use of nuclear energy, security and emergency preparedness arrangements, and waste management are specified in more detail in the Government Decrees and Regulatory Guides issued by STUK. Based on the assessment, STUK consideres that the Olkiluoto Nuclear Power Plant units 1 and 2 meet the set safety requirements for operational nuclear power plants, the emergency preparedness arrangements are sufficient and the necessary control to prevent the proliferation of nuclear weapons has been appropriately arranged. The physical protection of the Olkiluoto nuclear power plant is not yet completely in compliance with the requirements of Government Decree 734/2008, which came into force in December 2008. Further requirements concerning this issue based also on the principle of continuous improvement were included in the decision relating to the periodic safety review. The safety of the Olkiluoto nuclear power plant was assessed in compliance with the Government Decree on the Safety of Nuclear Power Plants (733/2008), which came into force in 2008. The decree notes that existing nuclear power plants need not meet all the requirements set out for new plants. Most of the design bases pertaining to the Olkiluoto 1 and 2 nuclear power plant units were set in the 1970s. Substantial modernisations have been carried out at the Olkiluoto 1 and 2 nuclear power plant units since their commissioning to improve safety. This is in line with

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

International Nuclear Information System (INIS)

Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

2015-01-01

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy's (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

Energy Technology Data Exchange (ETDEWEB)

Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

2015-01-31

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

Assessment and management of ageing of major nuclear power plant components important to safety: Steam generators

International Nuclear Information System (INIS)

1997-11-01

This TECDOC is one in a series of reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety which was issued by the IAEA in 1992. They have been compiled using contributions from technical experts in typically 10 to 12 countries for each report, a feedback from a September 1994 Technical Committee Meeting attended by 53 technical experts from 21 Member States (who reviewed first drafts in specialized working groups), and review comments from invited specialists

Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume I

International Nuclear Information System (INIS)

2000-12-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume II

International Nuclear Information System (INIS)

2000-12-01

At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

76 FR 58812 - Patient Safety Organizations: Delisting for Cause of Patient Safety Organization One, Inc.

Science.gov (United States)

2011-09-22

... Organizations: Delisting for Cause of Patient Safety Organization One, Inc. AGENCY: Agency for Healthcare Research and Quality (AHRQ), HHS. ACTION: Notice of Delisting. SUMMARY: Patient Safety Organization One, Inc.: AHRQ has delisted Patient Safety Organization One, Inc. as a Patient Safety Organization (PSO...

Development Trends in Nuclear Technology and Related Safety Aspects

International Nuclear Information System (INIS)

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

2002-01-01

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO sub-task 'User Requirements and Nuclear Energy Development Criteria in the Area of Safety' have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R and D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle. (authors)

Transactions of the twenty-fifth water reactor safety information meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

Transactions of the twenty-fifth water reactor safety information meeting

International Nuclear Information System (INIS)

Monteleone, S.

1997-09-01

This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session

Literature review of environmental qualification of safety-related electric cables: Summary of past work. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Subudhi, M. [Brookhaven National Lab., Upton, NY (United States)

1996-04-01

This report summarizes the findings from a review of published documents dealing with research on the environmental qualification of safety-related electric cables used in nuclear power plants. Simulations of accelerated aging and accident conditions are important considerations in qualifying the cables. Significant research in these two areas has been performed in the US and abroad. The results from studies in France, Germany, and Japan are described in this report. In recent years, the development of methods to monitor the condition of cables has received special attention. Tests involving chemical and physical examination of cable`s insulation and jacket materials, and electrical measurements of the insulation properties of cables are discussed. Although there have been significant advances in many areas, there is no single method which can provide the necessary information about the condition of a cable currently in service. However, it is possible that further research may identify a combination of several methods that can adequately characterize the cable`s condition.

Literature review of environmental qualification of safety-related electric cables: Summary of past work. Volume 1

International Nuclear Information System (INIS)

Subudhi, M.

1996-04-01

This report summarizes the findings from a review of published documents dealing with research on the environmental qualification of safety-related electric cables used in nuclear power plants. Simulations of accelerated aging and accident conditions are important considerations in qualifying the cables. Significant research in these two areas has been performed in the US and abroad. The results from studies in France, Germany, and Japan are described in this report. In recent years, the development of methods to monitor the condition of cables has received special attention. Tests involving chemical and physical examination of cable's insulation and jacket materials, and electrical measurements of the insulation properties of cables are discussed. Although there have been significant advances in many areas, there is no single method which can provide the necessary information about the condition of a cable currently in service. However, it is possible that further research may identify a combination of several methods that can adequately characterize the cable's condition

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Science.gov (United States)

2010-01-01

..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... functions. Risk-Informed Safety Class (RISC)-2 structures, systems and components (SSCs) means nonsafety-related SSCs that perform safety significant functions. Risk-Informed Safety Class (RISC)-3 structures...

Periodic safety review of the experimental fast reactor JOYO. Review of aging management

International Nuclear Information System (INIS)

Isozaki, Kazunori; Ogawa, To-ru; Nishino, Kazunari

2005-05-01

Periodic safety review (Review of the aging management) which consisted of ''Technical review on aging for the safety related structures, systems and components'' and ''Establishment a long term maintenance program'' was carried out up to April 2005. 1. Technical review on aging for the safety related structures, systems and components. It was technically confirmed to prevent the loss of function of the safety related structures, systems and components due to aging phenomena, which (1) irradiation damage, (2) corrosion, (3) abrasion and erosion, (4) thermal aging, (5) creep and fatigue, (6) Stress Corrosion Cracking, (7) insulation deterioration and (8) general deterioration, under the periodic monitoring or renewal of them. 2. Establishment of long term maintenance program. The long term maintenance during JFY2005 to 2014 were established based on the technical review on aging for the safety related structures, systems and components. It was evaluated that the inspection and renewal based on the long term maintenance program, in addition to the spontaneous inspection of the long term voluntary long-term inspection plan, could prevent the loss of function of the safety related structures, systems and components. (author)

Progress summary of the Chernobyl accident

International Nuclear Information System (INIS)

Iddekinge, F.W. van

1986-01-01

Based on two IAEA documents (the report of the USSR State Committee on the Utilization of Atomic Energy named 'The accident at the Chernobyl nuclear power plant and its consequences' prepared for the IAEA Experts Meeting held in Vienna on 25-29 August, 1986 and the INSAG (International Nuclear Safety Advisory Group) summary report on the Post-accident review meeting on the Chernobyl accident, drawn up in Vienna from August 30 until September 5, 1986, this publication tries to present a logic relation between the special features of the RMBK-1000 LWGR, the cause of the accident, and the technical countermeasures. (Auth.)

Age-related differences in working memory updating components.

Science.gov (United States)

Linares, Rocío; Bajo, M Teresa; Pelegrina, Santiago

2016-07-01

The aim of this study was to investigate possible age-related changes throughout childhood and adolescence in different component processes of working memory updating (WMU): retrieval, transformation, and substitution. A set of numerical WMU tasks was administered to four age groups (8-, 11-, 14-, and 21-year-olds). To isolate the effect of each of the WMU components, participants performed different versions of a task that included different combinations of the WMU components. The results showed an expected overall decrease in response times and an increase in accuracy performance with age. Most important, specific age-related changes in the retrieval component were found, demonstrating that the effect of retrieval on accuracy was larger in children than in adolescents or young adults. These findings indicate that the availability of representations from outside the focus of attention may change with age. Thus, the retrieval component of updating could contribute to the age-related changes observed in the performance of many updating tasks. Copyright © 2016 Elsevier Inc. All rights reserved.

Challenges in the management of gas voids in safety related systems

International Nuclear Information System (INIS)

Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M.

2009-01-01

Gas intrusion into Safety Related Systems, such as the Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) and Containment Spray (CS) in nuclear power plants is undesirable and can lead to pump binding (depending on the void fraction and flow rate) and damaging water hammer events. Gas ingestion in pumps can result in total or momentary loss of hydraulic performance resulting in possible pump shaft seizure rendering the pumps unable to perform their safety functions or reduce the pump discharge pressure and flow capacity to the point that the system cannot perform its design function. Extreme cases of gas water hammer can result in physical damage to system piping, components and supports, and possible relief valve lifting events with consequential loss of inventory. NRC Generic Letter GL 2008 01, 'Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,' requires US utilities to demonstrate that suitable design, operational and testing measures are in place to maintain licensing commitments. The Generic Letter (GL 2008 01) outlines a number of actions that are detailed in nature, such as establishing pump void tolerance limits; establishing limits on pump suction void fractions, assuring adequate system venting capability, identification of all possible sources of gas intrusion, preventing vortex formation in tanks, and determining acceptable limits of gas in system discharge piping.. Regarding one of these issues, GL 2008 01 indicates that the amount of gas that can be ingested without significant impact on pump design, gas dispersion and flow rate. Each US nuclear power plant licensee is required to evaluate their ECCS, DHR and CS system design, operation and test procedures to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of 10 CFR 50 Appendix B. Typically, gas pockets get into the safety related systems through a number

Challenges in the management of gas voids in safety related systems

Energy Technology Data Exchange (ETDEWEB)

Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M. [Westinghouse Electric Company LLC, Pittsburgh (United States)

2009-04-15

Gas intrusion into Safety Related Systems, such as the Emergency Core Cooling System (ECCS), Decay Heat Removal (DHR) and Containment Spray (CS) in nuclear power plants is undesirable and can lead to pump binding (depending on the void fraction and flow rate) and damaging water hammer events. Gas ingestion in pumps can result in total or momentary loss of hydraulic performance resulting in possible pump shaft seizure rendering the pumps unable to perform their safety functions or reduce the pump discharge pressure and flow capacity to the point that the system cannot perform its design function. Extreme cases of gas water hammer can result in physical damage to system piping, components and supports, and possible relief valve lifting events with consequential loss of inventory. NRC Generic Letter GL 2008 01, 'Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems,' requires US utilities to demonstrate that suitable design, operational and testing measures are in place to maintain licensing commitments. The Generic Letter (GL 2008 01) outlines a number of actions that are detailed in nature, such as establishing pump void tolerance limits; establishing limits on pump suction void fractions, assuring adequate system venting capability, identification of all possible sources of gas intrusion, preventing vortex formation in tanks, and determining acceptable limits of gas in system discharge piping.. Regarding one of these issues, GL 2008 01 indicates that the amount of gas that can be ingested without significant impact on pump design, gas dispersion and flow rate. Each US nuclear power plant licensee is required to evaluate their ECCS, DHR and CS system design, operation and test procedures to assure that gas intrusion is minimized and monitored in order to maintain system operability and compliance with the requirements of 10 CFR 50 Appendix B. Typically, gas pockets get into the safety related systems through

Aviation safety and maintenance under major organizational changes, investigating non-existing accidents.

Science.gov (United States)

Herrera, Ivonne A; Nordskag, Arve O; Myhre, Grete; Halvorsen, Kåre

2009-11-01

The objective of this paper is to discuss the following questions: Do concurrent organizational changes have a direct impact on aviation maintenance and safety, if so, how can this be measured? These questions were part of the investigation carried out by the Accident Investigation Board, Norway (AIBN). The AIBN investigated whether Norwegian aviation safety had been affected due to major organizational changes between 2000 and 2004. The main concern was the reduction in safety margins and its consequences. This paper presents a summary of the techniques used and explains how they were applied in three airlines and by two offshore helicopter operators. The paper also discusses the development of safety related indicators in the aviation industry. In addition, there is a summary of the lessons learned and safety recommendations. The Norwegian Ministry of Transport has required all players in the aviation industry to follow up the findings and recommendations of the AIBN study.

Transactions of the Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session

Transactions of the Twenty-First Water Reactor Safety Information Meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

Transactions of the twenty-second water reactor safety information meeting

International Nuclear Information System (INIS)

1994-10-01

This report contains summaries of papers on reactor safety research to be presented at the 22nd Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 24--26, 1994. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session. Individual papers have been cataloged separately

Association Between Awareness of Hypertension and Health-Related Quality of Life in a Cross-Sectional Population-Based Study in Rural Area of Northwest China.

Science.gov (United States)

Mi, Baibing; Dang, Shaonong; Li, Qiang; Zhao, Yaling; Yang, Ruihai; Wang, Duolao; Yan, Hong

2015-07-01

Hypertensive patients have more complex health care needs and are more likely to have poorer health-related quality of life than normotensive people. The awareness of hypertension could be related to reduce health-related quality of life. We propose the use of quantile regression to explore more detailed relationships between awareness of hypertension and health-related quality of life. In a cross-sectional, population-based study, 2737 participants (including 1035 hypertensive patients and 1702 normotensive participants) completed the Short-Form Health Survey. A quantile regression model was employed to investigate the association of physical component summary scores and mental component summary scores with awareness of hypertension and to evaluate the associated factors. Patients who were aware of hypertension (N = 554) had lower scores than patients who were unaware of hypertension (N = 481). The median (IQR) of physical component summary scores: 48.20 (13.88) versus 53.27 (10.79), P mental component summary scores: 50.68 (15.09) versus 51.70 (10.65), P = 0.03. adjusting for covariates, the quantile regression results suggest awareness of hypertension was associated with most physical component summary scores quantiles (P mental component summary scores quantiles in which the β-estimates from -3.47 (95% CI: -6.65 to -0.39) to -2.18 (95% CI: -4.30 to -0.06). The awareness of hypertension has a greater effect on those with intermediate physical component summary status: the β-estimates were equal to -2.04 (95% CI: -3.51 to -0.57, P Awareness of hypertension was negatively related to health-related quality of life in hypertensive patients in rural western China, which has a greater effect on mental component summary scores with the poorer status and on physical component summary scores with the intermediate status.

Development of Non-safety System Architecture and Evaluation of Components/Systems

International Nuclear Information System (INIS)

Oh, I. S.; Lee, C. K.; Kim, D. H.; Lee, J. W.; Lee, D. Y.; Park, W. M.; Hwang, I. K.; Hur, S.; Kim, J. T.; Park, J. C.; Lee, J. W.

2007-10-01

We describe in this report the works performed for a technical evaluation of the non-safety digital control system of the KNICS, the non-safety process control system of the KNICS, a communication load analysis for the MMIS (including both the non-safety and the safety systems) of the KNICS, the development of MMI and an implementation of the logic for the CVCS, and the works performed to support writing a proposal needed for bidding an I and C system based on the KNICS. The technical evaluation results were aimed to be used by the designers to detect parts needed to be corrected or to be newly inserted, and also by the developers during the development phase. The requirement specifications and the data requirement characteristics have been identified for each subsystem of the determined KNICS structure. For each communication node, the specifications related to the data transfer including the data capacity for interfaces, delay time for the data transfer, and the marginal availability of its performance capabilities have been analyzed to identify the amount of data transfer and hence to verify that both of the designed structures for the safety related communications network and for the digital communications network are appropriate. The results of the supporting work performed for writing the technical specifications related to each subsystem of the KNICS structure, are expected to be useful in writing a proposal for the expected Uljin new units 1 and 2, and in the I and C upgrade for any of the existing nuclear power plants under operation. Also included in this report are the descriptions on a design of the chemical volume control system (CVCS), on the supporting work performed to draw the logic diagrams for CVCS using the tool ISaGRAF, and on the generation of a set of system displays to be used as references

Development of Non-safety System Architecture and Evaluation of Components/Systems

Energy Technology Data Exchange (ETDEWEB)

Oh, I. S.; Lee, C. K.; Kim, D. H.; Lee, J. W.; Lee, D. Y.; Park, W. M.; Hwang, I. K.; Hur, S.; Kim, J. T.; Park, J. C.; Lee, J. W

2007-10-15

We describe in this report the works performed for a technical evaluation of the non-safety digital control system of the KNICS, the non-safety process control system of the KNICS, a communication load analysis for the MMIS (including both the non-safety and the safety systems) of the KNICS, the development of MMI and an implementation of the logic for the CVCS, and the works performed to support writing a proposal needed for bidding an I and C system based on the KNICS. The technical evaluation results were aimed to be used by the designers to detect parts needed to be corrected or to be newly inserted, and also by the developers during the development phase. The requirement specifications and the data requirement characteristics have been identified for each subsystem of the determined KNICS structure. For each communication node, the specifications related to the data transfer including the data capacity for interfaces, delay time for the data transfer, and the marginal availability of its performance capabilities have been analyzed to identify the amount of data transfer and hence to verify that both of the designed structures for the safety related communications network and for the digital communications network are appropriate. The results of the supporting work performed for writing the technical specifications related to each subsystem of the KNICS structure, are expected to be useful in writing a proposal for the expected Uljin new units 1 and 2, and in the I and C upgrade for any of the existing nuclear power plants under operation. Also included in this report are the descriptions on a design of the chemical volume control system (CVCS), on the supporting work performed to draw the logic diagrams for CVCS using the tool ISaGRAF, and on the generation of a set of system displays to be used as references.

Time dependent unavailability analysis of nuclear safety systems considering periodically tested components

International Nuclear Information System (INIS)

Goes, Alexandre Gromann de Araujo

1988-01-01

It is of utmost importance to have a computer code in order to analyze how different parameters (like test duration time) affect the unavailability of safety systems of nuclear. In this context, a study was performed in order to evaluate the model employed by the FRANTIC computer code, which performs detailed calculations on the contribution to the system unavailability originated by hardware failures, component tests and repairs, aiming at considering the influence of different test schemes on the system unavailability. It was shown, by means of the results attained that the numerical model used by the FRANTIC code and the analytical model proposed by APOSTOLAKIS and CHU (4) give unavailability values much similar when the component tests are supposed to be perfect. When a test is supposed to be imperfect (that is, when it may induce a test is supposed to be imperfect (that is, when it may induce a failure on the component being tested), the analytical model presents more conservative results. (author)

Summary report on the use of plant safety performance indicators

International Nuclear Information System (INIS)

2001-09-01

In 1998, the OECD/NEA committee on Nuclear Regulatory Activities (CNRA) initiated an activity with the objective of advancing the discussion on how to enhance and measure regulatory effectiveness in relation to nuclear installations. One of the outcome of this activity was to establish a Task group to develop internal (direct) performance indicators which would be used to monitor regulatory efficiency. In parallel, a joint CNRA/CSNI group was launched in December 2000 to exchange information and develop external (indirect) indicators to measure regulatory effectiveness, i.e. impact on licensee's safety performance. These external indicators are, in other words, the traditional plant performance indicators (PI's) and these are the ones that this report deals with. This report presents the work performed by the joint CNRA/CSNI task group mentioned above. It provides a summary of the sets of PI's being used by different regulatory bodies and WANO, it describes the national practices on the use of PI's and proposes a set of PI's that could be used nationally describing regulatory effectiveness and also as a basis for an international system. The task force consisted of regulators, organisations which have a performance indicators system in operation or under testing. The task force met in Paris on February 19-20, 2001. Each participant provided a brief description of the PI System at his organisation and its usage. The group identified a list of PI's that are recommended to be used nationally by regulators. This paper has been elaborated based on the information exchanged and discussions held in the February meeting. The participating countries (Spain, Finland, US, Sweden) and WANO were asked to provide an overview of systems in use. The presently used Performance Indicators were reviewed in a three steps process. 1. First indicators used in at least two agencies were identified. 2. The second step was to identify the most used indicators. 3. The third step was to

Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

International Nuclear Information System (INIS)

1996-03-01

On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff's basis for issuing GL 94-03, as well as the staff's assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date

Status of the safety concept and safety demonstration for an HLW repository in salt. Summary report

Energy Technology Data Exchange (ETDEWEB)

Bollingerfehr, W.; Buhmann, D.; Filbert, W.; and others

2013-12-15

Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safety assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

Human reliability analysis methods for probabilistic safety assessment

International Nuclear Information System (INIS)

Pyy, P.

2000-11-01

Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

Using the Job Demands-Resources model to investigate risk perception, safety climate and job satisfaction in safety critical organizations.

Science.gov (United States)

Nielsen, Morten Birkeland; Mearns, Kathryn; Matthiesen, Stig Berge; Eid, Jarle

2011-10-01

Using the Job Demands-Resources model (JD-R) as a theoretical framework, this study investigated the relationship between risk perception as a job demand and psychological safety climate as a job resource with regard to job satisfaction in safety critical organizations. In line with the JD-R model, it was hypothesized that high levels of risk perception is related to low job satisfaction and that a positive perception of safety climate is related to high job satisfaction. In addition, it was hypothesized that safety climate moderates the relationship between risk perception and job satisfaction. Using a sample of Norwegian offshore workers (N = 986), all three hypotheses were supported. In summary, workers who perceived high levels of risk reported lower levels of job satisfaction, whereas this effect diminished when workers perceived their safety climate as positive. Follow-up analyses revealed that this interaction was dependent on the type of risks in question. The results of this study supports the JD-R model, and provides further evidence for relationships between safety-related concepts and work-related outcomes indicating that organizations should not only develop and implement sound safety procedures to reduce the effects of risks and hazards on workers, but can also enhance other areas of organizational life through a focus on safety. © 2011 The Authors. Scandinavian Journal of Psychology © 2011 The Scandinavian Psychological Associations.

Summary of Great East Japan Earthquake response at Onagawa Nuclear Power Station and further safety improvement measures

International Nuclear Information System (INIS)

Sato, Toru

2013-01-01

A large earthquake occurred on March 11, 2011 and tsunami was generated following it. The East Japan suffered serious damage by the earthquake and tsunami. This is called the Great East Japan Earthquake. Onagawa Nuclear Power Station (NPS) is located closest to the epicenter of Great East Japan Earthquake. We experienced intense shake by the earthquake and some flooding from the tsunami, however, we have succeeded safely cold shutdown of the reactors. In this paper, we introduce summary of Great East Japan Earthquake response a Onagawa NPS and safety improvement measures which are based on both experience of Onagawa NPS and lesson from Fukushima Daiichi NPS accident. (author)

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

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

International Nuclear Information System (INIS)

Ramanauskiene, A.

2000-01-01

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

The eNanoMapper database for nanomaterial safety information.

Science.gov (United States)

Jeliazkova, Nina; Chomenidis, Charalampos; Doganis, Philip; Fadeel, Bengt; Grafström, Roland; Hardy, Barry; Hastings, Janna; Hegi, Markus; Jeliazkov, Vedrin; Kochev, Nikolay; Kohonen, Pekka; Munteanu, Cristian R; Sarimveis, Haralambos; Smeets, Bart; Sopasakis, Pantelis; Tsiliki, Georgia; Vorgrimmler, David; Willighagen, Egon

2015-01-01

The NanoSafety Cluster, a cluster of projects funded by the European Commision, identified the need for a computational infrastructure for toxicological data management of engineered nanomaterials (ENMs). Ontologies, open standards, and interoperable designs were envisioned to empower a harmonized approach to European research in nanotechnology. This setting provides a number of opportunities and challenges in the representation of nanomaterials data and the integration of ENM information originating from diverse systems. Within this cluster, eNanoMapper works towards supporting the collaborative safety assessment for ENMs by creating a modular and extensible infrastructure for data sharing, data analysis, and building computational toxicology models for ENMs. The eNanoMapper database solution builds on the previous experience of the consortium partners in supporting diverse data through flexible data storage, open source components and web services. We have recently described the design of the eNanoMapper prototype database along with a summary of challenges in the representation of ENM data and an extensive review of existing nano-related data models, databases, and nanomaterials-related entries in chemical and toxicogenomic databases. This paper continues with a focus on the database functionality exposed through its application programming interface (API), and its use in visualisation and modelling. Considering the preferred community practice of using spreadsheet templates, we developed a configurable spreadsheet parser facilitating user friendly data preparation and data upload. We further present a web application able to retrieve the experimental data via the API and analyze it with multiple data preprocessing and machine learning algorithms. We demonstrate how the eNanoMapper database is used to import and publish online ENM and assay data from several data sources, how the "representational state transfer" (REST) API enables building user friendly

Concept of a new method for fatigue monitoring of nuclear power plant components

International Nuclear Information System (INIS)

Zafosnik, M.; Cizelj, L.

2007-01-01

Fatigue is one of the well-understood aging mechanisms affecting mechanical components in many industrial facilities including nuclear power plants. Operational experience of nuclear power plants worldwide to date confirmed adequate design of safety related components against fatigue. In some cases however, for example when the plant life extension is envisioned, it may be very useful to monitor the remaining fatigue life of safety related components. Nuclear power plants components are classified into safety classes regarding their importance in mitigating the consequences of hypothetic accidents. Service life of components subjected to fatigue loading can be estimated with Usage Factor uk. A concept of the new method aiming both at monitoring the current state of the component and predicting its remaining lifetime in the life-extension conditions is presented. The method is based on determination of partial Usage Factor of components in which operating transients will be considered and compared to design transients. (author)

78 FR 17212 - Patient Safety Organizations: Voluntary Relinquishment From Universal Safety Solution PSO

Science.gov (United States)

2013-03-20

... DEPARTMENT OF HEALTH AND HUMAN SERVICES Agency for Healthcare Research and Quality Patient Safety... Research and Quality (AHRQ), HHS. ACTION: Notice of delisting. SUMMARY: The Patient Safety and Quality Improvement Act of 2005 (Patient Safety Act), Public Law 109-41, 42 U.S.C. 299b-21--b-26, provides for the...

A combined deterministic and probabilistic procedure for safety assessment of components with cracks - Handbook.

Energy Technology Data Exchange (ETDEWEB)

Dillstroem, Peter; Bergman, Mats; Brickstad, Bjoern; Weilin Zang; Sattari-Far, Iradj; Andersson, Peder; Sund, Goeran; Dahlberg, Lars; Nilsson, Fred (Inspecta Technology AB, Stockholm (Sweden))

2008-07-01

SSM has supported research work for the further development of a previously developed procedure/handbook (SKI Report 99:49) for assessment of detected cracks and tolerance for defect analysis. During the operative use of the handbook it was identified needs to update the deterministic part of the procedure and to introduce a new probabilistic flaw evaluation procedure. Another identified need was a better description of the theoretical basis to the computer program. The principal aim of the project has been to update the deterministic part of the recently developed procedure and to introduce a new probabilistic flaw evaluation procedure. Other objectives of the project have been to validate the conservatism of the procedure, make the procedure well defined and easy to use and make the handbook that documents the procedure as complete as possible. The procedure/handbook and computer program ProSACC, Probabilistic Safety Assessment of Components with Cracks, has been extensively revised within this project. The major differences compared to the last revision are within the following areas: It is now possible to deal with a combination of deterministic and probabilistic data. It is possible to include J-controlled stable crack growth. The appendices on material data to be used for nuclear applications and on residual stresses are revised. A new deterministic safety evaluation system is included. The conservatism in the method for evaluation of the secondary stresses for ductile materials is reduced. A new geometry, a circular bar with a circumferential surface crack has been introduced. The results of this project will be of use to SSM in safety assessments of components with cracks and in assessments of the interval between the inspections of components in nuclear power plants

The implementation of a burnup credit based criticality safety assessment in the THORP head end plant

International Nuclear Information System (INIS)

Gulliford, J.; Edge, J.A.; Gracey, J.; Harris, N.

2003-01-01

A new criticality safety assessment based on Actinide-Only Burnup Credit has been developed to cover operations in BNFL's Thermal Oxide Reprocessing Plant (THORP). Reduction of the gadolinium concentration leads to significant reduction in active waste volumes. Detailed description of the methodology was presented at ICNC 1999 and the basic components of the approved safety case have remained unchanged from those proposed then. This paper presents a brief summary of the new methodology, and describes further analyses carried out to quantify additional safety margins. These additional margins are not credited in the derivation of the operating limits, but provide further evidence of the fault tolerance inherent in the new regime. As part of the arrangements to monitor the overall performance of the plant and instrumentation under the new regime, various analyses of plant data are made, including 'on-line' cross checks of measured versus expected fuel parameters (i.e. in addition to the checks on Residual Enrichment). Statistical analyses of data are made and compared with similar data from earlier batches. A summary of analyses made on some of the early fuel batches is presented here. A summary of the likely further development in the Burnup Credit methodology is given in this paper. (author)

Assessing propensity to learn from safety-related events

NARCIS (Netherlands)

Drupsteen, L.; Wybo, J.L.

2015-01-01

Most organisations aim to use experience from the past to improve safety, for instance through learning from safety-related incidents and accidents. Whether an organisation is able to learn successfully can however only be determined afterwards. So far, there are no proactive measures to assess

Ventilation systems and components of nuclear power plants

International Nuclear Information System (INIS)

1997-01-01

The most important radiation and nuclear safety requirements for the design and manufacture of nuclear power plant ventilation systems and components are presented in the guide. Also the regulatory activities of the Finnish Centre for Radiation and Nuclear Safety (STUK) as regards the ventilation systems and components are explained. Documents and data which shall be submitted to STUK during the various phases of the regulatory procedure relating to the design, construction, commissioning and operation of the nuclear power plants are presented. (13 refs.)

EDF ageing management program of nuclear components: a safety and economical issue

International Nuclear Information System (INIS)

Faidy, C.

2005-01-01

Ageing management of Nuclear Power Plants is an essential issue for utilities, in term of safety and availability and corresponding economical consequences. Practically all nuclear countries have developed a systematic program to deal with ageing of components on their plants. This paper presents the ageing management program developed by EDF and that are compared with different other approaches in other countries (IAEA guidelines and GALL report). The paper presents a general overview of the programs, the major results, recommendations and conclusions. (author)

Reliability Prediction Of System And Component Of Process System Of RSG-GAS Reactor

International Nuclear Information System (INIS)

Sitorus Pane, Jupiter

2001-01-01

The older the reactor the higher the probability of the system and components suffer from loss of function or degradation. This phenomenon occurred because of wear, corrosion, and fatigue. Study on component reliability was generally performed deterministically and statistically. This paper would describe an analysis of using statistical method, i.e. regression Cox, in order to predict the reliability of the components and their environmental influence's factors. The result showed that the dynamics, non safety related, and mechanic components have higher risk of failure, whereas static, safety related, and electric have lower risk of failures. The relative risk value for variable of components dynamics, quality, dummy 1 and dummy 2 are of 1.54, 1.59, 1.50, and 0.83 compare to other components type with each variable. Component with the higher risk have lower reliability than lower one

Prevent recurrence of nuclear disaster (4). Future tasks in the field of structure and components

International Nuclear Information System (INIS)

Okamoto, Koji; Takagi, Toshiyuki; Ueda, Susumu

2012-01-01

Structure and components subcommittee under the special committee of seismic safety of nuclear power stations of the Atomic Energy Society of Japan discussed future activities related with technical problems of seismic design of structures, components and piping system and evaluation of seismic effects in collaboration with the Japan Society of Mechanical Engineers. These problems were arranged by 'logic of seismic safety' and tabulated just enough, and then their roadmap was prepared. This article described selected relevant problems and discussed safety margins of seismic design and their related problems, referring to state of countermeasures and evaluated results of nuclear power stations after Great East Japan Earthquake occurred in March 11, 2011. Main problems were related with seismic safety margins of structure and components, consideration of ground motion index, rationalization and upgrade of seismic design, application of new technology, integrity evaluation of structure and components after or at earthquake, and upgrade of seismic probabilistic risk assessment methodology. (T. Tanaka)

Executive Summary

International Nuclear Information System (INIS)

2012-01-01

The OECD Nuclear Energy Agency (NEA) Integration Group for the Safety Case (IGSC) organised a workshop to assess current understanding on the use of cementitious materials in radioactive waste disposal. The workshop was hosted by the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (Ondraf/Niras), in Brussels, Belgium on 17-19 November 2009. The workshop brought together a wide range of people involved in supporting safety case development and having an interest in cementitious materials: namely, cement and concrete experts, repository designers, scientists, safety assessors, disposal programme managers and regulators. The workshop was designed primarily to consider issues relevant to the post-closure safety of radioactive waste disposal, but also addressed some related operational issues, such as cementitious barrier emplacement. Where relevant, information on cementitious materials from analogous natural and anthropogenic systems was also considered. The workshop agenda is included as Appendix A. The workshop focused on: - The uses of different cementitious materials in various repository designs. - The evolution of cementitious materials over long time scales (1000s to 100000s of years). - The interaction of cementitious materials with surrounding components of the repository (e.g. waste, container, buffer, backfill, host rock). - The workshop comprised: - Plenary sessions in which the state-of-the-art on repository design and understanding the phenomenology of cementitious materials and their interactions were presented and discussed. - Dedicated working group sessions, which were used to discuss key safety assessment and safety case questions in more detail. For example: How strong is the scientific basis for incorporating the various aspects of the behaviour and interactions of cementitious materials in safety assessments and safety cases? How can the behaviour and interactions of cementitious materials best be incorporated within the

An initial examination of aging related degradation in turbine drives and governors for safety related pumps

International Nuclear Information System (INIS)

Cox, D.F.

1991-01-01

This study is being performed to examine the relationship between time dependent degradation, and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized water reactor (PWR) plants, and the Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) systems for Boiling Water Reactor (BWR) facilities. This research has been conducted by examining current information in NPRDS, reviewing Licensee Event Reports, and thoroughly investigating contacts with operating plant personnel, and by personal observation. The reported information was reviewed to determine the cause of the event and the method of discovery. From this data attempts have been made at determining the predictability of events and possible preventive measures that may be implemented. Findings in a recent study on the Auxiliary Feedwater System (NUREG/CR-5404) indicate that the turbine drive is the single largest contributor to AFW system degradation. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design seem to indicate that this equipment can be a reliable component in safety systems

Summary Of Session 5: How Should We Handle Safety?

Energy Technology Data Exchange (ETDEWEB)

Albert, M.; Roy, G

2001-07-01

This session was originally titled 'Safety: Who cares?' in a fairly provocative way. A clear conclusion of this session and discussions that were held at the workshop is that there is a wide concern for safety among the people in charge of control room operations. This was shown as well by the quality of the seven talks presented in this session on subjects ranging from safety standards to a practical case of a safety incident. (author)

Summary Of Session 5: How Should We Handle Safety?

International Nuclear Information System (INIS)

Albert, M.; Roy, G.

2001-01-01

This session was originally titled 'Safety: Who cares?' in a fairly provocative way. A clear conclusion of this session and discussions that were held at the workshop is that there is a wide concern for safety among the people in charge of control room operations. This was shown as well by the quality of the seven talks presented in this session on subjects ranging from safety standards to a practical case of a safety incident. (author)

Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Bourham, Mohamed A.; Gilligan, John G.

1999-08-14

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

Aging techniques and qualified life for safety system components

International Nuclear Information System (INIS)

Weaver, W.W.

1980-01-01

Presently, the qualified life objective for Class IE safety system components in nuclear power plants is somewhat of a subjective engineering judgment. When the desired qualified life is ascertained, there are other choices that must be made (which may be influenced by the desired qualified life) such as selecting the aging procedure to use in the qualification process. Adding complexity to the situation is the fact that there are some limitations in aging techniques at the present time. This article presents (1) a discussion of the limitations in aging procedures, (2) the general philosophy of qualification, and (3) a proposed method for specifying a desired qualified life, which uses a probabilistic approach. The probabilistic approach proposed in item 3 can be applied to natural aging programs and eventually to accelerated aging once the present technical difficulties are overcome

FY 1996 activity summary

International Nuclear Information System (INIS)

1997-01-01

The US Department of Energy Office of Nuclear and Facility Safety provides nuclear safety policy, independent technical evaluation, and technical support. A summary of these activities is provided in this report. These include: (1) changing the mission of the former production facilities to storage and waste management; (2) stabilizing nuclear materials not recycled due to production cessation or interruptions; (3) reformulating the authorization basis for existing facilities to convert to a standards based approach for operations consistent with modern expectations; and (4) implementing a modern regulatory framework for nuclear facilities. Enforcement of the Price-Anderson Amendments Act is also reported

Safety margins in deterministic safety analysis

International Nuclear Information System (INIS)

Viktorov, A.

2011-01-01

The concept of safety margins has acquired certain prominence in the attempts to demonstrate quantitatively the level of the nuclear power plant safety by means of deterministic analysis, especially when considering impacts from plant ageing and discovery issues. A number of international or industry publications exist that discuss various applications and interpretations of safety margins. The objective of this presentation is to bring together and examine in some detail, from the regulatory point of view, the safety margins that relate to deterministic safety analysis. In this paper, definitions of various safety margins are presented and discussed along with the regulatory expectations for them. Interrelationships of analysis input and output parameters with corresponding limits are explored. It is shown that the overall safety margin is composed of several components each having different origins and potential uses; in particular, margins associated with analysis output parameters are contrasted with margins linked to the analysis input. While these are separate, it is possible to influence output margins through the analysis input, and analysis method. Preserving safety margins is tantamount to maintaining safety. At the same time, efficiency of operation requires optimization of safety margins taking into account various technical and regulatory considerations. For this, basic definitions and rules for safety margins must be first established. (author)

77 FR 6411 - Training, Qualification, and Oversight for Safety-Related Railroad Employees

Science.gov (United States)

2012-02-07

... Oversight for Safety-Related Railroad Employees AGENCY: Federal Railroad Administration (FRA), Department of... establishing minimum training standards for each category and subcategory of safety-related railroad employee... or contractor that employs one or more safety-related railroad employee to develop and submit a...

Second ANS workshop on the safety of Soviet-designed nuclear power plants. Summary report

International Nuclear Information System (INIS)

Bari, R.A.

1995-03-01

The Second American Nuclear Society Workshop on the Safety of Soviet-Designed Nuclear Power Plants was held in Washington, DC, in November, 1994. The Workshop consisted of both plenary sessions and working sessions with three hundred participants overall. All countries with operating Soviet-Designed nuclear power plants were represented and representatives from several other countries also participated. In addition to the status and plans related to technical issues, the Workshop also included discussions of economic, political, legal, and social issues as they relate to the safety of these nuclear power plants

Safety-related operator actions: methodology for developing criteria

International Nuclear Information System (INIS)

Kozinsky, E.J.; Gray, L.H.; Beare, A.N.; Barks, D.B.; Gomer, F.E.

1984-03-01

This report presents a methodology for developing criteria for design evaluation of safety-related actions by nuclear power plant reactor operators, and identifies a supporting data base. It is the eleventh and final NUREG/CR Report on the Safety-Related Operator Actions Program, conducted by Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The operator performance data were developed from training simulator experiments involving operator responses to simulated scenarios of plant disturbances; from field data on events with similar scenarios; and from task analytic data. A conceptual model to integrate the data was developed and a computer simulation of the model was run, using the SAINT modeling language. Proposed is a quantitative predictive model of operator performance, the Operator Personnel Performance Simulation (OPPS) Model, driven by task requirements, information presentation, and system dynamics. The model output, a probability distribution of predicted time to correctly complete safety-related operator actions, provides data for objective evaluation of quantitative design criteria

Safety Test Program Summary SNAP 19 Pioneer Heat Source Safety Program

Energy Technology Data Exchange (ETDEWEB)

None,

1971-07-01

Sixteen heat source assemblies have been tested in support of the SNAP 19 Pioneer Safety Test Program. Seven were subjected to simulated reentry heating in various plasma arc facilities followed by impact on earth or granite. Six assemblies were tested under abort accident conditions of overpressure, shrapnel impact, and solid and liquid propellant fires. Three capsules were hot impacted under Transit capsule impact conditions to verify comparability of test results between the two similar capsule designs, thus utilizing both Pioneer and Transit Safety Test results to support the Safety Analysis Report for Pioneer. The tests have shown the fuel is contained under all nominal accident environments with the exception of minor capsule cracks under severe impact and solid fire environments. No catastrophic capsule failures occurred in this test which would release large quantities of fuel. In no test was fuel visible to the eye following impact or fire. Breached capsules were defined as those which exhibit thoria contamination on its surface following a test, or one which exhibited visible cracks in the post test metallographic analyses.

Nordic studies in reactor safety

International Nuclear Information System (INIS)

Pershagen, N.

1993-01-01

The Nordic Nuclear Safety Research Programme SIK programme in reactor safety is part of a major joint Nordic research effort in nuclear safety. The report summarizes the achievements of the SIK programme, which was carried out during 1990-1993 in collaboration between Nordic nuclear utilities, safety authorities, and research institutes. Three main projects were successfully completed dealing with: 1) development and application of a living PSA concept for monitoring the risk of core damage, and of safety indicators for early warning of possible safety problems; 2) review and intercomparison of severe accident codes, case studies of potential core melt accidents in nordic reactors, development of chemical models for the MAAP code, and outline of a system for computerized accident management support; 3) compilation of information about design and safety features of neighbouring reactors in Germany, Lithuania and Russia, and for naval reactors and nuclear submarines. The report reviews the state-of-the-art in each subject matter as an introduction to the individual project summaries. The main findings of each project are highlighted. The report also contains an overview of reactor safety research in the Nordic countries and a summary of fundamental reactor safety principles. (au) (69 refs.)

Inroads into Equestrian Safety: Rider-Reported Factors Contributing to Horse-Related Accidents and Near Misses on Australian Roads

Science.gov (United States)

Thompson, Kirrilly; Matthews, Chelsea

2015-01-01

Simple Summary Riding horses on roads can be dangerous, but little is known about accidents and near misses. To explore road safety issues amongst Australian equestrians, we conducted an online survey. More than half of all riders (52%) reported having experienced at least one accident or near miss in the 12 months prior to the survey, mostly attributed to speed. Whilst our findings confirmed factors identified overseas, we also identified issues around road rules, hand signals and road rage. This paper suggests strategies for improving the safety of horses, riders and other road users. Abstract Horse riding and horse-related interactions are inherently dangerous. When they occur on public roads, the risk profile of equestrian activities is complicated by interactions with other road users. Research has identified speed, proximity, visibility, conspicuity and mutual misunderstanding as factors contributing to accidents and near misses. However, little is known about their significance or incidence in Australia. To explore road safety issues amongst Australian equestrians, we conducted an online survey. More than half of all riders (52%) reported having experienced at least one accident or near miss in the 12 months prior to the survey. Whilst our findings confirm the factors identified overseas, we also identified issues around rider misunderstanding of road rules and driver misunderstanding of rider hand signals. Of particular concern, we also found reports of potentially dangerous rider-directed road rage. We identify several areas for potential safety intervention including (1) identifying equestrians as vulnerable road users and horses as sentient decision-making vehicles; (2) harmonising laws regarding passing horses; (3) mandating personal protective equipment; (4) improving road signage; (5) comprehensive data collection; (6) developing mutual understanding amongst road-users; (7) safer road design and alternative riding spaces; and (8) increasing investment

The Qualification Experiences for Safety-critical Software of POSAFE-Q

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Son, Kwang Seop; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-05-15

Programmable Logic Controllers (PLC) have been applied to the Reactor Protection System (RPS) and the Engineered Safety Feature (ESF)-Component Control System (CCS) as the major safety system components of nuclear power plants. This paper describes experiences on the qualification of the safety-critical software including the pCOS kernel and system tasks related to a safety-grade PLC, i.e. the works done for the Software Verification and Validation, Software Safety Analysis, Software Quality Assurance, and Software Configuration Management etc.

Summary of the law relating to atomic energy and radioactive substances

International Nuclear Information System (INIS)

Sim, D.F.; Ritchie, K.J.S.

1981-04-01

A summary is given of the law relating to atomic energy and radioactive substances revised as at 31 March 1981 under the following headings: (1) The common law. (2) The legislation. (3) Regulations under the factories act 1961. (4) Regulations relating to educational establishments. (5) Regulations and orders relating to food and medicines. (6) Regulations, rules, etc. affecting the transport of radioactive materials. (7) Regulations under the social security act 1975. (8) Control of import and export. (9) The Euratom treaty. (10) Important nonstatutory codes of practice, etc.. (11) International conventions, regulations, etc. relating to the peaceful use of atomic energy and radioactive substances, in which the United Kingdom is interested. (12) Foreign legislation. (U.K.)

Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

Energy Technology Data Exchange (ETDEWEB)

Lowe, Andrew; Hayward, Brent (Dedale Asia Pacific, Albert Park VIC 3206 (Australia))

2006-08-15

This report documents a study with the objective of enhancing safety culture in the Swedish nuclear power industry. A primary objective of this study was to ensure that the latest thinking on human factors principles was being recognised and applied by nuclear power operators as a means of ensuring optimal safety performance. The initial phase of the project was conducted as a pilot study, involving the senior management group at one Swedish nuclear power-producing site. The pilot study enabled the project methodology to be validated after which it was repeated at other Swedish nuclear power industry sites, providing a broad-ranging analysis of opportunities across the industry to enhance safety culture. The introduction to this report contains an overview of safety culture, explains the background to the project and sets out the project rationale and objectives. The methodology used for understanding and analysing the important safety culture issues at each nuclear power site is then described. This section begins with a summary of the processes used in the information gathering and data analysis stage. The six components of the Management Workshops conducted at each site are then described. These workshops used a series of presentations, interactive events and group exercises to: (a) provide feedback to site managers on the safety culture and safety leadership issues identified at their site, and (b) stimulate further safety thinking and provide 'take-away' information and leadership strategies that could be applied to promote safety culture improvements. Section 3, project Findings, contains the main observations and output from the project. These include: - a brief overview of aspects of the local industry operating context that impinge on safety culture; - a summary of strengths or positive attributes observed within the safety culture of the Swedish nuclear industry; - a set of identified opportunities for further improvement; - the aggregated

NIF special equipment construction health and safety plan

Energy Technology Data Exchange (ETDEWEB)

Sawicki, R.H.

1997-07-28

The purpose of this plan is to identify how the construction and deployment activities of the National Ignition Facility (NIF) Special Equipment (SE) will be safely executed. This plan includes an identification of (1) the safety-related responsibilities of the SE people and their interaction with other organizations involved; (2) safety related requirements, policies, and documentation; (3) a list of the potential hazards unique to SE systems and the mechanisms that will be implemented to control them to acceptable levels; (4) a summary of Environmental Safety and Health (ES&H) training requirements; and (5) requirements of contractor safety plans that will be developed and used by all SE contractors participating in site activities. This plan is a subsidiary document to the NIF Construction Safety Program (CSP) and is intended to compliment the requirements stated therein with additional details specific to the safety needs of the SE construction-related activities. If a conflict arises between these two documents, the CSP will supersede. It is important to note that this plan does not list all of the potential hazards and their controls because the design and safety analysis process is still ongoing. Additional safety issues win be addressed in the Final Safety Analysis Report, Operational Safety Procedures (OSPs), and other plans and procedures as described in Section 3.0 of this plan.

NIF special equipment construction health and safety plan

International Nuclear Information System (INIS)

Sawicki, R.H.

1997-01-01

The purpose of this plan is to identify how the construction and deployment activities of the National Ignition Facility (NIF) Special Equipment (SE) will be safely executed. This plan includes an identification of (1) the safety-related responsibilities of the SE people and their interaction with other organizations involved; (2) safety related requirements, policies, and documentation; (3) a list of the potential hazards unique to SE systems and the mechanisms that will be implemented to control them to acceptable levels; (4) a summary of Environmental Safety and Health (ES ampersand H) training requirements; and (5) requirements of contractor safety plans that will be developed and used by all SE contractors participating in site activities. This plan is a subsidiary document to the NIF Construction Safety Program (CSP) and is intended to compliment the requirements stated therein with additional details specific to the safety needs of the SE construction-related activities. If a conflict arises between these two documents, the CSP will supersede. It is important to note that this plan does not list all of the potential hazards and their controls because the design and safety analysis process is still ongoing. Additional safety issues win be addressed in the Final Safety Analysis Report, Operational Safety Procedures (OSPs), and other plans and procedures as described in Section 3.0 of this plan

Safety related events at nuclear installations in 1995

DEFF Research Database (Denmark)

Korsbech, Uffe C C

1996-01-01

Nuclear safety related events of significance at least corresponding to level 2 of the International Nuclear Event Scale are described. In 1995 only two events occured at nuclear power plants, and four events occured at plants using ionizing radiation for processing or research.......Nuclear safety related events of significance at least corresponding to level 2 of the International Nuclear Event Scale are described. In 1995 only two events occured at nuclear power plants, and four events occured at plants using ionizing radiation for processing or research....

Selling safety: the use of celebrities in improving awareness of safety in commercial aviation.

Science.gov (United States)

Molesworth, Brett R C; Seneviratne, Dimuth; Burgess, Marion

2016-07-01

The aim of this study was to investigate the influential power of a celebrity to convey key safety messages in commercial aviation using a pre-flight safety briefing video. In addition, the present research sought to examine the effectiveness of subtitles in aiding the recall of these important messages as well as how in-cabin aircraft noise affects recall of this information. A total of 101 participants were randomly divided into four groups (no noise without subtitles, no noise with subtitles, noise without subtitles and noise with subtitles) and following exposure to a pre-recorded pre-flight safety briefing video were tested for recall of key safety messages within that video. Participants who recognised and recalled the name of the celebrity in the safety briefing video recalled significantly more of the messages than participants who did not recognise the celebrity. Subtitles were also found to be effective, however, only in the presence of representative in-cabin aircraft noise. Practitioner Summary: Passenger attention to pre-flight safety briefings on commercial aircraft is poor. Utilising the celebrity status of a famous person may overcome this problem. Results suggest that celebrities do increase the recall of safety-related information.

Some Subjects and Relations According to the Act about Safety at Work

Directory of Open Access Journals (Sweden)

Marino Đ. Učur

2015-01-01

Full Text Available Complex relations in the field of safety at work could not be present without the subjects which have a specific status and specific rights, obligations and responsibilities regulated by the Occupational Health and Safety Act. This paper deals with: employer’s designated employee for the implementation of occupational health and safety activities, employees’ elected representative for health and safety protection at work, occupational medicine specialist, occupational health and safety specialist and the committee for safety at work in the relations of safety at work.

Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446). Supplement No. 11

International Nuclear Information System (INIS)

1985-05-01

Supplement 11 to the Safety Evaluation Report for the Texas Utilities Electric Company application for a license to operate Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445, 50-446), located in Somervell County, Texas, has been jointly prepared by the Office of Nuclear Reactor Regulation and the Comanche Peak Technical Review Team (TRT) of the US Nuclear Regulatory Commission (NRC) and is in two parts. Part 1 (Appendix 0) of this supplement provides the results of the TRT's evaluation of approximately 124 concerns and allegations relating specifically to quality assurance and quality control (QA/QC) issues regarding construction proctices at the Comanche Peak facility. Part 2 (Appendix P) contains an overall summary and conclusion of the QA/QC aspects of the NRC Technical Review Team efforts as reported in supplemental Safety Evaluation Report SERs 7, 8, 9, and 10. Since QA/QC issues are also contained in each of the other supplements, the TRT considered that such a summary and conclusion from all supplements was necessary for a complete TRT description of QA/QC activities at Comanche Peak

Enhancement of reliability of PLT-safety devices by utilization of process control components; Steigerung der Verfuegbarkeit von PLT-Schutzeinrichtungen durch Mitbenutzung von Komponenten des Prozessleitsystems

Energy Technology Data Exchange (ETDEWEB)

Gabriel, T.; Litz, L. [Technische Univ. Kaiserslautern (Germany); Schroers, B. [Material-Science AG, Leverkusen (Germany)

2008-01-15

According to the standard IEC 61511 each safety-related loop is assigned to one of the four Safety Integrity Levels (SILs). For every safety-related loop a SIL-specific Probability of Failure on Demand (PFD) must be proven. Usually, the PFD calculation is performed based upon the failure rates of each loop component aided by commercial software tools. However, this bottom-up approach suffers from many uncertainties. Especially, a lack of reliable failure rate data causes many problems. Reference data collected in different environments are available to solve this situation. However, this pragmatism leads to a PFD bandwidth, not to a single PFD value as desired. In order to make a decision for a numerical value appropriate for the chemical and pharmaceutical process industry a data ascertainment has been initiated by the European NAMUR. Its results display large deficiencies for the bottom-up approach. The error sources leading to this situation are located and analyzed. (GL)

Test to prove the resistance to incidents of components of electric and control systems in the safety containment of nuclear power plants

International Nuclear Information System (INIS)

1982-01-01

The marginal program for proving the suitability of safety-relevant components of electric and control systems in the safety containment during a loss-of-coolant incident is described. Variant test conditions are established in the component-specific test program. Special attention has been paid to the representation of the course of pressure and temperature for the performance test of the valve room of the Nuclear Power Plant Philippsburg 2. (DG) [de

Components selection for ageing management

International Nuclear Information System (INIS)

Mingiuc, C.; Vidican, D.

2002-01-01

Full text: The paper presents a synthesis of methods and activities realized for the selection of critical components to assure plant safety and availability (as electricity supplier). There are presented main criteria for selection, screening process. For the resulted categories of components shall be applied different category of maintenance (condition oriented, scheduled or corrective), function of the importance and financial effort necessary to fulfil the task. 1. Systems and components screening for plant safety assurance For the systems selection, from Safety point of view, was necessary first, to define systems which are dangerous in case of failure (mainly by rupture/ release of radioactivity) and the safety systems which have to mitigate the effects. This is realized based on accident analysis (from Safety Report). Also where taken in to account the 4 basic Safety Principles: 'Reactor shut down; Residual heat removal; Radioactivity products confinement; NPP status monitoring in normal and accident conditions'. Following step is to establish safety support systems, which have to action to assure main safety systems operation. This could be realized based on engineering judgement, or on PSA Level I analysis. Finally shall be realized chains of the support systems, which have to work, till primary systems. For the critical components selection, was realized a Failure Mode and Effect Analysis (FMEA), considering the components effects of failures, on system safety function. 2. Systems and components screening for plant availability assurance The work was realized in two steps: Systems screening; Components screening The systems screening, included: General, analyze of the plant systems list and the definition of those which clearly have to run continue to assure the nominal power; Realization of a complex diagram to define interdependence between the systems (e.g. PHT and auxiliaries, moderator and auxiliaries, plant electrical diagram); Fill of special

Safety-related instrumentation and control systems for nuclear power plants

International Nuclear Information System (INIS)

1984-01-01

This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety but are not safety systems. The Guide is intended to expand paragraphs 3.1, 3.2 and 3.3 of the Code of Practice on Design for Safety of Nuclear Power Plants (IAEA Safety Series No.50-C-D) in the area of I and C systems important to safety and refers to them as safety-related I and C systems. It also gives guidance and enumerates requirements for multiplexing and the use of the digital computers employed in this area

An analysis of electronic health record-related patient safety incidents.

Science.gov (United States)

Palojoki, Sari; Mäkelä, Matti; Lehtonen, Lasse; Saranto, Kaija

2017-06-01

The aim of this study was to analyse electronic health record-related patient safety incidents in the patient safety incident reporting database in fully digital hospitals in Finland. We compare Finnish data to similar international data and discuss their content with regard to the literature. We analysed the types of electronic health record-related patient safety incidents that occurred at 23 hospitals during a 2-year period. A procedure of taxonomy mapping served to allow comparisons. This study represents a rare examination of patient safety risks in a fully digital environment. The proportion of electronic health record-related incidents was markedly higher in our study than in previous studies with similar data. Human-computer interaction problems were the most frequently reported. The results show the possibility of error arising from the complex interaction between clinicians and computers.

The effectiveness of a bicycle safety program for improving safety-related knowledge and behavior in young elementary students.

Science.gov (United States)

McLaughlin, Karen A; Glang, Ann

2010-05-01

The purpose of this study was to evaluate the "Bike Smart" program, an eHealth software program that teaches bicycle safety behaviors to young children. Participants were 206 elementary students in grades kindergarten to 3. A random control design was employed to evaluate the program, with students assigned to either the treatment condition (Bike Smart) or the control condition (a video on childhood safety). Outcome measures included computer-based knowledge items (safety rules, helmet placement, hazard discrimination) and a behavioral measure of helmet placement. Results demonstrated that regardless of gender, cohort, and grade the participants in the treatment group showed greater gains than control participants in both the computer-presented knowledge items (p > .01) and the observational helmet measure (p > .05). Findings suggest that the Bike Smart program can be a low cost, effective component of safety training packages that include both skills-based and experiential training.

Final summary report of the Nordic Nuclear Safety Research Program 1994 - 1997

International Nuclear Information System (INIS)

Bennerstedt, T.; Lemmens, A.

1999-11-01

This is a summary report of the NKS research program carried out 1994 - 1997. It is basically a compilation of the executive summaries of the final reports on the nine scientific projects carried out during that period. It highlights the conclusions, recommendations and other results of the projects. (au)

Technical Letter Report: Evaluation and Analysis of a Few International Periodic Safety Review Summary Reports

Energy Technology Data Exchange (ETDEWEB)

Chopra, Omesh K. [Argonne National Lab., IL (United States). Environmental Science Division; Diercks, Dwight R. [Argonne National Lab., IL (United States). Nuclear Engineering Division; Ma, David Chia-Chiun [Argonne National Lab., IL (United States). Environmental Science Division; Garud, Yogendra S. [Argonne National Lab., IL (United States). Environmental Science Division

2013-12-17

At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commission (NRC) incorporate lessons learned from periodic safety reviews (PSRs) performed in other countries as an input to the NRC’s assessment processes. In the U.S., commercial nuclear power plants (NPPs) are granted an initial 40-year operating license, which may be renewed for additional 20-year periods, subject to complying with regulatory requirements. The NRC has established a framework through its inspection, and operational experience processes to ensure the safe operation of licensed nuclear facilities on an ongoing basis. In contrast, most other countries do not impose a specific time limit on the operating licenses for NPPs, they instead require that the utility operating the plant perform PSRs, typically at approximately 10-year intervals, to assure continued safe operation until the next assessment. The staff contracted with Argonne National Laboratory (Argonne) to perform a pilot review of selected translated PSR assessment reports and related documentation from foreign nuclear regulatory authorities to identify any potential new regulatory insights regarding license renewal-related topics and NPP operating experience (OpE). A total of 14 PSR assessment documents from 9 countries were reviewed. For all of the countries except France, individual reports were provided for each of the plants reviewed. In the case of France, three reports were provided that reviewed the performance assessment of thirty-four 900-MWe reactors of similar design commissioned between 1978

Technical Letter Report: Evaluation and Analysis of a Few International Periodic Safety Review Summary Reports

International Nuclear Information System (INIS)

Chopra, Omesh K.; Diercks, Dwight R.; Ma, David Chia-Chiun; Garud, Yogendra S.

2013-01-01

At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commission (NRC) incorporate lessons learned from periodic safety reviews (PSRs) performed in other countries as an input to the NRC's assessment processes. In the U.S., commercial nuclear power plants (NPPs) are granted an initial 40-year operating license, which may be renewed for additional 20-year periods, subject to complying with regulatory requirements. The NRC has established a framework through its inspection, and operational experience processes to ensure the safe operation of licensed nuclear facilities on an ongoing basis. In contrast, most other countries do not impose a specific time limit on the operating licenses for NPPs, they instead require that the utility operating the plant perform PSRs, typically at approximately 10-year intervals, to assure continued safe operation until the next assessment. The staff contracted with Argonne National Laboratory (Argonne) to perform a pilot review of selected translated PSR assessment reports and related documentation from foreign nuclear regulatory authorities to identify any potential new regulatory insights regarding license renewal-related topics and NPP operating experience (OpE). A total of 14 PSR assessment documents from 9 countries were reviewed. For all of the countries except France, individual reports were provided for each of the plants reviewed. In the case of France, three reports were provided that reviewed the performance assessment of thirty-four 900-MWe reactors of similar design commissioned between 1978 and

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

International Nuclear Information System (INIS)

Lofgren, E.V.; Varcolik, F.

1982-11-01

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

Current State of Dental Education: Executive Summary.

Science.gov (United States)

Formicola, Allan J

2017-08-01

This executive summary for Section 1 of the "Advancing Dental Education in the 21 st Century" project provides a composite picture of information from 12 background articles on the current state of dental education in the United States. The summary includes the following topics: the current status of the dental curriculum, the implications of student debt and dental school finances, the expansion of enrollment, student diversity, pre- and postdoctoral education, safety net status of dental school clinics, and trends in faculty.

Method to classify the safety class of Structure, System and Components in a Defueled Condition of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Hak; Jeon, Dang-Hee [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

During pre-decommissioning phase, licensing and engineering work need to change the design basis of the plant such as safety analysis report, downgrade of systems, technical specifications and program and procedures to change of NPP condition from in an operation condition to in a defueled condition. The many systems to need to operate in an operational condition will not be operated during in a defueled condition and the function of systems will be changed from in an operation condition to in a defueled condition. So a downgrade of systems may be needed and reclassifying the safety class of structure, system and component (SSC) may be conducted. By the reclassification of SSC, activity related with quality assurance and maintenance of SSC is affected. In this paper, the method to reclassify SSC in a defueled condition is studied. The many systems to need to operate in an operational condition will not be operated during in a defueled condition and the function of systems will be changed from in an operation condition to in a defueled condition. The operation of NPP during a defueled condition need to conduct licensing and engineering work need to change the design basis of the plant optimize by downgrading systems and reclassifying the safety class of SSC. In this paper, the method to reclassify safety class for a defueled condition is studied.

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.

FY 1974 report on the results of the Sunshine Project. Study on a system to refine/transport/store hydrogen and the safety technology (Summary); 1974 nendo suiso no seisei yuso chozo system oyobi hoan gijutsu ni kansuru kenkyu seika hokokusho. Yoyakuban

Energy Technology Data Exchange (ETDEWEB)

NONE

1975-05-30

This study aims at the following. As to an alternative plan of the hydrogen distribution system, this paper makes the comprehensive evaluation from viewpoints of possibility, adaptability to society, economical efficiency, reliability, dynamic characteristics, etc., selects the most rational system and picked up the problems for realization of the system. It also sets up standards for safety technology relating to the hydrogen energy system and acquires the basic knowledge/information on concrete disaster prevention technology which backs up the standards. Among those, in this fiscal year, as system analysis of the hydrogen distribution system, study was made on system analysis of the system alternative plan and problems on system realization. Next, as an investigational study on safety technology in hydrogen energy system, the paper conducted the survey/collection in Japan and abroad of the present law, standards for handling, and examples of disaster/accident in relation to hydrogen, and the survey research on the concrete safety problem predicted. The paper studied the technology related to the safety problem required for safety. Industrial Research Institute, Japan, and the four companies under re-consignment made each report in which the results of the research were detailedly arranged. This paper is a summary of these reports. (NEDO)

Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

International Nuclear Information System (INIS)

Bourham, Mohamed A.; Gilligan, John G.

1999-01-01

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m 2 over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER

Component Fragility Research Program: Phase 1 component prioritization

International Nuclear Information System (INIS)

Holman, G.S.; Chou, C.K.

1987-06-01

Current probabilistic risk assessment (PRA) methods for nuclear power plants utilize seismic ''fragilities'' - probabilities of failure conditioned on the severity of seismic input motion - that are based largely on limited test data and on engineering judgment. Under the NRC Component Fragility Research Program (CFRP), the Lawrence Livermore National Laboratory (LLNL) has developed and demonstrated procedures for using test data to derive probabilistic fragility descriptions for mechanical and electrical components. As part of its CFRP activities, LLNL systematically identified and categorized components influencing plant safety in order to identify ''candidate'' components for future NRC testing. Plant systems relevant to safety were first identified; within each system components were then ranked according to their importance to overall system function and their anticipated seismic capacity. Highest priority for future testing was assigned to those ''very important'' components having ''low'' seismic capacity. This report describes the LLNL prioritization effort, which also included application of ''high-level'' qualification data as an alternate means of developing probabilistic fragility descriptions for PRA applications

Ranking related entities: components and analyses

NARCIS (Netherlands)

Bron, M.; Balog, K.; de Rijke, M.

2010-01-01

Related entity finding is the task of returning a ranked list of homepages of relevant entities of a specified type that need to engage in a given relationship with a given source entity. We propose a framework for addressing this task and perform a detailed analysis of four core components;

Topics raised by the summary report of the first review meeting

International Nuclear Information System (INIS)

2001-01-01

This document presents the fulfilling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 7 of the document contains topics raised by the summary report of the first review meeting. The chapter covers status and position of the regulatory body, collective doses, effluents release trends, adoption of the ICRP 60 and basic safety standards, regulatory strategy, quality assurance, monitoring safety management, international cooperation, probabilistic safety assessment and safety improvement programmes

IAEA activity related to safety of nuclear desalination

International Nuclear Information System (INIS)

Gasparini, M.

2000-01-01

The nuclear plants for desalination to be built in the future will have to meet the standards of safety required for the best nuclear power plants currently in operation or being designed. The current safety approach, based on the achievement of the fundamental safety functions and defence in depth strategy, has been shown to be a sound foundation for the safety and protection of public health, and gives the plant the capability of dealing with a large variety of sequences, even beyond the design basis. The Department of Nuclear Safety of the IAEA is involved in many activities, the most important of which are to establish safety standards, and to provide various safety services and technical knowledge in many Technical Co-operation assistance projects. The department is also involved in other safety areas, notably in the field of future reactors. The IAEA is carrying out a project on the safety of new generation reactors, including those used for desalination, with the objective of fostering an exchange of information on safety approaches, promoting harmonization among Member States and contributing towards the development and revision of safety standards and guidelines for nuclear power plant design. The safety, regulatory and environmental concerns in nuclear powered desalination are those related directly to nuclear power plants, with due consideration given to the coupling process. The protection of product water against radioactive contamination must be ensured. An effective infrastructure, including appropriate training, a legal framework and regulatory regime, is a prerequisite to considering use of nuclear power for desalination plants, also in those countries with limited industrial infrastructures and little experience in nuclear technology or safety. (author)

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

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

Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

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

Safety of Transmission in Railway Applications

Directory of Open Access Journals (Sweden)

Maria Franekova

2003-01-01

Full Text Available The paper deals with the problems of data security in safety - related transmission systems defined within railway process. It is intent on summary of treats and security tools against them within closed and open security systems. In details is analysed possibility of using channel coding techniques and cryptography mechanisms, which are defined according to norms EN 50159. For chosen security mechanisms are deal mathematical apparatus of probability of undetected error determination for channel decoders, relations for determination decipher encoder's complexity and error probability of cryptography code word.

Safety and design impact of hurricane Andrew

International Nuclear Information System (INIS)

Guey, Ching N.

2004-01-01

Turkey Point completed the IPE in June of 1991. Hurricane Andrew landed at Turkey Point on August 24, 1992. Although the safety related systems, components and structures were not damaged by the Hurricane Andrew, certain nonsafety related components and the neighboring fossil plant sustained noticeable damage. Among the major components that were nonsafety related but would affect the PRA of the plant included the service water pumps and the high tower. This paper discusses the safety and design impact of Hurricane Andrew on Turkey Point Nuclear Power Plant. The risk of hurricanes on the interim and evolving plant configurations are briefly described. The risk of the plant from internal events as a result of damage incurred during Hurricane Andrew are discussed. The design change as the result of Hurricane Andrew and its impact on the PRA are presented. (author)

Relation between water chemistry and operational safety

International Nuclear Information System (INIS)

Oliveira, M.F. de.

1991-01-01

This report describes the relation between chemistry/radiochemistry and operational safety, the technics bases for chemical and radiochemical parameters and an analysis of the Annual Report of Angra I Operation and OSRAT Mission report to 1989 in this area too. Furthermore it contains the transcription of the technical Specifications related to the chemistry and radiochemistry for Angra I. (author)

Metatextual Awareness, reading comprehension and story summary – possible relations in a psycholinguistic perspective

Directory of Open Access Journals (Sweden)

Marília Marques Lopes

2016-10-01

Full Text Available Even before formal education, children have a long term for developing their abilities of reading and writing, and these abilities relate to both reading comprehension and awareness of texts, whatever they are. This Metatextual Awareness, according to Jean-Émile Gombert (1992, comprises coherence, cohesion and text structure, and is a factor that can determine a good comprehension, as well as summary writing. The thesis this article is about employed Gombert’s theoretical framework on that subject, and dealt with correlations among metatextual awareness, reading comprehension and summary writing of narratives by 5th and 6th grade students in three public schools in Porto Alegre. To check these correlations, we applied simple choice tests to evaluate metatextual awareness and reading comprehension, and also a test of summary writing through a reference framework comprising categories like main events, writing autonomy and narrative structure. The results of these three aspects were related and we found that for summarizing a story the most important is the subject’s metatextual ability, or his attention to source text. At the same time, this metatextual awareness is also relevant to text comprehension as it is a kind of frame that guides him in understanding a story.

Definitions of engineered safety features and related features for nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

In light water moderated, light water cooled nuclear power plants, definitions are given of engineered safety features which are designed to suppress or prevent dispersion of radioactive materials due to damage etc. of fuel at the times of power plant failures, and of related features which are designed to actuate or operate the engineered safety features. Contents are the following: scope of engineered safety features and of related features; classification of engineered safety features (direct systems and indirect systems) and of related features (auxiliaries, emergency power supply, and protective means). (Mori, K.)

Waste Isolation Pilot Plant Safety Analysis Report

International Nuclear Information System (INIS)

1995-11-01

The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

Waste Isolation Pilot Plant Safety Analysis Report

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-11-01

The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

Safety considerations concerning light water reactors in Sweden

International Nuclear Information System (INIS)

Nilsson, T.

1977-01-01

In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

Quantification of the safety status of PWR steam generators

International Nuclear Information System (INIS)

Bartonicek, J.; Jonas, O.; Schoeckle, F.

1996-01-01

The operational safety of components is affected during the period of operation (e.g., by loads and environmental effects). Additionally, changes in codes and standards can affect the safety margin of a component. In order to determine a quantitative status of the safety, the existing status of the component and the actual knowledge have to be regarded and evaluated. Differences (deficits) with respect to actual demands must be balanced by use of redundant measure to guarantee safety. The procedure is demonstrated using the steam generators of GKN 1 (German nuclear power plant--in operation since 1975) as an example. The existing status of the component regarding design and construction, material, load history (results of monitoring), the results of nondestructive testing and operational experience is determined and related to the actual regulatory and code compliance and to the state of the art in technics. Non-specified loads were measured in operation. Stress and fatigue analysis computations are performed on the base of realistic load patterns (results from monitoring). Thus the safety status can be quantified. Finally the GKN-concept of measures to guarantee the reliability and the safety of the component in future operation is discussed

Plutonium safety training course

International Nuclear Information System (INIS)

Moe, H.J.

1976-03-01

This course seeks to achieve two objectives: to provide initial safety training for people just beginning work with plutonium, and to serve as a review and reference source for those already engaged in such work. Numerous references have been included to provide information sources for those wishing to pursue certain topics more fully. The first part of the course content deals with the general safety approach used in dealing with hazardous materials. Following is a discussion of the four properties of plutonium that lead to potential hazards: radioactivity, toxicity, nuclear properties, and spontaneous ignition. Next, the various hazards arising from these properties are treated. The relative hazards of both internal and external radiation sources are discussed, as well as the specific hazards when plutonium is the source. Similarly, the general hazards involved in a criticality, fire, or explosion are treated. Comments are made concerning the specific hazards when plutonium is involved. A brief summary comparison between the hazards of the transplutonium nuclides relative to 239 Pu follows. The final portion deals with control procedures with respect to contamination, internal and external exposure, nuclear safety, and fire protection. The philosophy and approach to emergency planning are also discussed

Summary of the activities of the ISAM Confidence Building Working Group

International Nuclear Information System (INIS)

Dolinar, G.M.; Batandjieva, B.

2002-01-01

During the early stages of the ISAM project confidence building was a relatively new topic in the radioactive waste disposal literature, but it was beginning to receive some attention. Although almost all safety assessment activities are intended to provide a level of confidence in the results of the assessment, considering the activities from the viewpoint of how they contributed to the decision making of various 'audiences' was relatively new. The ISAM project included the Confidence Building Working Group (CBWG) to examine the topic of Confidence Building and this paper provides a summary of the working group findings. (author)

Developments related to the National Nuclear Safety Authority of Romania

International Nuclear Information System (INIS)

Baciu, Florin

1998-01-01

The contribution presents the status of the National Commission for Nuclear Activity Control (CNCAN) as indicated by the provisions of a Romanian Government Decision of May 1998. As specified in the art.3 the main tasks of the Commission are the following: to issue authorization and exercise permits of activities in nuclear field; to supervise the applications of the provisions stipulated by the law concerning development in safety conditions of nuclear activities; to develop instructions as well as nuclear safety regulations to ensure the quality assurance and functioning in safety conditions of the nuclear facilities and plants, the protection against nuclear radiation of the professionally exposed personnel, of the population, of the environment and of the material goods, the physical protection, the records, preservation and transport of radioactive material and of fissionable materials as well as the management of radioactive waste; organizes expert and is responsible for the state control concerning the integrated application of the law provisions in the field of quality constructions in which nuclear installations of national interest are located, during all the phases and for all the components of the quality system in this field; issues specialty and information documentation specific to its own activity, provides the information of the public through official publication, official statements to the press and other specific form of information; carries out any other tasks provided by law in the field of regulations and control of nuclear activity. Author presents also the CNCAN staff number evolution, the new structure, the staff distribution at headquarters, local agencies and national radiation monitoring network. Finally, the author discusses the legal provisions related to management manual procedures

Nuclear safety regulation on nuclear safety equipment activities in relation to human and organizational factors

International Nuclear Information System (INIS)

Li Tianshu

2013-01-01

Based on years of knowledge in nuclear safety supervision and experience of investigating and dealing with violation events in repair welding of DFHM, this paper analyzes major faults in manufacturing and maintaining activities of nuclear safety equipment in relation to human and organizational factors. It could be deducted that human and organizational factors has definitely become key features in the development of nuclear energy and technology. Some feasible measures to reinforce supervision on nuclear safety equipment activities have also been proposed. (author)

Inpatient Prospective Payment System (IPPS) Provider Summary for the Top 100 Diagnosis-Related Groups (DRG)

Data.gov (United States)

U.S. Department of Health & Human Services — A provider level summary of Inpatient Prospective Payment System (IPPS) discharges, average charges and average Medicare payments for the Top 100 Diagnosis-Related...

Using Active Learning to Identify Health Information Technology Related Patient Safety Events.

Science.gov (United States)

Fong, Allan; Howe, Jessica L; Adams, Katharine T; Ratwani, Raj M

2017-01-18

The widespread adoption of health information technology (HIT) has led to new patient safety hazards that are often difficult to identify. Patient safety event reports, which are self-reported descriptions of safety hazards, provide one view of potential HIT-related safety events. However, identifying HIT-related reports can be challenging as they are often categorized under other more predominate clinical categories. This challenge of identifying HIT-related reports is exacerbated by the increasing number and complexity of reports which pose challenges to human annotators that must manually review reports. In this paper, we apply active learning techniques to support classification of patient safety event reports as HIT-related. We evaluated different strategies and demonstrated a 30% increase in average precision of a confirmatory sampling strategy over a baseline no active learning approach after 10 learning iterations.

Conference summaries of the Canadian Nuclear Association 30. annual conference, and the Canadian Nuclear Society 11. annual conference

International Nuclear Information System (INIS)

1990-01-01

This volume contains conference summaries for the 30. annual conference of the Canadian Nuclear Association, and the 11. annual conference of the Canadian Nuclear Society. Topics of discussion include: energy needs and challenges facing the Canadian nuclear industry; the environment and nuclear power; the problems of maintaining and developing industrial capacity; the challenges of the 1990's; programmes and issues for the 1990's; thermalhydraulics; reactor physics and fuel management; nuclear safety; small reactors; fuel behaviour; energy production and the environment; computer applications; nuclear systems; fusion; materials handling; and, reactor components

Work-related driver safety: A multi-level investigation

OpenAIRE

AMANDA ROSE WARMERDAM

2017-01-01

This program of research explored the organisational determinants of work-related road traffic injury in light vehicle fleets. The landscape of risk management in workplace road safety in Australia and organisational practices that influence safe driver behaviour were investigated. Key findings included that safe driving is influenced by factors at multiple levels, including senior managers, supervisors and individual fleet drivers and workplace road safety is not well integrated within curre...

Comparative analysis of safety related site characteristics

International Nuclear Information System (INIS)

Andersson, Johan

2010-12-01

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Comparative analysis of safety related site characteristics

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan (ed.)

2010-12-15

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

DART - for design basis justification and safety related information management

International Nuclear Information System (INIS)

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

2000-01-01

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

DART - for design basis justification and safety related information management

International Nuclear Information System (INIS)

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

2001-01-01

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

Proceedings of the NKS-R and NKS-B joint summary seminar

International Nuclear Information System (INIS)

Gwynn, J.P.; Isaksson, P.

2009-08-01

Nordic Nuclear Safety Research (NKS) is a platform for Nordic cooperation and competence in nuclear safety and related radiation protection issues including emergency preparedness and protection of the environment. Its purpose is to carry out joint activities producing seminars, exercises, scientific articles, technical reports and other types of reference material, with special efforts made to engage young scientists. The region in question is the five Nordic countries, i.e., Denmark (including the Faroe Islands and Greenland), Finland, Iceland, Norway and Sweden, who share a common cultural and historic heritage. The Nordic countries have cooperated in the field of nuclear safety for approximately half a century, developing informal networks for exchange of information, strengthening the region's potential for fast, coordinated and adequate response to nuclear threats, incidents and accidents. Activities are financed and supported by Nordic authorities, research institutions, power companies, contractors and other organizations, with results used by participating organizations in their decision making processes and information efforts.The NKS-R and NKS-B Joint Summary Seminar held at the Armemuseum, Stockholm on the 26th - 27th of March 2009 showcased a range of activities supported by NKS over the previous 2 years and provided an opportunity to bring together researchers and end users from the wider NKS community. This summary seminar was the first joint venture between the NKS-R and NKS-B Programmes since the current two programme format was adopted by NKS. One of our intentions in organising the Joint Summary Seminar was to further the reciprocal awareness of ongoing research and issues under the respective NKS-R and NKS-B Programmes, with the aim of promoting new networking opportunities and generating new ideas and approaches to solving existing problems. (au)

Proceedings of the NKS-R and NKS-B joint summary seminar

Energy Technology Data Exchange (ETDEWEB)

Gwynn, J.P.; Isaksson, P. (eds.) (Nordic Nuclear Safety Research, (NKS) (Denmark))

2009-08-15

Nordic Nuclear Safety Research (NKS) is a platform for Nordic cooperation and competence in nuclear safety and related radiation protection issues including emergency preparedness and protection of the environment. Its purpose is to carry out joint activities producing seminars, exercises, scientific articles, technical reports and other types of reference material, with special efforts made to engage young scientists. The region in question is the five Nordic countries, i.e., Denmark (including the Faroe Islands and Greenland), Finland, Iceland, Norway and Sweden, who share a common cultural and historic heritage. The Nordic countries have cooperated in the field of nuclear safety for approximately half a century, developing informal networks for exchange of information, strengthening the region's potential for fast, coordinated and adequate response to nuclear threats, incidents and accidents. Activities are financed and supported by Nordic authorities, research institutions, power companies, contractors and other organizations, with results used by participating organizations in their decision making processes and information efforts. The NKS-R and NKS-B Joint Summary Seminar held at the Armemuseum, Stockholm on the 26th - 27th of March 2009 showcased a range of activities supported by NKS over the previous 2 years and provided an opportunity to bring together researchers and end users from the wider NKS community. This summary seminar was the first joint venture between the NKS-R and NKS-B Programmes since the current two programme format was adopted by NKS. One of our intentions in organising the Joint Summary Seminar was to further the reciprocal awareness of ongoing research and issues under the respective NKS-R and NKS-B Programmes, with the aim of promoting new networking opportunities and generating new ideas and approaches to solving existing problems. (au)

Resistance ability evaluation of safety-related structures for the simulated aircraft accident

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Kim, Sung Woon; Choi, Jang Kyu [Daewoo E and C Co., Ltd., Suwon (Korea, Republic of)] (and others)

2003-03-15

Aircraft accidents on nuclear safety-related structures can cause severe damage to the safety of NPP(Nuclear Power Plant)s. To assess the safety of nuclear safety-related structures, the local damage and the dynamic response of global structures should be investigated together. This study have compared several local damage assessment formulas suggested for aircraft as an impactor, and have set the assessment system of local damage for impact-proof design of NPP containment buildings. And the local damage of nuclear safety-related structures in operation in Korea for commercial aircraft as impactor have been estimated. Impact load-time functions of the aircraft crash have been decided to assessment the safety of nuclear safety-related structures against the intentional colliding of commercial aircraft. Boeing 747 and Boeing 767 is selected as target aircraft based on the operation frequencies and weights. Comparison of the fire analysis methods showed that the method considering heat convection and radiation is adequate for the temperature analysis of the aircraft fuel fire. Finally, the study covered the analysis of the major structural drawings and design drawings with which three-dimensional finite element model analysis is expected to be performed.

The orthopaedic error index: development and application of a novel national indicator for assessing the relative safety of hospital care using a cross-sectional approach.

Science.gov (United States)

Panesar, Sukhmeet S; Netuveli, Gopalakrishnan; Carson-Stevens, Andrew; Javad, Sundas; Patel, Bhavesh; Parry, Gareth; Donaldson, Liam J; Sheikh, Aziz

2013-11-21

The Orthopaedic Error Index for hospitals aims to provide the first national assessment of the relative safety of provision of orthopaedic surgery. Cross-sectional study (retrospective analysis of records in a database). The National Reporting and Learning System is the largest national repository of patient-safety incidents in the world with over eight million error reports. It offers a unique opportunity to develop novel approaches to enhancing patient safety, including investigating the relative safety of different healthcare providers and specialties. We extracted all orthopaedic error reports from the system over 1 year (2009-2010). The Orthopaedic Error Index was calculated as a sum of the error propensity and severity. All relevant hospitals offering orthopaedic surgery in England were then ranked by this metric to identify possible outliers that warrant further attention. 155 hospitals reported 48 971 orthopaedic-related patient-safety incidents. The mean Orthopaedic Error Index was 7.09/year (SD 2.72); five hospitals were identified as outliers. Three of these units were specialist tertiary hospitals carrying out complex surgery; the remaining two outlier hospitals had unusually high Orthopaedic Error Indexes: mean 14.46 (SD 0.29) and 15.29 (SD 0.51), respectively. The Orthopaedic Error Index has enabled identification of hospitals that may be putting patients at disproportionate risk of orthopaedic-related iatrogenic harm and which therefore warrant further investigation. It provides the prototype of a summary index of harm to enable surveillance of unsafe care over time across institutions. Further validation and scrutiny of the method will be required to assess its potential to be extended to other hospital specialties in the UK and also internationally to other health systems that have comparable national databases of patient-safety incidents.

Nuclear Safety Project

International Nuclear Information System (INIS)

1978-11-01

The 13th semi-annual report 1/78 is a description of work within the Nuclear Safety Project performed in the first six months of 1978 in the nuclear safety field by KFK institutes and departments and by external institutions on behalf of KfK. It includes for each individual research activity short summaries on - work completed, - essential results, - plans for the near future. (orig./RW) [de

Software FMEA analysis for safety-related application software

International Nuclear Information System (INIS)

Park, Gee-Yong; Kim, Dong Hoon; Lee, Dong Young

2014-01-01

Highlights: • We develop a modified FMEA analysis suited for applying to software architecture. • A template for failure modes on a specific software language is established. • A detailed-level software FMEA analysis on nuclear safety software is presented. - Abstract: A method of a software safety analysis is described in this paper for safety-related application software. The target software system is a software code installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system (DRPS). For the ATIP software safety analysis, at first, an overall safety or hazard analysis is performed over the software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA analysis is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA analysis, being applied to the ATIP software code, which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) that could not be identified during various system tests

Improving operational safety management through probabilistic safety assessment on personal computers

International Nuclear Information System (INIS)

1988-10-01

The Technical Committee Meeting considered the current effort in the implementation and use of PSA information for day-to-day operational safety management on Personal Computers. Due to the very recent development of the necessary hardware and software for Personal Computers, the application of PSA information for day-to-day operational safety management on PCs is essentially still in a pioneering stage. There is at present only one such system for end users existing, the PRISIM (Plant Risk Status Information Management) program for which a limited practical application experience is available. Others are still in the development stage. The main aim of the Technical Committee Meeting was to discuss the present status of PSA based systems for operational safety management support on small computers, to consider practical aspects when implementing these systems into a nuclear installation and to address problems related to the further work in the area. A separate abstract was prepared for the summary of the Technical Committee Meeting and for the 8 papers presented by the participants. Refs, figs and tabs

Reliability of containment and safety-related structures

International Nuclear Information System (INIS)

Nessim, M.A.

1995-09-01

A research program on Reliability of Containment and Safety-related Structures has been developed and is described in this document. This program is designed to support AECB's regulatory activities aimed at ensuring the safety of these structures. These activities include evaluating submissions by operators and requesting special assessments when necessary. The results of the proposed research will also be useful in revising and enhancing the CSA design standards for containment and safety-related structures. The process of developing the research program started with an information collection and review phase. The sources of information included C-FER's previous work in the area, various recent research publications, regulatory documents and relevant design standards, and a detailed discussion with AECB staff. The second step was to outline the process of reliability evaluation, and identify the required models and parameters. Comparison between the required and available information was used to identify gaps in the state-of-the-art, and the research program was designed to fill these gaps. The program is organized in four major topics, namely: development of an approach for reliability analysis; compilation and development of the required analysis tools; application to specific problems related to design, assessment, maintenance and testing of structures; and testing and validation. It is suggested that the program should be supported by an on-going process of communication and consultation between AECB staff and industry experts. This will lend credibility to the results and facilitate their future application. (author). 1 fig

Ninth ITER technical meeting on safety and environment

International Nuclear Information System (INIS)

Raeder, J.; Gordon, C.

2001-01-01

The ninth ITER Technical Meeting on safety and environment, the last in the course of the ITER Engineering Design Activities (EDA), was held at the ITER Garching Joint Work site, 8 to 10 May 2001. At this Meeting, safety experts from the House Teams worked together with the members of the Safety, Environment and Health Groups (SEHG) of the ITER Joint Central Team (JCT) in the following areas: finalization of the Generic Site Safety report (GSSR) which is considered to be the most important objective of the present work; summary of the safety related R and D work done by the Home Teams for ITER during EDA; review of verification and validation work done on computer codes being applied for Safety and Environment (S and E) analyses; outline of work considered necessary for improving the S and E database, quantifying uncertainties of the code results and preparing the adaptation of ITER to a specific site

Constrained principal component analysis and related techniques

CERN Document Server

Takane, Yoshio

2013-01-01

In multivariate data analysis, regression techniques predict one set of variables from another while principal component analysis (PCA) finds a subspace of minimal dimensionality that captures the largest variability in the data. How can regression analysis and PCA be combined in a beneficial way? Why and when is it a good idea to combine them? What kind of benefits are we getting from them? Addressing these questions, Constrained Principal Component Analysis and Related Techniques shows how constrained PCA (CPCA) offers a unified framework for these approaches.The book begins with four concre

Nuclear power plants. Electrical equipment of the safety system. Qualification

International Nuclear Information System (INIS)

2001-01-01

This International Standard applies to electrical parts of safety systems employed at nuclear power plants, including components and equipment of any interface whose failure could affect unfavourably properties of the safety system. The standard also applies to non-electrical safety-related interfaces. Furthermore, the standard describes the generic process of qualification certification procedures and methods of qualification testing and related documentation. (P.A.)

Safety status of Russian research reactors

International Nuclear Information System (INIS)

Morozov, S.I.

2001-01-01

Gosatomnadzor of Russia is conducting the safety regulation and inspection activity related to nuclear and radiation safety at nuclear research facilities, including research reactors, critical assemblies and sub-critical assemblies. It implies implementing three major activities: 1) establishing the laws and safety standards in the field of research reactors nuclear and radiation safety; 2) research reactors licensing; and 3) inspections (or license conditions tracking and inspection). The database on nuclear research facilities has recently been updated based on the actual status of all facilities. It turned out that many facilities have been shutdown, whether temporary or permanently, waiting for the final decision on their decommissioning. Compared to previous years the situation has been inevitably changing. Now we have 99 nuclear research facilities in total under Gosatomnadzor of Russia supervision (compared to 113 in previous years). Their distribution by types and operating organizations is presented. The licensing and conduct of inspection processes are briefly outlined with emphasis being made on specific issues related to major incidents that happened in 2000, spent fuel management, occupational exposure, effluents and emissions, emergency preparedness and physical protection. Finally, a summary of problems at current Russian research facilities is outlined. (author)

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

Increased component safety through improved methods for residual stress analysis. Subprojects. Consideration of real component geometries (phase 1). Final report

International Nuclear Information System (INIS)

Nau, Andreas; Scholtes, B.

2014-01-01

Residual stresses can be result in both detrimental as well as beneficial consequences on the component's strength and lifetime. A most detailed knowledge of the residual stress state is a pre-requisite for the assessment of the component's performance. The mechanical methods for residual stress measurements are classified in non-destructive, destructive and semi-destructive methods. The two commonly used (semi-destructive) mechanical methods are the hole drilling and the ring core method. In the context of reactor safety research of the Federal Ministry of Economic Affairs and Energy (BMWi), two fundamental and interacting weak points of the hole drilling as well as of the ring core method are investigated. On the one hand, there are effects concerning geometrical boundary conditions of the components and on the other hand, there are influences of plasticity due to notch effects. Both aspects affect the released strain field, when the material is removed and finally, the calculated residual stresses. The first issue mentioned above is under the responsibility of Institute of Materials Engineering - Metallic Materials (Kassel University) and the last one will be investigated by University of Stuttgart-Otto-Graf-Institut - materials testing institute. Within the framework of this project it could be demonstrated that updated calibration coefficients lead to more reliable residual stress calculation in contrast to existing ones. These findings are valid for points of measurements on components without geometrical boundary effects like edges or shoulders. Reasons are high developed Finite-Element software packages and the opportunity of modelling the point of measurement (hole geometry, layout of the strain gauges) and its vicinity more in detail. Special challenges are multi-axial residual stress depth distributions and the geometry of components composing edges and claddings. Unlike existing analyses considering uni-axial and homogeneous stress states, bi

Analysis of safety culture components based on site interviews

International Nuclear Information System (INIS)

Ueno, Akira; Nagano, Yuko; Matsuura, Shojiro

2002-01-01

Safety culture of an organization is influenced by many factors such as employee's moral, safety policy of top management and questioning attitude among site staff. First this paper analyzes key factors of safety culture on the basis of site interviews. Then the paper presents a safety culture composite model and its applicability in various contexts. (author)

Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

International Nuclear Information System (INIS)

Lowe, Andrew; Hayward, Brent

2006-08-01

This report documents a study with the objective of enhancing safety culture in the Swedish nuclear power industry. A primary objective of this study was to ensure that the latest thinking on human factors principles was being recognised and applied by nuclear power operators as a means of ensuring optimal safety performance. The initial phase of the project was conducted as a pilot study, involving the senior management group at one Swedish nuclear power-producing site. The pilot study enabled the project methodology to be validated after which it was repeated at other Swedish nuclear power industry sites, providing a broad-ranging analysis of opportunities across the industry to enhance safety culture. The introduction to this report contains an overview of safety culture, explains the background to the project and sets out the project rationale and objectives. The methodology used for understanding and analysing the important safety culture issues at each nuclear power site is then described. This section begins with a summary of the processes used in the information gathering and data analysis stage. The six components of the Management Workshops conducted at each site are then described. These workshops used a series of presentations, interactive events and group exercises to: (a) provide feedback to site managers on the safety culture and safety leadership issues identified at their site, and (b) stimulate further safety thinking and provide 'take-away' information and leadership strategies that could be applied to promote safety culture improvements. Section 3, project Findings, contains the main observations and output from the project. These include: - a brief overview of aspects of the local industry operating context that impinge on safety culture; - a summary of strengths or positive attributes observed within the safety culture of the Swedish nuclear industry; - a set of identified opportunities for further improvement; - the aggregated results of the