1E Qualification of Electrical Equipment - Requirement for Safety Nuclear Power Plants

International Nuclear Information System (INIS)

Geambasu, C.; Segarceanu, D.; Albu, J.

2002-01-01

The paper presents the qualification methods of the safety related equipment according to the safety class 1E. There are presented the qualification principles, procedure and documents, emphasis being laid on the qualification approach by type tests. This approach assumes the equipment test under both normal and accident conditions (design basis events) simulating the operational conditions and covers the largest part of electrical equipment from a nuclear power plant.The safety related equipment is to be qualified is subjected to a sequential test that will be detailed in the paper. (author)

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

Procurement strategic analysis of nuclear safety equipment

International Nuclear Information System (INIS)

Wu Caixia; Yang Haifeng; Li Xiaoyang; Li Shixin

2013-01-01

The nuclear power development plan in China puts forward a challenge on procurement of nuclear safety equipment. Based on the characteristics of the procurement of nuclear safety equipment, requirements are raised for procurement process, including further clarification of equipment technical specification, establishment and improvement of the expert database of the nuclear power industry, adoption of more reasonable evaluation method and establishment of a unified platform for nuclear power plants to procure nuclear safety equipment. This paper makes recommendation of procurement strategy for nuclear power production enterprises from following aspects, making a plan of procurement progress, dividing procurement packages rationally, establishing supplier database through qualification review and implementing classified management, promoting localization process of key equipment continually and further improving the system and mechanism of procurement of nuclear safety equipment. (authors)

Guidelines for evaluation of anchorage adequacy for safety-related equipment typically used on WWER-type NPPs

International Nuclear Information System (INIS)

Masopust, R.

1999-01-01

This report describes the criteria which should be met when the capacity evaluation of anchorage of safety related equipment is performed for the WWER type NPPs. It should be noted that these criteria were developed specifically for anchorage of WWER type equipment and components to the concrete or steel building structures and they cover different types of anchor bolts and other anchorage details which are typical just for the existing, constructed or reconstructed WWER type NPPs. The screening approach for verifying of equipment anchorage presented in this report is based on a combination of inspections, calculations, and engineering judgement

49 CFR 238.230 - Safety appliances-new equipment.

Science.gov (United States)

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false Safety appliances-new equipment. 238.230 Section... I Passenger Equipment § 238.230 Safety appliances—new equipment. (a) Applicability. This section..., 2007, that is equipped with a safety appliance, required by the “manner of application” provisions in...

Analysis of Medical Equipment Management in Relation to the Mandatory Medical Equipment Safety Manager (MESM in Japan

Directory of Open Access Journals (Sweden)

Kai Ishida

2014-01-01

Full Text Available Half a decade has passed since the fifth revision of the medical law and mandatory appointment of a medical equipment safety manager (MESM in hospitals in Japan. During this period, circumstances have changed regarding maintenance of medical equipment (ME. We conducted a survey to examine these changes and the current situation in ME management. Maintenance of ME and related work were found to have increased in many hospitals, but the number of clinical engineering technologists (CETs has only slightly increased. The appointed MESM was a CET or physician in most hospitals. In hospitals where physicians were appointed as the MESM, 81% had operation managers. Many respondents commented that it was difficult for one person to cover all the tasks required by the MESM, due to a lack of knowledge, too much work, or other reasons. This suggests the importance of an operation manager for ME to work under the MESM.

Regulatory Activities on Civil Nuclear Safety Equipment in China

International Nuclear Information System (INIS)

Gaoshang, Lu; Choi, Kwang Sik

2011-01-01

It is stipulated in IAEA Fundamental Safety Principles (SF1) that the fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. The fundamental safety objective applies for all facilities and activities and for all stages over the lifetime of a facility or radiation source, including planning, sitting, design, manufacturing, construction, commissioning and operation, as well as decommissioning and closure. So, according to the requirement, the related activities such as design, manufacturing, installation and non-destructive test that conducted on civil nuclear equipment should be well controlled by the vendors, the owner of the nuclear power plants and the regulatory body. To insure the quality of those equipment, Chinese government had taken a series of measures to regulate the related activities on them

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.)

Margins related to equipment design

International Nuclear Information System (INIS)

Devos, J.

1994-01-01

Safety margins related to design of reactor equipment are defined according to safety regulations. Advanced best estimate methods are proposed including some examples which were computed and compared to experimental results. Best estimate methods require greater computation effort and more material data but give better variable accuracy and need careful experimental validation. Simplified methods compared to the previous are less sensitive to material data, sometimes are more accurate but very long to elaborate

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

Plutonium finishing plant safety systems and equipment list

International Nuclear Information System (INIS)

Bergquist, G.G.

1995-01-01

The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex

Navigation Tools and Equipment and How They Have Improved Aviation Safety

OpenAIRE

Sulaiman D. S Alsahli FadalahassanALfadala

2017-01-01

This paper highlights the impact of navigation tools and equipment, such as the GPS, navigation radar, and other communications tools, which aid in ensuring aviation safety. It emphasizes the need for aviation safety and how these navigation methods are of great help to reduce the hazards and clearly indicate the problems related to the aircraft, aircraft traffic management, weather disturbances, among others. It also recommends how these tools and equipment must be further developed to promo...

Screening of external hazards for NPP with bank type reactor. Modeling of safety related systems and equipment for RBMK. Probabilistic assessment of NPP safety on aircraft impact. Progress report

International Nuclear Information System (INIS)

Kostarev, V.

1999-01-01

This progress report was produced within the frame of IAEA research project on screening the hazards for NPP with bank type reactor. It covers the following tasks; development of the model for the primary loop system of RBMK; developing the models for safety related equipment of RBMK; developing of models for safety related models of EGP-6 type reactor (Bilibinskaya Nuclear Co-generated heat and Power Plant); and probabilistic assessment of NPP safety on aircraft impact

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

Evaluation of Generic Issue 57: Effects of fire protection system actuation on safety-related equipment

International Nuclear Information System (INIS)

Lambright, J.; Bohn, M.; Lynch, J.; Ross, S.; Brosseau, D.

1992-12-01

Nuclear power plants have experienced actuations of fire protection systems (FPSs) under conditions for which these systems were not intended to actuate and also have experienced advertent actuations with the presence of a fire. These actuations have often damaged safety-related equipment. A review of the impact of past occurrences of both types of such events and their impact on plant safety systems, an analysis of the risk impacts of such events on nuclear power plant safety, and a cost-benefit analysis of potential corrective measures have been performed. Thirteen different scenarios leading to actuation of fire protection systems due to a variety of causes were identified. These scenarios ranged from inadvertent actuation caused by human error to hardware failure, and include seismic root causes and seismic/fire interactions. A quantification of these thirteen root causes, where applicable, was performed on generically applicable scenarios. This document, Volume 4, contains appendices E and F of this report

Comparison and Analysis of IEEE 344 and IEC 60980 standards for harmonization of seismic qualification of safety-related equipment

International Nuclear Information System (INIS)

Lee, Young Ok; Kim, Jong Seog; Seo, Jeong Ho; Kim, Myung Jun

2011-01-01

The seismic qualification of safety related equipment in nuclear power plants should demonstrate an equipment's ability to perform its safety function during/or after the time it is subjected to the forces resulting from one SSE. In addition, the equipment must withstand the effects of a number of OBEs, preceding the SSE. IEEE 344 and IEC 60980 present the criteria for establishing procedures demonstrating that the Class 1E equipment can meet its performance requirement during seismic events. Currently, IEEE 344 is used for regulation of nuclear power plant in the United State whereas IEC 60980 is mainly used in Europe. In particular, NPPs of France and China apply with RCC-E and GB that are domestic standards, respectively. Equipment supplier and Utility have difficulties because of different applicable standards. Equipment supplier to export S/R components/equipment to other standard area performs additional seismic qualification. For example, equipment are qualifies according to IEC 60980, RCC-E, GB although they have been qualified in accordance with IEEE 344. Also, utility to attempt power up-rate, life extension of NPP constructed under rules of RCC-E such as Ulchin NPP 1 and 2 has similar difficulties. RCC-E endorses IEC 60980 and GB is almost same as IEC 60980 except minor difference of earthquake environment definition. Therefore this paper surveys the similarities and differences between IEEE 344 and IEC 60980. In addition, this paper considers how the two sets of standards may be used in a complementary fashion to be possible using one or the other standard area

Guideline on evaluation and acceptance of commercial grade digital equipment for nuclear safety applications

International Nuclear Information System (INIS)

1996-10-01

Nuclear power plants are increasingly upgrading their instrumentation and control (I ampersand C) systems with commercial digital equipment, which allows them to continue meeting safety and reliability requirements while controlling operating costs. However, the use of commercial software-based devices for safety related applications has raised new issues that impact design, procurement, and licensing activities. This guideline describes a consistent, comprehensive approach for the evaluation and acceptance of commercial digital equipment for nuclear safety systems

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

International Nuclear Information System (INIS)

Anon.

1980-01-01

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

30 CFR 250.806 - Safety and pollution prevention equipment quality assurance requirements.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 2 2010-07-01 2010-07-01 false Safety and pollution prevention equipment... Gas Production Safety Systems § 250.806 Safety and pollution prevention equipment quality assurance... install only certified safety and pollution prevention equipment (SPPE) in wells located on the OCS. SPPE...

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

46 CFR 189.60-10 - Cargo Ship Safety Equipment Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Cargo Ship Safety Equipment Certificate. 189.60-10... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 189.60-10 Cargo Ship Safety Equipment Certificate. (a) All vessels on an international voyage...

46 CFR 91.60-10 - Cargo Ship Safety Equipment Certificate.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Cargo Ship Safety Equipment Certificate. 91.60-10... VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 91.60-10 Cargo Ship Safety Equipment Certificate. (a) All vessels on an international voyage are...

Regulatory analysis for the resolution of generic issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment

International Nuclear Information System (INIS)

Woods, H.W.

1993-10-01

Actuation of Fire Protection Systems (FPS) in Nuclear Power Plants have resulted in adverse interactions with equipment important to safety. Precursor operational experience has shown that 37% of all FPS actuations damaged some equipment, and 20% of all FPS actuations have resulted in a plant transient and reactor trip. On an average 0.17 FPS actuations per reactor year have been experienced in nuclear power plants in this country. This report presents the regulatory analysis for GI-57, ''Effects of Fire Protection System Actuation on Safety-Related Equipment''. The risk reduction estimates, cost/benefit analyses, and other insights gained during this effort have shown that implementation of the recommendations contained in this report can significantly reduce risk, and that these improvements can be warranted in accordance with the backfit rule, 10 CFR 50.109(a)(3). However, plant specific analyses are required in order to identify such improvements. Generic analyses can not serve to identify improvements that could be warranted for individual, specific plants. Plant specific analyses of the type needed for this purpose are underway as part of the Individual Plant Examination of External Events (IPEEE) program

Experience of safety and performance improvement for fuel handling equipment

International Nuclear Information System (INIS)

Gyoon Chang, Sang; Hee Lee, Dae

2014-01-01

The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

Control of Nuclear Materials and Special Equipment (Nuclear Safety Regulations)

International Nuclear Information System (INIS)

Cizmek, A.; Prah, M.; Medakovic, S.; Ilijas, B.

2008-01-01

Based on Nuclear Safety Act (OG 173/03) the State Office for Nuclear Safety (SONS) in 2008 adopted beside Ordinance on performing nuclear activities (OG 74/06) and Ordinance on special conditions for individual activities to be performed by expert organizations which perform activities in the area of nuclear safety (OG 74/06) the new Ordinance on the control of nuclear material and special equipment (OG 15/08). Ordinance on the control of nuclear material and special equipment lays down the list of nuclear materials and special equipment as well as of nuclear activities covered by the system of control of production of special equipment and non-nuclear material, the procedure for notifying the intention to and filing the application for a license to carry out nuclear activities, and the format and contents of the forms for doing so. This Ordinance also lays down the manner in which nuclear material records have to be kept, the procedure for notifying the State administration organization (regulatory body) responsible for nuclear safety by the nuclear material user, and the keeping of registers of nuclear activities, nuclear material and special equipment by the State administration organization (regulatory body) responsible for nuclear safety, as well as the form and content of official nuclear safety inspector identification card and badge.(author)

Safety requirements and feedback of commonly used material handling equipment

International Nuclear Information System (INIS)

Pathak, M.K.

2009-01-01

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

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

Science.gov (United States)

2010-10-01

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

Safety code 19: recommended safety procedures for the selection, installation and use of x-ray diffraction equipment

International Nuclear Information System (INIS)

1984-01-01

This document is one of a series of Safety Codes prepared by the Radiation Protection Bureau to set out requirements for the safe use of radiation emitting devices. The equipment and installation guidelines and safety procedures detailed in this Code are primarily for the instruction and guidance of persons employed in Federal Public Service Departments and Agencies, as well as those coming under the jurisdiction of the Canada Labour Code. This Safety Code is also intended to assist other users of X-ray diffraction equipment to select safe equipment and to install and use it so that the radiation hazard to the operator and other persons in its vicinity is negligible. It should be noted that facilities under provincial jurisdiction may be subject to requirements specified under provincial statutes. This Code supersedes Safety Code RPD-SC-7, entitled 'Requirements For Non-Medical X-Ray Equipment, Use and Installation', insofar as X-ray diffraction equipment is concerned, and it is intended to complement X-ray equipment design, construction and performance standards promulgated under the Radiation Emitting Devices Act

The french safety approach regarding equipment qualification

International Nuclear Information System (INIS)

Jalouneix, J.; Deletre, G.; Henry, J.Y.; Rousseau, L.

1989-10-01

The qualification of the equipment, used in France on nuclear reactors and reactor components, is discussed. The approach on safety involves the qualification of mechanical and electrical components, under operating and under seismic conditions. The principles of the qualification program and the list of the equipment concerned, are given. The conditions in which qualification tests are performed and the methods applied, are summarized. The actions for preserving the qualification procedures, are considered

Comparison of safety equipment between London underground and Beijing subway

Science.gov (United States)

Chen, T.; Zhang, S. Y.; Zhao, L. Z.; Xia, J. J.; Fu, X. C.; Bao, Z. M.; Chen, Y.; Zhang, X. Z.; Wang, R. J.; Hu, C.; Jing, L. S.; Wang, Y.

2017-06-01

The purpose of this paper was to improve the safety equipment’s effectiveness through the comparison. Firstly, the history and safety accident of London Underground and Beijing Subway were shown. Secondly, fire equipment between these two cities was compared including station’s hardware installations and carriage’s hardware installations. Thirdly, the relative software installations were also compared such as emergency drills. The results showed that Beijing Subway’s hardware installations were better than London. However, London Underground’s some installations were more effective than Beijing. Both cities would pay more attention on anti-terrorist in tunnel.

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

Experience on environmental qualification of safety-related components for Darlington Nuclear Generating Station

International Nuclear Information System (INIS)

Yu, A.S.; Kukreti, B.M.

1987-01-01

The proliferation of Nuclear Power Plant safety concerns has lead to increasing attention over the Environmental Qualification (EQ) of Nuclear Power Plant Safety-Related Components to provide the assurance that the safety related equipment will meet their intended functions during normal operation and postulated accident conditions. The environmental qualification of these components is also a Licensing requirement for Darlington Nuclear Generating Station. This paper provides an overview of EQ and the experience of a pilot project, in the qualification of the Main Moderator System safety-related functions for the Darlington Nuclear Generating Station currently under construction. It addresses the various phases of qualification from the identification of the EQ Safety-Related Components List, definition of location specific service conditions (normal, adbnormal and accident), safety-related functions, Environmental Qualification Assessments and finally, an EQ system summary report for the Main Moderator System. The results of the pilot project are discussed and the methodology reviewed. The paper concludes that the EQ Program developed for Darlington Nuclear Generating Station, as applied to the qualification of the Main Moderator System, contained all the elements necessary in the qualification of safety-related equipment. The approach taken in the qualification of the Moderator safety-related equipment proves to provide a sound framework for the qualification of other safety-related components in the station

Design review report for modifications to RMCS safety class equipment

International Nuclear Information System (INIS)

Corbett, J.E.

1997-01-01

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

Design review report for modifications to RMCS safety class equipment

Energy Technology Data Exchange (ETDEWEB)

Corbett, J.E.

1997-05-30

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

Safety for Compressed Gas and Air Equipment. Module SH-26. Safety and Health.

Science.gov (United States)

Center for Occupational Research and Development, Inc., Waco, TX.

This student module on safety for compressed gas and air equipment is one of 50 modules concerned with job safety and health. This module presents technical data about commonly used gases and stresses the procedures necessary for safe handling of compressed gases. Following the introduction, 14 objectives (each keyed to a page in the text) the…

The effect of using road safety equipment and systems and ...

African Journals Online (AJOL)

The effect of using road safety equipment and systems and determine their role on ... traffic control equipment situation and by multi-criteria weighting systems AHP ... The results have shown that indices median, lighting and panel type and the ...

The use of probabilistic safety assessment 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 order 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 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)

A safety equipment list for rotary mode core sampling systems operation in single shell flammable gas tanks

International Nuclear Information System (INIS)

SMALLEY, J.L.

1999-01-01

This document identifies all interim safety equipment to be used for rotary mode core sampling of single-shell flammable gas tanks utilizing Rotary Mode Core Sampling systems (RMCS). This document provides the safety equipment for RMCS trucks HO-68K-4600, HO-68K-4647, trucks three and four respectively, and associated equipment. It is not intended to replace or supersede WHC-SD-WM-SEL-023, (Kelly 1991), or WHC-SD-WM-SEL-032, (Corbett 1994), which classifies 80-68K-4344 and HO-68K-4345 respectively. The term ''safety equipment'' refers to safety class (SC) and safety significant (SS) equipment, where equipment refers to structures, systems and components (SSC's). The identification of safety equipment in this document is based on the credited design safety features and analysis contained in the Authorization Basis (AB) for rotary mode core sampling operations in single-shell flammable gas tanks. This is an interim safety classification since the AB is interim. This document will be updated to reflect the final RMCS equipment safety classification designations upon completion of a final AB which will be implemented with the release of the Final Safety Analysis Report (FSAR)

Safety systems I/C equipment reliability analyses of the Kozloduy NPP units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Halev, G; Christov, N [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

The purpose of the analysis is to assess the safety systems I/C equipment reliability. The assessment includes: quantification of the safety systems unavailability due to component failures; definition of the minimal cut sets leading to the analysed safety systems failure; quantification of the I/C equipment importance measures of the dominant contribution components. The safety systems I/C equipment reliability has been analysed using PSAPACK (a code for probabilistic safety assessment). Fault trees for the following safety systems of the Kozloduy-3 and Kozloduy-4 reactors have been constructed: neutron flow control equipment, reactor protection system, main coolant pumps, pressurizer safety valves `Sempell`, steam dump systems, spray system, low pressure injection system, emergency feeding water system, essential service water system. THree separate reports have been issued containing the performed analyses and results. 1 ref.

38 CFR 17.155 - Minimum standards of safety and quality for automotive adaptive equipment.

Science.gov (United States)

2010-07-01

... safety and quality for automotive adaptive equipment. 17.155 Section 17.155 Pensions, Bonuses, and... Minimum standards of safety and quality for automotive adaptive equipment. (a) The Under Secretary for... officials that it meets implicit standards of safety and quality adopted by the industry or as later...

Safety equipment and methods for evaluating its effectiveness

Energy Technology Data Exchange (ETDEWEB)

Evdokimov, F I; Nadtoka, T B [DPI (Ukraine)

1993-05-01

Analyzes relations between technologies (especially for roof support) used in black coal mining and work safety in mines. The share of manual work and accident rate are compared for mining by narrow and wide web shearer loaders and by coal plows with powered and individual support. Protection from occupational injury is discussed at three levels: safety engineering, work organization and the human factor. A method of evaluating the social and economic effectiveness of protection from occupational injury developed at the DPI institute is presented. The method uses the knowledge of probability distribution of failure situations, failures and protective means to determine the probabilistic characteristics of the functioning of protection systems and to calculate, for a given period, the occurrence probability and mean number of accidents. Each state of the system is characterized by determined social and/or economic results. The method was used in designing equipment intended for protective power cut-off in electric mine networks.

Safety of nursing staff and determinants of adherence to personal protective equipment.

Science.gov (United States)

Neves, Heliny Carneiro Cunha; Souza, Adenícia Custódia Silva e; Medeiros, Marcelo; Munari, Denize Bouttelet; Ribeiro, Luana Cássia Miranda; Tipple, Anaclara Ferreira Veiga

2011-01-01

A qualitative study conducted in a teaching hospital with 15 nursing professionals. Attempted to analyze the reasons, attitudes and beliefs of nursing staff regarding adherence to personal protective equipment. Data were collected through focus groups, analyzed by the method of interpretation of meanings, considering Rosenstock's model of health beliefs as a reference framework. Data revealed two themes: Occupational safety and Interpersonal Relationship. We identified several barriers that interfere in matters of safety and personal protective equipment, such as communication, work overload, physical structure, accessibility of protective equipment and organizational and management aspects. Adherence to personal protective equipment is determined by the context experienced in the workplace, as well as by individual values and beliefs, but the decision to use the personal protective equipment is individual.

Equipment line-up developed for structuring programmed digital systems important to safety

International Nuclear Information System (INIS)

Remus, L.; Colling, J.M.; Buisson, J.

1986-01-01

The use of the programmed digital techniques in systems important to safety led to developing equipment line-ups which allow to procedure, through using a restricted number of sub-assemblies, a variety of facilities capable of performing diversified functions. Upon remembering the requirements that such equipment are required to meet (quality assurance both at the design and production stages, qualification, testing opportunities, etc.) the equipments produced by MERLIN GERIN are described in this paper. Such equipments include a whole set of printed circuit boards, plug-in units with self-engaging connectors and the related cabinets to accomodate them. The following systems are made up of such equipments: Digital integrated protection system; Rod control and monitoring system; Reactor power distribution monitoring equipment. This equipment line-up has been gradually complemented and is still further expanding both as a result of the component further development or for the purpose of meeting new requirements or applications (new digital neutron instrumentation, local area networks). Experience gained thus far provides us with the capability of working out in a short time programmed systems for further applications in the nuclear power field

A safety equipment list for rotary mode core sampling systems operation in single shell flammable gas tanks; TOPICAL

International Nuclear Information System (INIS)

SMALLEY, J.L.

1999-01-01

This document identifies all interim safety equipment to be used for rotary mode core sampling of single-shell flammable gas tanks utilizing Rotary Mode Core Sampling systems (RMCS). This document provides the safety equipment for RMCS trucks HO-68K-4600, HO-68K-4647, trucks three and four respectively, and associated equipment. It is not intended to replace or supersede WHC-SD-WM-SEL-023, (Kelly 1991), or WHC-SD-WM-SEL-032, (Corbett 1994), which classifies 80-68K-4344 and HO-68K-4345 respectively. The term ''safety equipment'' refers to safety class (SC) and safety significant (SS) equipment, where equipment refers to structures, systems and components (SSC's). The identification of safety equipment in this document is based on the credited design safety features and analysis contained in the Authorization Basis (AB) for rotary mode core sampling operations in single-shell flammable gas tanks. This is an interim safety classification since the AB is interim. This document will be updated to reflect the final RMCS equipment safety classification designations upon completion of a final AB which will be implemented with the release of the Final Safety Analysis Report (FSAR)

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)

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)

Standardization and improvement of safety for radioisotope equipped instruments

International Nuclear Information System (INIS)

Sumi, Tetsuo

1980-01-01

The safety for radioisotope-equipped instruments is considered. The one is the safety for the source assembly. The radioisotopes employed for radioisotope-equipped instruments are sealed sources which are used in the state of being contained in the enclosures. Many of the enclosures are provided with shutter mechanism for the purpose of emitting radiation only during the period required. If the possible troubles that might lead to the accidents are sampled out of the results of field operation of radiation instruments, and the safety measures for source enclosures are considered in connection with these troubles, it is no exaggeration to say that the safety for source enclosures has been maintained by preventing the critical accidents by the management of users and the cooperation of manufactures though there were the chance for investigating the safety in the common field and the establishment of JIS Z 4614 standard. Another consideration is concerned with the measures to improve the safety. No accident in the past never guarantees no accident in the future. Accumulation of experience is most effective for those measures, and the more experiences the better. It may be most effective that the manufacturers disclose their experiences each other from the wide outlook overcoming the barrier of trade secret. Fortunately, such consciousness has risen since a few years ago, and the investigation group is doing the works in the Japan Radioisotope Association. On the other hand, the reasonable revision of the radiation injury prevention law is desired. (Wakatsuki, Y.)

L-038: EPR-First Responders: Forces / safety equipment. Action Guides

International Nuclear Information System (INIS)

2011-01-01

This conference is about the actions carry out by the forces and the safety equipment in a radiological emergency. The security area, the victims, the hospitals, the police vehicles area, the safety cordon, the evacuation, the contamination level and the risk of life are important aspects to be considered by the first responders.

Evidence for the safety assurance of the radiation associated equipments for stat-up check

International Nuclear Information System (INIS)

Takahashi, Yasuyuki; Igarashi, Hiroshi; Saito, Kyoko; Kawaharada, Yasuhiro; Hirano, Kunihiro; Murase, Kenya; Mochizuki, Teruhito

2008-01-01

Based on the new amendment of the Medical Service Law and the Pharmaceutical Affairs Law, the safety control for the medical equipments and devices was strengthened; and a radiological technologist should be stipulated as a safety control manager for the medical equipments and devices. Although it is thought that Start-up check is important for the safety assurance, the enforcement situation is not clear until now. In this report, on the occasion of the start-up of the new Law, questionnaire about the start up check (time and a check items) were investigated from 56 hospitals applying the visit interview. Modality by modality check up time was obtained from 261 institutions by the descriptive study paper type. Start-up check and arming up were applied for almost all modalities about 30 minutes before start of the clinical tests. The check up items for the medical equipments and devices depended on each hospital. Therefore, safety assurance should be established, cooperating with the medical equipment distributors, the hospital sharing the checked results for the patients' safety. (author)

Practical approaches to implementing facility wide equipment strengthening programs

International Nuclear Information System (INIS)

Kincaid, R.H.; Smietana, E.A.

1989-01-01

Equipment strengthening programs typically focus on components required to ensure operability of safety related equipment or to prevent the release of toxic substances. Survival of non-safety related equipment may also be crucial to ensure rapid recovery and minimize business interruption losses. Implementing a strengthening program for non-safety related equipment can be difficult due to the large amounts of equipment involved and limited budget availability. EQE has successfully implemented comprehensive equipment strengthening programs for a number of California corporations. Many of the lessons learned from these projects are applicable to DOE facilities. These include techniques for prioritizing equipment and three general methodologies for anchoring equipment. Pros and cons of each anchorage approach are presented along with typical equipment strengthening costs

Safety analysis report for packaging (onsite) decontaminated equipment self-container

International Nuclear Information System (INIS)

Boehnke, W.M.

1998-01-01

The purpose of this Safety Analysis Report for Packaging (SARP) is to demonstrate that specific decontaminated equipment can be safely used as its own self-container. As a Decontaminated Equipment Self-Container (also referred to as a self-container), no other packaging, such as a burial box, would be required to transport the equipment onsite. The self-container will consist of a piece of equipment or apparatus which has all readily removable interior contamination removed, all of its external openings sealed, and all external surfaces decontaminated to less than 2000 dpm/100 cm for gamma-emitting radionuclides and less than 220 dpm/100 CM2 for alpha-emitting radionuclides

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

International Nuclear Information System (INIS)

Roh, Young Hoon; Kim, Jung Min

2015-01-01

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

Improvement suggestions on license extension management for civil nuclear safety equipment activities

International Nuclear Information System (INIS)

Sun Xingjian; Liu Hongji; Han Guoli; Jia Fengcai

2013-01-01

Based on the concepts of Clear Requirements, Comprehensive Review, Objective Assessment, Dynamic Management, this paper gives improvement suggestions on license extension management for civil nuclear safety equipment design, manufacture, installation and non-destructive examination activities, which include establishing a relatively unified license extension review standard, combining multi-views and close linking license review and supervision, full utilizing the daily supervision and inspection results, as well as further improving motivation and elimination mechanism. (authors)

A study on enforcement effects of radiation safety control regulations for diagnostic X-ray equipment

International Nuclear Information System (INIS)

Sung, Mo IL; Park, Myeong Hwan; Kwon, Duk Moon; Lee, Joon IL

1999-01-01

The purposes of this study are to analyze the realities after enforcements of safety control regulations for diagnostic X-ray equipment and to suggest means for an improvement of low radiation safety control. A questionnaire survey for medical radiologic technologists was carried out to determine enforcement effects of the safety control regulations. The results of analysis from the survey are as follows. That is, most of he respondents realized the importance of the radiation safety control system, but about a half of them revealed that regulations were not well observed in accordance with their purposes. Only 43.9 percent of the respondents took an active part in quality control of radiation. And responsibility, sex, age, and knowledge for safety control were important indicators for observations of the regulations. Training for the safety control regulations are needed to ensure safety control and proper usage of diagnostic X-ray equipment. And management of organizations using diagnostic X-ray equipment have to understand and stress the importance of radiation safety control system. (author)

Nuclear safety: a large scale quality audit of pressurized equipment

International Nuclear Information System (INIS)

Faudon, Valerie

2016-01-01

This article notably refers to, quotes and comments a hearing organised by the French Public Office for the Assessment Scientific and Technological Choices (OPECST) on the issue of safety of pressurized equipment in nuclear reactors, and which gathered the main concerned actors (Areva, EDF, IRSN, ASN) to have an overview of quality controls in AREVA NP fabrication plants. Two different issues have been addressed: a technical metallurgical issue related to some boiler-making parts, and an issue related to quality assurance. These issues concern different older reactors (Fessenheim for example) as well as new ones (EPR Flamanville). The article indicates the different measures planned, envisaged or already implemented by the concerned actors in order to improve knowledge in the boiler-making industry, and to ensure a better quality

Evaluation of safety requirements of erbium laser equipment used in dentistry

International Nuclear Information System (INIS)

Braga, Flavio Hamilton

2002-01-01

The erbium laser (Er:YAG) has been used in several therapeutic processes. Erbium lasers, however, operate with energies capable to produce lesions in biological tissues. Aiming the safe use, the commercialization of therapeutic laser equipment is controlled in Brazil, where the equipment should comply with quality and safety requirement prescribed in technical regulations. The objective of this work is to evaluate the quality and safety requirements of a commercial therapeutic erbium laser according to Brazilian regulations, and to discuss a risk control program intended to minimize the accidental exposition at dangerous laser radiation levels. It was verified that the analyzed laser can produce lesions in the skin and eyes, when exposed to laser radiation at distances smaller than 80 cm by 10 s or more. In these conditions, the use of protection glasses is recommended to the personnel that have access to the laser operation ambient. It was verified that the user's training and the presence of a target indicator are fundamental to avoid damages in the skin and buccal cavity. It was also verified that the knowledge and the correct use of the equipment safety devices, and the application of technical and administrative measures is efficient to minimize the risk of dangerous expositions to the laser radiation. (author)

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

26 CFR 1.187-1 - Amortization of certain coal mine safety equipment.

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 3 2010-04-01 2010-04-01 false Amortization of certain coal mine safety... (CONTINUED) INCOME TAX (CONTINUED) INCOME TAXES Itemized Deductions for Individuals and Corporations (continued) § 1.187-1 Amortization of certain coal mine safety equipment. (a) Allowance of deduction—(1) In...

Factors Influencing the Safety Behavior of German Equestrians: Attitudes towards Protective Equipment and Peer Behaviors

Directory of Open Access Journals (Sweden)

Christina-Maria Ikinger

2016-02-01

Full Text Available Human interactions with horses entail certain risks. Although the acceptance and use of protective gear is increasing, a high number of incidents and very low or inconsistent voluntary use of safety equipment are reported. While past studies have examined factors influencing the use of safety gear, they have explored neither their influence on the overall safety behavior, nor their relative influence in relation to each other. The aim of the present study is to fill this gap. We conducted an online survey with 2572 participants. By means of a subsequent multiple regression analysis, we explored 23 different variables in view of their influence on the protective behavior of equestrians. In total, we found 17 variables that exerted a significant influence. The results show that both having positive or negative attitudes towards safety products as well as the protective behavior of other horse owners or riding pupils from the stable have the strongest influence on the safety behavior of German equestrians. We consider such knowledge to be important for both scientists and practitioners, such as producers of protective gear or horse sport associations who might alter safety behavior in such a way that the number of horse-related injuries decreases in the long term.

INDIVIDUAL AND ORGANIZATIONAL VARIABLES AFFECTING THE USE OF PERSONAL PROTECTIVE EQUIPMENT

Directory of Open Access Journals (Sweden)

TUNÇ DEMİRBİLEK

2013-06-01

Full Text Available Management’s safety approach and commitment are related to occupational health and safety applications in workplaces. However, personal characteristics and attitudes on their health of employees are very influential factors on occupational health and safety applications. For examine of interaction between behaviors and the health and safety applications it is required that determine of employee’s perceptions. For this reason this paper’s purpose is to establish of safety needs, attitudes and behaviors of employees and their perceptions regarding management’s safety approach and applications. The study’s focal point is use of personal protective equipment as an indicator “safety behavior”. In empirical study, assumption variables regarding the use of personal protective equipment are safety effectively, safety need, management attitudes and protective equipment conditions. The fundamental hypothesis of this paper is use of personal protective equipment is related to individual and organizational variables. Implications of research findings are to indicate impelling the workers to safety behaviour or using the personal protective equipment by their feeling of the safety behaviour and availability of personal protective equipment.

A Handbook for Public Playground Safety. Volume II: Technical Guidelines for Equipment and Surfacing.

Science.gov (United States)

Consumer Product Safety Commission, Washington, DC.

This handbook suggests safety guidelines for public playground equipment and describes various surfaces used under the equipment and possible injuries resulting from falls. The handbook is intended for use mainly by manufacturers, installers, school and park officials, and others interested in technical criteria for public playground equipment.…

An overview of equipment survivability studies at Sandia National Laboratories (SNL)

International Nuclear Information System (INIS)

Bonzon, L.L.; Craft, Ch.M.; McCulloch, W.H.; Sebrell, W.A.

1983-01-01

The USNRC sponsors a number of programs at Sandia National Laboratories (SNL) specifically addressing safety-related equipment survivability. The major thrust of these programs has been the physical testing of equipment. Test results illustrate the importance of a dedicated equipment design effort giving particular attention to the safety implications of the equipment operation. Several equipment survivability tests here have revealed equipment design and test-related deficiencies

Risk-based evaluation tool for safety-related maintenance involving scaffolding

International Nuclear Information System (INIS)

Stevens, C.; Azizi, M.; Massman, M.

1988-01-01

The US Nuclear Regulatory Commission (NRC) has expressed a general concern that transient materials in and around safety systems at nuclear power plants represent a seismic safety hazard to the plant, in particular, the uncontrolled use of scaffolding during maintenance activities. Currently, most plants perform a seismic safety analysis for all uses of scaffolding near safety-related equipment to determine appropriate tie-down locations, scaffolding reinforcements, etc. This is both time-consuming and, for the most part, unnecessary. A workable engineering solution based on risk analysis techniques has been developed and is being used at the Palo Verde nuclear generating station (PVNGS)

46 CFR 31.40-10 - Cargo Ship Safety Equipment Certificate-T/ALL.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Cargo Ship Safety Equipment Certificate-T/ALL. 31.40-10 Section 31.40-10 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY TANK VESSELS INSPECTION AND CERTIFICATION Certificates Under International Convention for Safety of Life at Sea, 1974 § 31.40-10 Cargo Ship...

An on-line monitoring system for navigation equipment

Science.gov (United States)

Wang, Bo; Yang, Ping; Liu, Jing; Yang, Zhengbo; Liang, Fei

2017-10-01

Civil air navigation equipment is the most important infrastructure of Civil Aviation, which is closely related to flight safety. In addition to regular flight inspection, navigation equipment's patrol measuring, maintenance measuring, running measuring under special weather conditions are the important means of ensuring aviation flight safety. According to the safety maintenance requirements of Civil Aviation Air Traffic Control navigation equipment, this paper developed one on-line monitoring system with independent intellectual property rights for navigation equipment, the system breakthroughs the key technologies of measuring navigation equipment on-line including Instrument Landing System (ILS) and VHF Omni-directional Range (VOR), which also meets the requirements of navigation equipment ground measurement set by the ICAO DOC 8071, it provides technical means of the ground on-line measurement for navigation equipment, improves the safety of navigation equipment operation, and reduces the impact of measuring navigation equipment on airport operation.

Used energy-related laboratory equipment grant program for institutions of higher learning. Eligible equipment catalog

Energy Technology Data Exchange (ETDEWEB)

1994-07-01

This is a listing of energy related equipment available through the Energy-Related Laboratory Equipment Grant Program which grants used equipment to institutions of higher education for energy-related research. Information included is an overview of the program, how to apply for a grant of equipment, eligibility requirements, types of equipment available, and the costs for the institution.

Seismic qualification method of equipment for nuclear power plant

International Nuclear Information System (INIS)

Kim, J.S.; Choi, T.H.; Sulaimana, R.A.

1995-01-01

Safety related equipment installed in Korean Nuclear Power Plants are required to perform a safety function during and after a seismic event. To accomplish this safety function, they must be seismically qualified in accordance with the intent and requirements of the USNRC Reg. Guide 1.100 Rev. 02 and IEEE Std. 344-1987. This paper defines and summarizes acceptable criteria and procedures, based on the Korean experience, for seismic qualification of purchased equipment to be installed in a nuclear power plant. As such the paper is intended to be a concise reference by equipment designers, architectural engineering company and plant owners in uniform implementation of commitments to nuclear regulatory agencies such as the USNRC or Korea Institute of Nuclear Safety (KINS) relating to adequacy of seismic Category 1 equipment. Thus, the paper provides the methodologies which can be used for qualifying equipment for safely related service in Nuclear Power Plants in a cost effective manner

Commercial-grade motors in safety-related applications: Final report

International Nuclear Information System (INIS)

Holzman, P.M.

1988-04-01

The objective of this project was to discuss the process necessary to utilize commercial grade equipment in safety related applications and to provide utilities with guidance for accepting commercial grade motors for safety-related applications. The generic commercial-grade concepts presented in this report can be successfully applied to motors. Commercial grade item utilization has the greatest applicability to motors in ''mild'' environments, because these motors are essentially similar to commercial grade motors in materials, construction methods, and capabilities. The acceptance process is less applicable to motors that are subject to ''harsh'' environments during postulated accidents, because of the unique design features and testing required to qualify these motors

Selection of equipment for equipment qualification

International Nuclear Information System (INIS)

Torr, K.G.

1989-01-01

This report describes the methodology applied in selecting equipment in the special safety systems for equipment qualification in the CANDU 600 MW nuclear generating stations at Gentilly 2 and Point Lepreau. Included is an explanation of the selection procedure adopted and the rationale behind the criteria used in identifying the equipment. The equipment items on the list have been grouped into three priority categories as a planning aid to AECB staff for a review of the qualification status of the special safety systems

Equipment qualification testing - a practical approach

International Nuclear Information System (INIS)

Davies, G.A.; McDougall, R.I.; Poirier, M.P.

1996-01-01

When nuclear safety equipment is credited with a Required Safety Function it must properly perform that function to facilitate safe control and/or shutdown of the plant during a design basis accident. When such equipment is required to be environmentally (EQ) and/or seismically qualified (SQ) for safety related use in CANDU nuclear power plants, the preferred method of qualification is by type testing. The qualification testing process requires that the test specimen equipment be subjected to the aging stressors associated with the normal service conditions that it would experience during it's required qualified (or service) life. Following the aging process, the test specimen is in a condition representative of that in which it would be at the end of its service life in the plant. The test specimen is then subjected to a simulated accident during which it must satisfy performance requirements thereby demonstrating that it can perform its required safety function. The performance requirements specified for the qualification testing must be designed to ensure that satisfactory performance of the safety function is demonstrated during the qualification program. This paper provides descriptions of practical methods used in the deriving and satisfying of relevant performance requirements during the qualification testing of safety related equipment. (author)

The basic discussion on nuclear power safety improvement based on nuclear equipment design

International Nuclear Information System (INIS)

Zhao Feiyun; Yao Yangui; Yu Hao; He Yinbiao; Gao Lei; Yao Weida

2013-01-01

The safety of strengthening nuclear power design was described based on nuclear equipment design after Fukushima nuclear accident. From these aspects, such as advanced standard system, advanced design method, suitable test means, consideration of beyond design basis event, and nuclear safety culture construction, the importance of nuclear safety improvement was emphatically presented. The enlightenment was given to nuclear power designer. (authors)

Discussion on unpacking inspection of imported civil nuclear safety equipment

International Nuclear Information System (INIS)

Li Chan; Zhang Wenguang; Li Maolin; Li Shixin; Jin Gang; Yao Yuan

2014-01-01

This paper introduces the purpose, contents, process and requirements of unpacking inspection which is the second stage of safety inspection of imported civil nuclear safety equipment, expresses review key points on application documents of unpacking inspection, processes of witness on-the-spot before unpacking inspection outside the civil nuclear facilities by the test agency, discusses understanding of unpacking inspection, supervision of manufacture, inspection prior to shipment, supervision of loading and acceptance by the unit operating civil nuclear facilities, reports on unpacking inspection. Some suggestions on reinforcing the unpacking inspection in China are concerned. (authors)

An overview of equipment survivability studies at Sandia National Labs. (SNL): Chapter 3

International Nuclear Information System (INIS)

Bonzon, L.L.; Craft, C.M.; McCulloch, W.H.; Sebrell, W.A.

1983-01-01

The USNRC sponsors a number of programs at Sandia National Laboratories (SNL) specifically addressing safety-related equipment survivability. The major thrust of these programs has been the physical testing of equipment. Test results illustrate the importance of a dedicated equipment design effort giving particular attention to the safety implications of the equipment operation. Several equipment survivability tests here have revealed equipment design and test-related deficiencies

Development of a Test Equipment for Performance Evaluation of Safety Systems

International Nuclear Information System (INIS)

Kim, S. J.; Kwon, S. M.; Lee, J. M.; Kim, C. K.; Cho, C. H.; Chun, J. H.; Park, M. K.

2004-07-01

The purpose of this study is to develop a test equipment for performance evaluation of safety systems in nuclear power plants. First, we develop an input-output simulator for reactor protection systems, ESF component control systems, and a data acquisition system for these I/O simulators as a hardware for this equipment. Then, we develop a software for human-machine interface system, which is easy-to-use and easy-to-modify. In addition, a simulation tool for a reactor trip switch gear is developed

Spent nuclear fuel project cold vacuum drying facility safety equipment list

International Nuclear Information System (INIS)

IRWIN, J.J.

1999-01-01

This document provides the safety equipment list (SEL) for the Cold Vacuum Drying Facility (CVDF). The SEL was prepared in accordance with the procedure for safety structures, systems, and components (SSCs) in HNF-PRO-516, ''Safety Structures, Systems, and Components,'' Revision 0 and HNF-PRO-097, Engineering Design and Evaluation, Revision 0. The SEL was developed in conjunction with HNF-SO-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998). The SEL identifies the SSCs and their safety functions, the design basis accidents for which they are required to perform, the design criteria, codes and standards, and quality assurance requirements that are required for establishing the safety design basis of the SSCs. This SEL has been developed for the CVDF Phase 2 Safety Analysis Report (SAR) and shall be updated, expanded, and revised in accordance with future phases of the CVDF SAR until the CVDF final SAR is approved

Material & equipment, procurement & maintenance: Impact on blood safety.

Science.gov (United States)

Emmanuel, Jean C

2010-01-01

Blood Transfusion Safety is dependent on effectively organised and managed blood services, which have adequate financial resources, skilled manpower, appropriate infrastructure and quality management systems in place. 80% of the world's population has access to 20% of the supply blood products, of which little is consistently safe. HIV highlighted the importance of blood safety. The lack of effective blood services in low human development index (LHDI), developing countries, has lead to international funding and capacity building for more than three decades. The initial strategies focused on providing HIV testing reagents to prevention transmission, however this only addresses one part of blood safety. Blood safety is not only dependent on preventing HIV transmission. In many populations there are other infectious agents, which have a higher prevalence. Ensuring the correct blood is provided to the patient depends on: well managed services with effective leadership and adequate budgets; capacity building and retention of skilled experienced staff; availability of laboratory equipment, correctly maintained; blood cold chain systems; procedures for tendering, purchasing and ensuring an unbroken supply of reagents and consumables; and quality management systems. Barriers for simplified effective tendering, procurement and contracting require urgent attention and coordination of all funding organisations to ensure an unbroken supply of reagents. Copyright 2009. Published by Elsevier Ltd.

Unresolved Safety Issue A-46 - seismic qualification of equipment in operating plants

International Nuclear Information System (INIS)

Anderson, N.

1985-01-01

Seismic Qualification of Equipment in Operating Plants was designated as an Unresolved Safety Issue (USI) in December, 1980. The USI A-46 program was developed in early 1981 to investigate the adequacy of mechanical and electrical equipment in operating plants to withstand a safe shutdown earthquake. The approach taken was to develop viable, cost effective alternatives to current seismic qualification licensing requirements which could be applied to operating nuclear power plants. The tasks investigated include: (1) identification of seismic sensitive systems and equipment; (2) assessment of adequacy of existing seismic qualification methods; (3) development and assessment of in-situ test procedures to assist in qualification of equipment; (4) seismic qualification of equipment using seismic experience data; and (5) development of methods to generate generic floor response spectra. Progress to date and plans for completion of resolution are reported

Evaluation of Electrical Characteristics of Protective Equipment - a Prerequisite for Ensuring Safety and Health of Workers at Work

Science.gov (United States)

Buică, G.; Beiu, C.; Antonov, A.; Dobra, R.; Păsculescu, D.

2017-06-01

The protecting electrical equipment in use are subject to various factors generated by the use, maintenance, storage and working environment, which may change the characteristics of protection against electric shock. The study presents the results of research on the behaviour over time of protective characteristics of insulating covers of material of work equipment in use, in order to determine the type and periodicity of safety tests. There were tested and evaluated safety equipment with plastic and insulating rubber covers used in operations of verifying functionality, safety and maintenance of machinery used in manufacturing industries and specific services from electric, energy and food sector.

[Prevention of medical device-related adverse events in hospitals: Specifying the recommendations of the German Coalition for Patient Safety (APS) for users and operators of anaesthesia equipment].

Science.gov (United States)

Bohnet-Joschko, Sabine; Zippel, Claus; Siebert, Hartmut

2015-01-01

The use and organisation of medical technology has an important role to play for patient and user safety in anaesthesia. Specification of the recommendations of the German Coalition for Patient Safety (APS) for users and operators of anaesthesia equipment, explore opportunities and challenges for the safe use and organisation of anaesthesia devices. We conducted a literature search in Medline/PubMed for studies dealing with the APS recommendations for the prevention of medical device-related risks in the context of anaesthesia. In addition, we performed an internet search for reports and recommendations focusing on the use and organisation of medical devices in anaesthesia. Identified studies were grouped and assigned to the recommendations. The division into users and operators was maintained. Instruction and training in anaesthesia machines is sometimes of minor importance. Failure to perform functional testing seems to be a common cause of critical incidents in anaesthesia. There is a potential for reporting to the federal authority. Starting points for the safe operation of anaesthetic devices can be identified, in particular, at the interface of staff, organisation, and (anaesthesia) technology. The APS recommendations provide valuable information on promoting the safe use of medical devices and organisation in anaesthesia. The focus will be on risks relating to the application as well as on principles and materials for the safe operation of anaesthesia equipment. Copyright © 2015. Published by Elsevier GmbH.

Changing nature of equipment and parts qualification

International Nuclear Information System (INIS)

Bucci, R.M.

1988-01-01

Ideally, the original supplier of a piece of nuclear safety-related equipment has performed a qualification program and will continue to support that equipment throughout the lifetime of the nuclear power plants in which in equipment is installed. The supplier's nuclear quality assurance program will be maintained and he will continue to offer all necessary replacement parts. These parts will be identical to the original parts, certified to the original purchase order requirements, and the parts will be offered at competitive prices. Due to the changing nature of the nuclear plant equipment market, however, one or more of those ideal features are frequently unavailable when safety-related replacement equipment or parts are required. Thus, the process of equipment and parts qualification has had to adjust in order to ensure obtaining qualified replacements when needed. This paper presents some new directions taken in the qualification of replacement equipment and parts to meet changes in the marketplace

Radiation protection in dentistry. Recommended safety procedures for the use of dental x-ray equipment. Safety code 30

International Nuclear Information System (INIS)

1994-01-01

The Radiation Protection Bureau has prepared a series of documents on safety codes to set out requirements for the safe use of radiation-emitting equipment. This Safety Code has been prepared to provide specific guidance to the dentist, dental hygienist, dental assistant and other support personnel concerned with safety procedures and equipment performance. Dental radiography is one of the most valuable tools used in modern dental health care. It makes possible the diagnosis of physical conditions that would otherwise be difficult to identify. The use of dental radiological procedures must be carefully managed, because x-radiation has the potential for damaging healthy cells and tissues. Although no known occurrence of cancer or genetic damage has been observed from radiation doses delivered in modern dentistry, and until more evidence is available, one should practice radiation hygiene with the same care as would be dictated if a hazard were known to exist. The aim of radiation protection in dentistry is to obtain the desired clinical information with minimal radiation exposure to patients, dental personnel and the public. 15 tabs

Radiation protection in dentistry. Recommended safety procedures for the use of dental x-ray equipment. Safety code 30

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

The Radiation Protection Bureau has prepared a series of documents on safety codes to set out requirements for the safe use of radiation-emitting equipment. This Safety Code has been prepared to provide specific guidance to the dentist, dental hygienist, dental assistant and other support personnel concerned with safety procedures and equipment performance. Dental radiography is one of the most valuable tools used in modern dental health care. It makes possible the diagnosis of physical conditions that would otherwise be difficult to identify. The use of dental radiological procedures must be carefully managed, because x-radiation has the potential for damaging healthy cells and tissues. Although no known occurrence of cancer or genetic damage has been observed from radiation doses delivered in modern dentistry, and until more evidence is available, one should practice radiation hygiene with the same care as would be dictated if a hazard were known to exist. The aim of radiation protection in dentistry is to obtain the desired clinical information with minimal radiation exposure to patients, dental personnel and the public. 15 tabs.

High field MRI: safety and protection problems related to the installation and the management of a 4T equipment

International Nuclear Information System (INIS)

Russo, A.; Speranza, A.; DELIA, R.; ANTOLINI, R.; FERRARI, P.; FILIPPI, M.

2006-01-01

The Physics Department of Trento University (Italy), in collaboration with Bruker Biospin decided to set up a 4 T MR system for research in the neurological field as well as on the biological effects of high field MRI. Nowadays the number of similar apparatus set up in the world is very low, several of these operate in the USA, some in Australia and one in Israel. Due to the very low diffusion of these equipments and to the few knowledge today available about the biological effects by clinical application of high field MRI, the problems related to the installation of the apparatus from the safety point of view are not well standardized. Therefore, in Trento, it has been necessary to deal with many problems regarding the set up of the MRI apparatus, taking also into account the current law in Italy with respect the safety and protection in MRI applications. In the present paper the Authors show the carried out studies and analyses, in relation to the organization of the safety aspects. These studies have resulted in arranging specific guidelines for: working in controlled way; emergency procedures for the patient advanced life support; procedures for the protocol of patient protection; rules for the safety of the operators, patients and general public. The carried out analyses have mainly taken into account the danger identification and the risk evaluation. For this reason it was set up a test protocol regarding:the performing tests, the very detailed test modalities for their periodic performance and the test sequence. For the patient safety and protection it has been necessary to put care, in the first step of the survey, to the subjects in critical situations, due to: working activity, clinical history, health conditions of the patient at the moment of the examination. So it has been necessary to set up an anamnesis questionnaire for the patient, finalized to spot: (1) the patients with particular contraindications for the MR examination; (2) the patients which

High field MRI: safety and protection problems related to the installation and the management of a 4T equipment

Energy Technology Data Exchange (ETDEWEB)

Russo, A.; Speranza, A. [National Research Council, Institute of Bioimaging and Biostructures- Dep.of Biomorphological and Functional Sciences - University Federico II, Napoli (Italy); DELIA, R. [University of Rome - Medicine and Surgery Faculty - La Sapienza Sez. Rieti (Italy); ANTOLINI, R. [University of Trento - Department of Physics, Trento (Italy); FERRARI, P. [University of Trento - Direction of Polo di Rovereto, Trento (Italy); FILIPPI, M. [University of Trento Prevention and Protection Service, Trento (Italy)

2006-07-01

The Physics Department of Trento University (Italy), in collaboration with Bruker Biospin decided to set up a 4 T MR system for research in the neurological field as well as on the biological effects of high field MRI. Nowadays the number of similar apparatus set up in the world is very low, several of these operate in the USA, some in Australia and one in Israel. Due to the very low diffusion of these equipments and to the few knowledge today available about the biological effects by clinical application of high field MRI, the problems related to the installation of the apparatus from the safety point of view are not well standardized. Therefore, in Trento, it has been necessary to deal with many problems regarding the set up of the MRI apparatus, taking also into account the current law in Italy with respect the safety and protection in MRI applications. In the present paper the Authors show the carried out studies and analyses, in relation to the organization of the safety aspects. These studies have resulted in arranging specific guidelines for: working in controlled way; emergency procedures for the patient advanced life support; procedures for the protocol of patient protection; rules for the safety of the operators, patients and general public. The carried out analyses have mainly taken into account the danger identification and the risk evaluation. For this reason it was set up a test protocol regarding:the performing tests, the very detailed test modalities for their periodic performance and the test sequence. For the patient safety and protection it has been necessary to put care, in the first step of the survey, to the subjects in critical situations, due to: working activity, clinical history, health conditions of the patient at the moment of the examination. So it has been necessary to set up an anamnesis questionnaire for the patient, finalized to spot: (1) the patients with particular contraindications for the MR examination; (2) the patients which

Method of operator safety assessment for underground mobile mining equipment

Science.gov (United States)

Działak, Paulina; Karliński, Jacek; Rusiński, Eugeniusz

2018-01-01

The paper presents a method of assessing the safety of operators of mobile mining equipment (MME), which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM). As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

Method of operator safety assessment for underground mobile mining equipment

Directory of Open Access Journals (Sweden)

Działak Paulina

2018-01-01

Full Text Available The paper presents a method of assessing the safety of operators of mobile mining equipment (MME, which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM. As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

The application of redundancy-related basic safety principles to the 1400 MWE reactor core standby cooling system

International Nuclear Information System (INIS)

Bertrand, R.

1990-01-01

This memorandum shall provide the background for the work of the European Community Commission which is to analyze safety principles relating to redundancy. The redundancy-related basic safety principles applied in French nuclear power plants are the following: . the single-failure criterion, . provisions additional to application of the single-failure criterion. These are mainly provisions made at the design stage to minimize risks associated with common cause failures or the risks of human error which can lead to such failures: - protection against hazards of internal and external origin, - the geographical or physical separation of equipment, - the independence of electrical power supplies and distribution systems, - the additional resources and associated operating procedures making it possible to accommodate total loss of the safety systems. The scope also includes the operating rules which ensure availability of redundant safety-related equipment. The provisions relating to the single-failure criterion are detailed in Basic Safety Rule 1.3.A appended. The application of these principles proposed by the operating organization and accepted by the safety authorities for the design and operation of the standby core cooling system (System RIS) is explained

Benefits of Digital Equipment Generic Qualification Activities

International Nuclear Information System (INIS)

Thomas, James E.; Steiman, Samuel C.

2002-01-01

As a result of nuclear power plant instrumentation and control obsolescence issues, there have been numerous activities during recent years relating to the qualification of digital equipment. Some of these activities have been 'generic' in nature in that the qualification was not limited to plant specific applications, but was intended to cover a broad base of potential applications of the digital equipment. These generic qualifications have been funded by equipment manufacturers and by utility groups and organizations. The generic activities sponsored by the Electric Power Research Institute (EPRI) have been pilot projects for an overall generic qualification approach. The primary benefit resulting from the generic qualification work to date is that a number of digital platforms and digital devices are now available for use in various nuclear safety-related applications. Many of the tests and evaluations necessary to support plant specific applications have been completed. The amount of data and documentation that each utility must develop on a case by case basis has been significantly reduced. There are also a number of additional benefits resulting from these industry efforts. The challenges and difficulties in qualifying digital equipment for safety-related applications are now more clearly understood. EPRI has published a lessons learned document (EPRI Report 1001452, Generic Qualification of Commercial Grade Digital Devices: Lessons Learned from Initial Pilots, which covers several different qualification areas, including device selection, project planning, vendor surveys and design reviews, and electromagnetic compatibility (EMC) qualification. Application of the experience and lessons learned from the EPRI pilot activities should help reduce the effort and cost required for future qualification work. Most generic qualification activities for commercial equipment have been conducted using the approach of EPRI TR-106439, Guideline on Evaluation and Acceptance

Safety plan for the cooperative telerobotic retrieval system equipment development area

Energy Technology Data Exchange (ETDEWEB)

Haney, T.J.; Jessmore, J.J.

1995-07-01

This plan establishes guidelines to minimize safety risks for the cooperative telerobotic retrieval project at the North Boulevard Annex (NBA). This plan has the dual purpose of minimizing safety risks to workers and visitors and of securing sensitive equipment from inadvertent damage by nonqualified personnel. This goal will be accomplished through physical control of work zones and through assigned responsibilities for project personnel. The scope of this plan is limited to establishing the working zone boundaries and entry requirements, and assigning responsibilities for project personnel. This plan does not supersede current safety organization responsibilities for the Landfill Stabilization Focus Area Transuranic (LSFA TRU) Arid outlined in the Environment, Safety, Health, and Quality Plan for the Buried Waste Integrated Demonstration Program; Tenant Manual; Idaho Falls Building Emergency Control Plan;; applicable Company Procedures; the attached Interface Agreement (Appendix A).

Safety plan for the cooperative telerobotic retrieval system equipment development area

International Nuclear Information System (INIS)

Haney, T.J.; Jessmore, J.J.

1995-07-01

This plan establishes guidelines to minimize safety risks for the cooperative telerobotic retrieval project at the North Boulevard Annex (NBA). This plan has the dual purpose of minimizing safety risks to workers and visitors and of securing sensitive equipment from inadvertent damage by nonqualified personnel. This goal will be accomplished through physical control of work zones and through assigned responsibilities for project personnel. The scope of this plan is limited to establishing the working zone boundaries and entry requirements, and assigning responsibilities for project personnel. This plan does not supersede current safety organization responsibilities for the Landfill Stabilization Focus Area Transuranic (LSFA TRU) Arid outlined in the Environment, Safety, Health, and Quality Plan for the Buried Waste Integrated Demonstration Program; Tenant Manual; Idaho Falls Building Emergency Control Plan;; applicable Company Procedures; the attached Interface Agreement (Appendix A)

Contributions to the research programs in nuclear and industrial electronics, domestic production of instrumentation, safety and control systems and equipment for nuclear reactors and auxiliary installations

International Nuclear Information System (INIS)

Talpariu, C; Talpariu, J.; Matei, C.

2001-01-01

Domestic production of component system and equipment for the control and safety of nuclear facilities was one of the priority objective of the Nuclear Research Institute Pitesti. The problems addressed were particularly related to design and production of analog and digital equipment for measurements, triggering and display of the values of process parameters as well as to regulating complex functions of this equipment. Associated to this effort were the research works concerning: - reliability and in-service life-time of the electronic components and equipment in the safety and control systems for nuclear processes; - radiation endurance of industrial electronic components; utilization of whirling currents in calandria tube testing; - expert systems and applications in nuclear reactor control and safety; design and testing methods of process real time software packages for safety in control critical systems for nuclear domain. There are presented characteristics of the following equipment: 1. amplifier for ionization chambers with triggering comparator circuits for the CANDU 600 reactor shut down system; 2. amplifier for ionization chambers without triggering comparator circuits for power regulating system; 3. safety and regulating computerized system for C9 and C5 cans; 4. acquisition system for dosimetric data in nuclear facilities; 5. program able digital comparator for the reactor shut down system; 6. stationary gamma areal monitors for CANDU 600 reactors and other nuclear facilities

29 CFR 1915.136 - Internal combustion engines, other than ship's equipment.

Science.gov (United States)

2010-07-01

... 29 Labor 7 2010-07-01 2010-07-01 false Internal combustion engines, other than ship's equipment. 1915.136 Section 1915.136 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR (CONTINUED) OCCUPATIONAL SAFETY AND HEALTH STANDARDS FOR SHIPYARD EMPLOYMENT Tools and Related Equipment §...

Chemistry laboratory safety manual available

Science.gov (United States)

Elsbrock, R. G.

1968-01-01

Chemistry laboratory safety manual outlines safe practices for handling hazardous chemicals and chemistry laboratory equipment. Included are discussions of chemical hazards relating to fire, health, explosion, safety equipment and procedures for certain laboratory techniques and manipulations involving glassware, vacuum equipment, acids, bases, and volatile solvents.

Considerations concerning the reliability of reactor safety equipment

International Nuclear Information System (INIS)

Furet, J.; Guyot, Ch.

1967-01-01

A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [fr

Regulatory analysis for resolution of Unresolved Safety Issue A-46, seismic qualification of equipment in operating plants

International Nuclear Information System (INIS)

Chang, T.Y.; Anderson, N.R.

1987-02-01

The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform required safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants must be reassessed to determine whether requalification is necessary. The objective of technical studies performed under Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring these plants to meet the criteria that are applied to new plants. This report presents the regulatory analysis for Unresolved Safety Issue (USI) A-46. It includes: Statement of the Problem; the Objective of USI A-46; a Summary of A-46 Tasks; a Proposed Implementation Procedure; a Value-Impact Analysis; Application of the Backfit Rule; 10 CFR 50.109; Implementation; and Operating Plants To Be Reviewed to USI A-46 Requirements

Study on a quantitative evaluation method of equipment maintenance level and plant safety level for giant complex plant system

International Nuclear Information System (INIS)

Aoki, Takayuki

2010-01-01

In this study, a quantitative method on maintenance level which is determined by the two factors, maintenance plan and field work implementation ability by maintenance crew is discussed. And also a quantitative evaluation method on safety level for giant complex plant system is discussed. As a result of consideration, the following results were obtained. (1) It was considered that equipment condition after maintenance work was determined by the two factors, maintenance plan and field work implementation ability possessed by maintenance crew. The equipment condition determined by the two factors was named as 'equipment maintenance level' and its quantitative evaluation method was clarified. (2) It was considered that CDF in a nuclear power plant, evaluated by using a failure rate counting the above maintenance level was quite different from CDF evaluated by using existing failure rates including a safety margin. Then, the former CDF was named as 'plant safety level' of plant system and its quantitative evaluation method was clarified. (3) Enhancing equipment maintenance level means an improvement of maintenance quality. That results in the enhancement of plant safety level. Therefore, plant safety level should be always watched as a plant performance indicator. (author)

Safety shutdowns and failures of the RA reactor equipment; Sigurnosna zaustavljanja i kvarovi opreme na reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Mitrovic, S [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

1966-07-01

This report is an attempt of statistical analysis of the failures occurred during RA reactor operation. A list of failures occurred on the RA equipment during 1965 is included. Failures were related to the following systems: dosimetry system (22%), safety and control system (7%), heavy water system (2%), technical water (4%), helium system (2%), measuring instruments (30%), transport, ventilation, power supply systems (32%). A review of safety shutdowns from 1962 to 1966 is included as well, as a comparison with three similar reactors. Although the number of events used for statistical analysis was not adequate, it has been concluded that RA reactor operation was stable and reliable.

Technical diagnostics - equipment monitoring for increasing safety and availability of nuclear power plants

International Nuclear Information System (INIS)

Sturm, A.; Foerster, R.

1977-01-01

Utilization of technical diagnostics in equipment monitoring of nuclear power plants for ensuring nuclear safety, economic availability, and for decision making on necessary maintenance is reviewed. Technical diagnostics is subdivided into inspection and early detection of malfunctions. Moreover, combination of technical diagnostics and equipment monitoring, integration of technical diagnostics into maintenance strategy, and problems of introducing early detection of malfunctions into maintenance management of nuclear power plants are also discussed. In addition, a compilation of measuring techniques used in technical diagnostics has been made. The international state of the art of equipment monitoring in PWR nuclear power plants is illustrated by description of the sound and vibration measuring techniques. (author)

Study concerning the power plant control and safety equipment by integrated distributed systems

International Nuclear Information System (INIS)

Optea, I.; Oprea, M.; Stanescu, P.

1995-01-01

The paper deals with the trends existing in the field of nuclear control and safety equipment and systems, proposing a high-efficiency integrated system. In order to enhance the safety of the plant and reliability of the structure system and components, we present a concept based on the latest computer technology with an open, distributed system, connected by a local area network with high redundancy. A modern conception for the control and safety system is to integrate all the information related to the reactor protection, active engineered safeguard and auxiliary systems parameters, offering a fast flow of information between all the agencies concerned so that situations can be quickly assessed. The integrated distributed control is based on a high performance operating system for realtime applications, flexible enough for transparent networking and modular for demanding configurations. The general design considerations for nuclear reactors instrumentation reliability and testing methods for real-time functions under dynamic regime are presented. Taking into account the fast progress in information technology, we consider the replacement of the old instrumentation of Cernavoda-1 NPP by a modern integrated system as an economical and efficient solution for the next units. (Author) 20 Refs

A Study on Performance and Safety Tests of Electrosurgical Equipment.

Science.gov (United States)

Tavakoli Golpaygani, A; Movahedi, M M; Reza, M

2016-09-01

Modern medicine employs a wide variety of instruments with different physiological effects and measurements. Periodic verifications are routinely used in legal metrology for industrial measuring instruments. The correct operation of electrosurgical generators is essential to ensure patient's safety and management of the risks associated with the use of high and low frequency electrical currents on human body. The metrological reliability of 20 electrosurgical equipment in six hospitals (3 private and 3 public) was evaluated in one of the provinces of Iran according to international and national standards. The achieved results show that HF leakage current of ground-referenced generators are more than isolated generators and the power analysis of only eight units delivered acceptable output values and the precision in the output power measurements was low. Results indicate a need for new and severe regulations on periodic performance verifications and medical equipment quality control program especially in high risk instruments. It is also necessary to provide training courses for operating staff in the field of meterology in medicine to be acquianted with critical parameters to get accuracy results with operation room equipment.

29 CFR 1918.106 - Payment for protective equipment.

Science.gov (United States)

2010-07-01

... pay for non-specialty safety-toe protective footwear (including steel-toe shoes or steel-toe boots... Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR (CONTINUED) SAFETY AND HEALTH REGULATIONS FOR LONGSHORING Personal Protective Equipment...

Equipment aging: an overview of status and research needs

International Nuclear Information System (INIS)

Carfagno, S.P.

1982-01-01

The problem of equipment aging in nuclear power generating stations is viewed from the standpoint of the qualification of safety-related equipment. A brief summary of developments related to equipment aging and qualification (in industry, the regulatory arena, and in the research area) is presented to focus on the current status of the problem. Several research topics are suggested for consideration as components of any general plan of research on plant aging. Among these is a topic concerned with an effort to establish the relative merit of different approaches to account for the functional degradation of equipment in terms of their cost and their contribution to assurance of public safety and a topic to investigate ways to improve the monitoring of functional degradation of equipment in service. Some factors to be considered in choosing among research topics and in establishing funding levels are mentioned

The method for assurance to operational safety of cryogenics equipment through definite maintenance

International Nuclear Information System (INIS)

Anghel, Vasile

2004-01-01

The present paper reports about a method of ensuring operational safety of cryogenics equipment through specific maintenance measures which imply a definite concept of maintainability and operational reliability, methods of maintenance, necessary logistic and necessary personnel. The stages of the concept should succeed in a logical order and must be based on detailed analyses of the process of maintenance. The method starts from the operation conditions of cryogenic equipment in nuclear regime. Thus maintenance policy must be well defined within the boundaries of the nuclear plant through definition of coherent objectives and responsibility delegation for the personnel implied in maintenance activity. This method ensures performance in maintenance and an efficient feedback from equipment analysis. (author)

The establish and application of equipment reliability database in Nuclear Power Plant

International Nuclear Information System (INIS)

Zheng Wei; Li He

2006-03-01

Take the case of Daya Bay Nuclear Power Plant, the collecting and handling of equipment reliability data, the calculation method of reliability parameters and the establish and application of reliability databases, etc. are discussed. The data source involved the design information of the equipment, the operation information, the maintenance information and periodically test record, etc. Equipment reliability database built on a base of the operation experience. It provided the valid tool for thoroughly and objectively recording the operation history and the present condition of various equipment of the plant; supervising the appearance of the equipment, especially the safety-related equipment, provided the very practical worth information for enhancing the safety and availability management of the equipment and insuring the safety and economic operation of the plant; and provided the essential data for the research and applications in safety management, reliability analysis, probabilistic safety assessment, reliability centered maintenance and economic management in nuclear power plant. (authors)

Co-operation Equipment Qualification for Safety grade I and C Equipment between Russia and Korea

International Nuclear Information System (INIS)

Ham, J. K.; Lee, D. H.; Park, J. W.; Jeong, M. H.; Choi, Y. H.; Suh, J. J.; Jang, T. H.

2007-06-01

Equipment Qualification technology for I and C system on nuclear grade have been approved with accurate and safe for verifying and validation. It contributes to acknowledge of reliability and to review of user approval. Also it did not completely satisfied the requirement of Russia, it can be satisfied by bit of modification on design. It is expected to support the export of safety PLC to nuclear power plant. Proven EQ technology is required on the further markets of nuclear. Therefore, comparison test with foreign country is necessary to support the mutual accreditation of EQ for export of domestic products

Co-operation Equipment Qualification for Safety grade I and C Equipment between Russia and Korea

Energy Technology Data Exchange (ETDEWEB)

Ham, J. K.; Lee, D. H.; Park, J. W.; Jeong, M. H.; Choi, Y. H.; Suh, J. J.; Jang, T. H. [Korea Testing Laboratory, Seoul (Korea, Republic of)

2007-06-15

Equipment Qualification technology for I and C system on nuclear grade have been approved with accurate and safe for verifying and validation. It contributes to acknowledge of reliability and to review of user approval. Also it did not completely satisfied the requirement of Russia, it can be satisfied by bit of modification on design. It is expected to support the export of safety PLC to nuclear power plant. Proven EQ technology is required on the further markets of nuclear. Therefore, comparison test with foreign country is necessary to support the mutual accreditation of EQ for export of domestic products.

Development of methodologies for optimization of surveillance testing and maintenance of safety related equipment at NPPs. Report of a research coordination meeting. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

This report summarizes the results of the first meeting of the Coordinated Research Programme (CRP) on Development of Methodologies for Optimization of Surveillance Testing and Maintenance of Safety Related Equipment at NPPs, held at the Agency Headquarters in Vienna, from 16 to 20 December 1996. The purpose of this Research Coordination Meeting (RCM) was that all Chief Scientific Investigators of the groups participating in the CRP presented an outline of their proposed research projects. Additionally, the participants discussed the objective, scope, work plan and information channels of the CRP in detail. Based on these presentations and discussions, the entire project plan was updated, completed and included in this report. This report represents a common agreed project work plan for the CRP. Refs, figs, tabs.

Commercial grade item (CGI) dedication of generators for nuclear safety related applications

International Nuclear Information System (INIS)

Das, R.K.; Hajos, L.G.

1993-01-01

The number of nuclear safety related equipment suppliers and the availability of spare and replacement parts designed specifically for nuclear safety related application are shrinking rapidly. These have made it necessary for utilities to apply commercial grade spare and replacement parts in nuclear safety related applications after implementing proper acceptance and dedication process to verify that such items conform with the requirements of their use in nuclear safety related application. The general guidelines for the commercial grade item (CGI) acceptance and dedication are provided in US Nuclear Regulatory Commission (NRC) Generic Letters and Electric Power Research Institute (EPRI) Report NP-5652, Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications. This paper presents an application of these generic guidelines for procurement, acceptance, and dedication of a commercial grade generator for use as a standby generator at Salem Generating Station Units 1 and 2. The paper identifies the critical characteristics of the generator which once verified, will provide reasonable assurance that the generator will perform its intended safety function. The paper also delineates the method of verification of the critical characteristics through tests and provide acceptance criteria for the test results. The methodology presented in this paper may be used as specific guidelines for reliable and cost effective procurement and dedication of commercial grade generators for use as standby generators at nuclear power plants

A Study on Performance and Safety Tests of Electrosurgical Equipment

Directory of Open Access Journals (Sweden)

Tavakoli Golpaygani A.

2016-09-01

Full Text Available Introduction: Modern medicine employs a wide variety of instruments with different physiological effects and measurements. Periodic verifications are routinely used in legal metrology for industrial measuring instruments. The correct operation of electrosurgical generators is essential to ensure patient’s safety and management of the risks associated with the use of high and low frequency electrical currents on human body. Material and Methods: The metrological reliability of 20 electrosurgical equipment in six hospitals (3 private and 3 public was evaluated in one of the provinces of Iran according to international and national standards. Results: The achieved results show that HF leakage current of ground-referenced generators are more than isolated generators and the power analysis of only eight units delivered acceptable output values and the precision in the output power measurements was low. Conclusion: Results indicate a need for new and severe regulations on periodic performance verifications and medical equipment quality control program especially in high risk instruments. It is also necessary to provide training courses for operating staff in the field of meterology in medicine to be acquianted with critical parameters to get accuracy results with operation room equipment.

Managing EHS of PV-Related Equipment at the National Renewable Energy Laboratory: Preprint

Energy Technology Data Exchange (ETDEWEB)

McCuskey, T.; Nelson, B. P.

2012-06-01

Managing environment, health, and safety (EHS) risks at a national laboratory, or university, can be intimidating to a researcher who is focused on research results. Laboratory research and development (R&D) operations are often driven by scientists with limited engineering support and lack well-refined equipment development resources. To add to the burden for a researcher, there is a plethora of codes, standards, and regulations that govern the safe installation and operation of photovoltaic-related R&D equipment -- especially those involving hazardous production materials. To help guide the researcher through the vast list of requirements, the EHS office at NREL has taken a variety of steps. Organizationally, the office has developed hazard-specific laboratory-level procedures to govern particular activities. These procedures are a distillation of appropriate international codes, fire agencies, SEMI standards, U.S. Department of Energy orders, and other industry standards to those necessary and sufficient to govern the safe operation of a given activity. The EHS office works proactively with researchers after a concept for a new R&D capability is conceived to help guide the safe design, acquisition, installation, and operation of the equipment. It starts with a safety assessment at the early stages such that requirements are implemented to determine the level of risk and degree of complexity presented by the activity so appropriate controls can be put in place to manage the risk. As the equipment requirements and design are refined, appropriate equipment standards are applied. Before the 'to-build' specifications are finalized, a process hazard analysis is performed to ensure that no single-point failure presents an unacceptable risk. Finally, as the tool goes through construction and installation stages, reviews are performed at logical times to ensure that the requisite engineering controls and design are in place and operational. Authorization to

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.

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

HAPO Plant and capital equipment budget for FY 1966 and revision of budget for FY 1965 equipment not related to construction projects

Energy Technology Data Exchange (ETDEWEB)

McDonald, J.E.

1964-05-12

This document is divided into: byproduct horizontal control rod system (5 reactors); high-speed scanning-effluent temperature monitoring system (KE reactor); improved reactor gas system (100-KE & KW); safety circuit system modifications (5 reactors); alternate process hot die sizing, 313 Building 300 Area; button line equipment (234-5 Building); in-tank waste solidification (3rd unit); and misc. minor equipment projects.

Official News relating to CERN Safety Rules

CERN Multimedia

HSE Unit

2015-01-01

The CERN Safety Rules listed below have been published on the HSE website (see here) and entered into force on the 9 June 2015:   Safety Regulation SR-M “Mechanical equipment”: http://cern.ch/safety-rules/SR-M_ENv2.htm; this SR-M (version 2) cancels and replaces SR-M (version 1) and the corresponding provisions of General Safety Instruction GSI-M3 “Special Equipment” (version 1).   General Safety Instruction GSI-M-1 “Lifting equipment and accessories”: http://cern.ch/safety-rules/GSI-M-1_ENv2.htm; this GSI-M-1 (version 2) cancels and replaces GSI-M1 (version 1). Specific Safety Instruction SSI-M-1-1 “Slings and lifting chains”: http://cern.ch/safety-rules/SSI-M-1-1_EN.htm; Specific Safety Instruction SSI-M-1-2 “Cranes, bridge cranes, gantry cranes and power-driven hoists”: http://cern.ch/safety-rules/SSI-M-1-2_EN.htm; Specific Safety Instruction SSI-M-1-3 “Non-f...

Development of nuclear equipment qualification technology

Energy Technology Data Exchange (ETDEWEB)

Choi, Heon O; Kim, Wu Hyun; Kim, Jin Wuk; Kim, Jeong Hyun; Lee, Jeong Kyu; Kim, Yong Han; Jeong, Hang Keun [Korea Institute of Machinery and Materials, Taejon (Korea)

1999-03-01

In order to enhance testing and evaluation technologies, which is one of the main works of the Chanwon branch of KIMM(Korea Institute of Machinery and Materials), in addition to the present work scope of the testing and evaluation in the industrial facilities such as petroleum and chemical, plants, the qualification technologies of the equipments important to safety used in the key industrial facilities such as nuclear power plants should be localized: Equipments for testing and evaluation is to be set up and the related technologies must be developed. In the first year of this study, of vibration aging qualification technologies of equipments important to safety used in nuclear power plants have been performed. (author). 27 refs., 81 figs., 17 tabs.

Design and installation of advanced computer safety related instrumentation

International Nuclear Information System (INIS)

Koch, S.; Andolina, K.; Ruether, J.

1993-01-01

The rapidly developing area of computer systems creates new opportunities for commercial utilities operating nuclear reactors to improve plant operation and efficiency. Two of the main obstacles to utilizing the new technology in safety-related applications is the current policy of the licensing agencies and the fear of decision making managers to introduce new technologies. Once these obstacles are overcome, advanced diagnostic systems, CRT-based displays, and advanced communication channels can improve plant operation considerably. The article discusses outstanding issues in the area of designing, qualifying, and licensing of computer-based instrumentation and control systems. The authors describe the experience gained in designing three safety-related systems, that include a Programmable Logic Controller (PLC) based Safeguard Load Sequencer for NSP Prairie Island, a digital Containment Isolation monitoring system for TVA Browns Ferry, and a study that was conducted for EPRI/NSP regarding a PLC-based Reactor Protection system. This article presents the benefits to be gained in replacing existing, outdated equipment with new advanced instrumentation

Safety Analysis of 'Older/Aged' Handling and Transportation Equipment for Heavy Loads, Radioactive Waste and Materials in Accordance with German Nuclear Standards KTA 3902, 3903 and 3905

International Nuclear Information System (INIS)

Macias, P.; Prucker, E.; Stang, W.

2006-01-01

The purpose of this paper is to present a general safety analysis of important handling and transportation processes and their related equipment ('load chains' consisting of cranes, load-bearing equipment and load-attaching points). This project was arranged by the responsible Bavarian ministry for environment, health and consumer protection (StMUGV) in agreement with the power plant operators of all Bavarian nuclear power plants to work out potential safety improvements. The range of the equipment (e.g. reactor building, crane, refuelling machine, load-bearing equipment and load-attaching points) covers the handling and transportation of fuel elements (e. g. with fuel flasks), heavy loads (e.g. reactor pressure vessel closure head, shielding slabs) and radioactive materials and waste (e.g. waste flasks, control elements, fuel channels, structure elements). The handling equipment was subjected to a general safety analysis taking into account the ageing of the equipment and the progress of standards. Compliance with the current valid requirements of the state of science and technology as required by German Atomic Act and particularly of the nuclear safety KTA-standards (3902, 3903 and 3905) was examined. The higher protection aims 'safe handling and transportation of heavy loads and safe handling of radioactive materials and waste' of the whole analysis are to avoid a criticality accident, the release of radioactivity and inadmissible effects on important technical equipment and buildings. The scope of the analysis was to check whether these protection aims were fulfilled for all important technical handling and transportation processes. In particularly the design and manufacturing of the components and the regulations of the handling itself were examined. (authors)

On safety 1E qualification of electrical equipment for nuclear power stations

International Nuclear Information System (INIS)

Segarceanu, D.; Geambasu, C.; Avramescu, M.

1995-01-01

Electrical equipment and systems for the emergency shutdown of a nuclear reactor are qualified according to safety class 1E. Methods of qualification meeting the requirements should be used, either individually or in combination include, type-test qualification, qualification by operating experience, qualification by analysis, combined qualification. These methods qualification principles, procedures and documents are discussed. (N.T.). 1 fig

Quality Control Activities Related to Mechanical Maintenance of Safety Related Components at Krsko NPP

International Nuclear Information System (INIS)

Djakovic, D.

2016-01-01

performing receipt control of complete units had found many deficiencies, excessive distortions, heat tints and corrosion spots. Heat tints could cause corrosion, which could accelerate in a harsh environment conditions. The reason for the faults identified in both cases was improper human performance. These findings confirm the need to control human performance factor in addition to standard quality control activities. Best tools to be used in such cases beside QC inspections are QNOD observations. The third specific maintenance case was a replacement of pressurizer spray control valve during the last outage. This was a complex activity on Safety Class 1 equipment. The QC department had to perform welding control and non-destructive examination of welds in very demanding radiological conditions. This activity affected the plant ISI program and pre-service inspection of the welds had to be performed. Activities affecting quality shall be accomplished under suitably controlled conditions and well performed maintenance of safety related components followed by successful QC control are essential for nuclear safety. (author).

Seismic qualification of equipment in operating nuclear power plants: Unresolved Safety Issue A-46

International Nuclear Information System (INIS)

Chang, T.Y.

1987-02-01

The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants must be reassessed to determine whether requalification is necessary. The objective of technical studies performed under the Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring qualification to the current criteria that are applied to new plants. This report summarizes the work accomplished on USI A-46. In addition, the collection and review of seismic experience data and existing seismic test data are presented. Staff assessment of work accomplished under USI A-46 leads to the conclusion that the use of seismic experience data provides the most reasonable alternative to current qualification criteria. Consideration of seismic qualification by use of experience data was a specific task in USI A-46. Several other A-46 tasks serve to support the use of an experienced data base. The principal technical finding of USI A-46 is that seismic experience data, supplemented by existing seismic test data, applied in accordance with the guidelines developed, can be used to verify the seismic adequacy of mechanical and electrical equipment in operating nuclear plants. Explicit seismic qualification should be required only if seismic experience data or existing test data on similar components cannot be shown to apply

Coordinated research programme on safety of RBMK type NPPs in relation to external events. V. 1. Working material

International Nuclear Information System (INIS)

1999-01-01

The present volume is a collection of progress reports which have been submitted within the scope of the CRP on safety of RBMK type NPPs in relation to external events including seismic related papers and man-induced events (explosions and airplane crash). It includes papers concerned with experience related to RBMK equipment testing and calculations of seismic resistance, soil-structure interactions analysis, safety assurance, aircraft impact qualification and other external events for RBMK type NPP, seismic stability of NPPs in Eastern Europe, probabilistic assessment of NPP safety under aircraft impact, dynamic analysis of NPPs, screening of external hazards for NPP

Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category I systems for Palisades nuclear power plant

International Nuclear Information System (INIS)

Collins, E.K.

1979-10-01

The technical evaluation is presented of Consumers Power Company's Palisades nuclear power plant to determine whether the failure of any non-Category I (seismic) equipment could result in a condition, such as flooding, that might potentially adversely affect the performance of safety-related equipment required for the safe shutdown of the facility or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection as well as measures taken by Consumers Power Company to minimize the danger of flooding and to protect safety-related equipment

Industrial high pressure applications. Processes, equipment and safety

Energy Technology Data Exchange (ETDEWEB)

Eggers, Rudolf (ed.) [Technische Univ. Hamburg-Harburg, Hamburg (Germany). Inst. fuer Thermische Verfahrenstechnik

2012-07-01

Industrial high pressure processes open the door to many reactions that are not possible under 'normal' conditions. These are to be found in such different areas as polymerization, catalytic reactions, separations, oil and gas recovery, food processing, biocatalysis and more. The most famous high pressure process is the so-called Haber-Bosch process used for fertilizers and which was awarded a Nobel prize. Following an introduction on historical development, the current state, and future trends, this timely and comprehensive publication goes on to describe different industrial processes, including methanol and other catalytic syntheses, polymerization and renewable energy processes, before covering safety and equipment issues. With its excellent choice of industrial contributions, this handbook offers high quality information not found elsewhere, making it invaluable reading for a broad and interdisciplinary audience.

78 FR 16358 - Safety Advisory 2013-02; Low-Speed, Wheel-Climb Derailments of Passenger Equipment With “Stiff...

Science.gov (United States)

2013-03-14

..., Staff Director, Motive Power and Equipment Division, Office of Railroad Safety, FRA, 1200 New Jersey Avenue SE., Washington, DC 20590, telephone (202) 493-6322; or Anna Nassif Winkle, Trial Attorney, Office... studies conducted by the equipment manufacturer and FRA using computer modeling indicate that an uneven...

Emergency Diesel: Safety-related instrumentation and control with programmable logic controllers

International Nuclear Information System (INIS)

Breidenich, G.; Luedtke, M.

2004-01-01

This report presents a new concept for the design of emergency diesel equipment protection circuits as a part of the safety related instrumentation in the nuclear power plant Biblis, units A and B. The concept was implemented with state of the art SIMATIC S7/316 programmable logic controllers (PLCs) and can be adapted to any system with high availability requirements (e.g. power plant turbines, aircraft engines, mining pumps etc). (orig.)

Modified enforcement policy relating to 10 CFR 50.49, ''Environmental qualification of electrical equipment important to safety for nuclear power plants'' (Generic Letter 88-07)

International Nuclear Information System (INIS)

Miraglia, F.J.

1992-01-01

Generic Letters, Bulletins, and Information Notices have been issued to provide guidance regarding the application and enforcement of 10 CFR 50.49, ''Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants.'' Generic Letter 85-15, issued August 6, 1985, and Generic Letter 86-15, issued September 22, 1986, provided information related to the deadlines for compliance with 10 CFR 50.49 and possible civil penalties applicable to licensees who were not in compliance with the rule as of the November 30, 1985 deadline. Upon review, the Commission found that the EQ Enforcement Policy promulgated in Generic Letter 86-15, could result in imposition of civil penalties that did not properly reflect the safety significance of EQ violations with respect to civil penalties imposed in the past. In the interest of continuing a tough but fair enforcement policy, the Commission determined that the EQ Enforcement Policy should be revised. The purpose of this letter is to provide a modification to the NRC's enforcement policy, as approved by the Commission, for environmental qualification (EQ) violations. This letter replaces the guidance provided in Generic Letters 85-15 and 86-15

Machine-related injuries in the US mining industry and priorities for safety research.

Science.gov (United States)

Ruff, Todd; Coleman, Patrick; Martini, Laura

2011-03-01

Researchers at the National Institute for Occupational Safety and Health studied mining accidents that involved a worker entangled in, struck by, or in contact with machinery or equipment in motion. The motivation for this study came from the large number of severe accidents, i.e. accidents resulting in a fatality or permanent disability, that are occurring despite available interventions. Accident descriptions were taken from an accident database maintained by the United States Department of Labor, Mine Safety and Health Administration, and 562 accidents that occurred during 2000-2007 fit the search criteria. Machine-related accidents accounted for 41% of all severe accidents in the mining industry during this period. Machinery most often involved in these accidents included conveyors, rock bolting machines, milling machines and haulage equipment such as trucks and loaders. The most common activities associated with these accidents were operation of the machine and maintenance and repair. The current methods to safeguard workers near machinery include mechanical guarding around moving components, lockout/tagout of machine power during maintenance and backup alarms for mobile equipment. To decrease accidents further, researchers recommend additional efforts in the development of new control technologies, training materials and dissemination of information on best practices.

Safety of nursing staff and determinants of adherence to personal protective equipment

OpenAIRE

Neves,Heliny Carneiro Cunha; Souza,Adenícia Custódia Silva e; Medeiros,Marcelo; Munari,Denize Bouttelet; Ribeiro,Luana Cássia Miranda; Tipple,Anaclara Ferreira Veiga

2011-01-01

A qualitative study conducted in a teaching hospital with 15 nursing professionals. Attempted to analyze the reasons, attitudes and beliefs of nursing staff regarding adherence to personal protective equipment. Data were collected through focus groups, analyzed by the method of interpretation of meanings, considering Rosenstock’s model of health beliefs as a reference framework. Data revealed two themes: Occupational safety and Interpersonal Relationship. We identified several barriers that i...

Safety analysis report for packaging onsite long-length contaminated equipment transport system

International Nuclear Information System (INIS)

McCormick, W.A.

1997-01-01

This safety analysis report for packaging describes the components of the long-length contaminated equipment (LLCE) transport system (TS) and provides the analyses, evaluations, and associated operational controls necessary for the safe use of the LLCE TS on the Hanford Site. The LLCE TS will provide a standardized, comprehensive approach for the disposal of approximately 98% of LLCE scheduled to be removed from the 200 Area waste tanks

Safety analysis report for packaging, onsite, long-length contaminated equipment transport system

Energy Technology Data Exchange (ETDEWEB)

McCormick, W.A.

1997-05-09

This safety analysis report for packaging describes the components of the long-length contaminated equipment (LLCE) transport system (TS) and provides the analyses, evaluations, and associated operational controls necessary for the safe use of the LLCE TS on the Hanford Site. The LLCE TS will provide a standardized, comprehensive approach for the disposal of approximately 98% of LLCE scheduled to be removed from the 200 Area waste tanks.

Constructing a User Interface for Cellular Phones Using Equipment and its Relations

Directory of Open Access Journals (Sweden)

Misayo Kitamura

2003-04-01

Full Text Available In a domain of SCADA (Supervisory Control And Data Acquisition systems, it is necessary to obtain information about plants such as water plants in remote places using a cellular phone in order to ascertain plant status in case of emergency.T o utilize the small screen of a cellular phone and to eliminate the engineering cost of creating de.nition data to show plant status, a method of constructing user interface using equipment in the plant and its relations is proposed. In this method, some equipment is selected from all supervised equipment using the relations between the equipment, and then the content to be displayed is generated dynamically using the selected equipment. The equipment in plants is organized as a graph structure, which involves the equipment and the relations between the equipment.T he relations adopted in this method are both the physical connections between the equipment and the conceptual relationships.The result of the selection depends on the relations and their parameter values called the context dependent weight, which changes dynamically by viewpoints.

Quality and safety based management of radiological and clinical equipment

International Nuclear Information System (INIS)

Arora, S.; Saini, P.S.

2001-01-01

TQM (Total quality management) is the 'buzz' acronym expected in almost every aspect of industrial manufacturing, product development and services solicited by the outside agencies. The correct management of the many technological resources in health institutions in terms of efficiency, effectiveness and safety is a necessity in view of the criteria on which modern medicine is based. Almost all activities -be they diagnostic, therapeutic and rehabilitative - carried out within a modern hospital are based on widespread application of equipment and technology, the efficiency of which can directly influence the quality of service offered to the patient

Ergonomic requirements for the operation of machines and technical equipment

Directory of Open Access Journals (Sweden)

Górny Adam

2017-01-01

Full Text Available In order to operate machinery and equipment safely, it is critical for the solutions in place to conform to design-related and operating requirements. Design-related requirements are primarily the responsibility of machine designers/developers and manufacturers. Operating requirements should be satisfied by employers, who are responsible for ensuring safe working conditions for their employees. Under applicable laws, machinery and equipment should be designed, produced and then operated without placing excessive loads on workers and in keeping with machine functionality and intended use. One should also ensure that machinery and equipment can be maintained and adjusted without exposing their operators to hazards. Ergonomic criteria are an integral part of such requirements. They ensure that human users and operators of technical equipment are enabled to function properly. Design-related requirements are viewed as a priority safety consideration. While they facilitate the use of technical tools, the actual safety of employees ultimately depends on the satisfaction of specific requirements during operation.

Advancements in the design of safety-related systems and components of the MARS nuclear plant

International Nuclear Information System (INIS)

Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

1992-01-01

In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

Safety and security profiles of industry networks used in safety- critical applications

Directory of Open Access Journals (Sweden)

Mária FRANEKOVÁ

2008-01-01

Full Text Available The author describes the mechanisms of safety and security profiles of industry and communication networks used within safety – related applications in technological and information levels of process control recommended according to standards IEC 61784-3,4. Nowadays the number of vendors of the safety – related communication technologies who guarantees besides the standard communication, the communication amongst the safety – related equipment according to IEC 61508 is increasing. Also the number of safety – related products is increasing, e. g. safety Fieldbus, safety PLC, safety curtains, safety laser scanners, safety buttons, safety relays and other. According to world survey the safety Fieldbus denoted the highest growth from all manufactured safety products.The main part of this paper is the description of the safety-related Fieldbus communication system, which has to guaranty Safety Integrity Level.

Requirements for industrial x-ray equipment

International Nuclear Information System (INIS)

1987-01-01

This safety code is concerned with the protection of all individuals who may be exposed to radiation emitted by X-ray equipment operating at energies up to 1 MeV as used in industrial radiography. This code presents basic radiation safety information for the protection of personnel operating and servicing X-ray equipment and other workers and the general public in the vicinity of areas where X-ray equipment is in operation. It specifies general safety features of design, construction and functioning of X-ray equipment and facilities; describes the responsibilities of the user, operator and maintenance personnel; contains recommendations to ensure that the X-ray equipment is used and maintained in accordance with the ALARA principle; and describes a program of personnel monitoring and radiation safety surveys. ( 6 refs., 5 tabs., 4 figs.)

The application and practice of predictive maintenance at CANDU equipment management

International Nuclear Information System (INIS)

Yu Guangting

2014-01-01

The equipment in Qinshan CANDU unit is characterized by large number and complex structure. Some equipment failure has no relation with the operation time, it is impossible to avoid the failure of these equipment only by periodical maintenance. To improve the equipment reliability, increase the equipment usability and decrease the maintenance cost, for important equipment related to nuclear safety and generating electricity, it is required to perform condition monitoring and the predictive maintenance (PdM). According to different characteristics of equipment, it is required to use suitable equipment condition monitoring method, content and frequency. In this way, some potential equipment failure can be found, preventive maintenance can be arranged in advance, and equipment maintenance management can be optimized. (author)

The Design of Transportation Equipment in Terms of Human Capabilities. The Role of Engineering Psychology in Transport Safety.

Science.gov (United States)

McFarland, Ross A.

Human factors engineering is considered with regard to the design of safety factors for aviation and highway transportation equipment. Current trends and problem areas are identified for jet air transportation and for highway transportation. Suggested solutions to transportation safety problems are developed by applying the techniques of human…

Modification and backfitting in safety related systems at Ringhals 2

Energy Technology Data Exchange (ETDEWEB)

Lidh, B. [KSU, Nykoeping (Sweden); Stroemqvist, E. [ES-Konsult AB, Stockholm (Sweden)

1995-08-01

This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Ringhals-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 4 refs.

Modification and backfitting in safety related systems at Ringhals 2

International Nuclear Information System (INIS)

Lidh, B.; Stroemqvist, E.

1995-08-01

This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Ringhals-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 4 refs

Situation analysis of physical independence of the equipment and safety circuits of Almaraz NPP regarding R.G. 1.75 rev.3 (2005)

International Nuclear Information System (INIS)

Seijas Portela, S.

2010-01-01

Situation analysis of physical independence of the electrical equipment and circuits CN safety Almaraz about R.G. 1.75 rev. 3. (2005) The aim of this paper is to present the work done in the analysis of the physical separation of redundant safety electrical equipment (emergency diesel generators, medium voltage, electrical cabinets, etc.) and physical separation of circuits and electrical conduits.

Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

International Nuclear Information System (INIS)

Lydell, B.

1998-03-01

This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood

Safety equipment list for 241-C-106 waste retrieval, Project W-320: Revision 1

International Nuclear Information System (INIS)

Conner, J.C.

1994-01-01

The goals of the C-106 sluicing operation are: (1) to stabilize the tank by reducing the heat load in the tank to less than 42 MJ/hr (40,000 Btu/hour), and (2) to initiate demonstration of single-shell tank (SST) retrieval technology. The purpose of this supporting document (SD) is as follows: (1) to provide safety classifications for items (systems, structures, equipment, components, or parts) for the waste retrieval sluicing system (WRSS), and (2) to document and methodology used to develop safety classifications. Appropriate references are made with regard to use of existing systems, structures, equipments, components, and parts for C-106 single-shell transfer tank located in the C Tank Farm, and 241-AY-102 (AY-102) double shell receiver tanks (DST) located in the Aging Waste Facility (AWF). The Waste Retrieval Sluicing System consists of two transfer lines that would connect the two tanks, one to carry the sluiced waste slurry to AY-102, and the other to return the supernatant liquid to C-106. The supernatant, or alternate fluid, will be used to mobilize waste in C-106 for the sluicing process. The equipment necessary for the WRSS include pumps in each tank, sluicers to direct the supernatant stream in C-106, a slurry distributor in AY-102, HVAC for C-106, instrumentation and control devices, and other existing components as required

does road safety projects relate to community capacity building?

African Journals Online (AJOL)

USER

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

Safety equipment list for the light duty utility arm system

International Nuclear Information System (INIS)

Barnes, G.A.

1998-01-01

The initial issue (Revision 0) of this Safety Equipment List (SEL) for the Light Duty Utility Arm (LDUA) requires an explanation for both its existence and its being what it is. All LDUA documentation leading up to creation of this SEL, and the SEL itself, is predicated on the LDUA only being approved for use in waste tanks designated as Facility Group 3, i.e., it is not approved for use in Facility Group 1 or 2 waste tanks. Facility Group 3 tanks are those in which a spontaneous or induced hydrogen gas release would be small, localized, and would not exceed 25% of the LFL when mixed with the remaining air volume in the dome space; exceeding these parameters is considered unlikely. Thus, from a NFPA flammable gas environment perspective the waste tank interior is not classified as a hazardous location. Furthermore, a hazards identification and evaluation (HNF-SD-WM-HIE-010, REV 0) performed for the LDUA system concluded that the consequences of actual LDUA system postulated accidents in Flammable Gas Facility Group 3 waste tanks would have either NO IMPACT or LOW IMPACT on the offsite public and onsite worker. Therefore, from a flammable gas perspective, there is not a rationale for classifying any of SSCs associated with the LDUA as either Safety Class (SC) or Safety Significant (SS) SSCs, which, by default, categorizes them as General Service (GS) SSCs. It follows then, based on current PHMC procedures (HNF-PRO-704 and HNF-IP-0842, Vol IV, Section 5.2) for SEL creation and content, and from a flammable gas perspective, that an SEL is NOT REQ at sign D HOWEVER exclamation point exclamation point exclamation point There is both a precedent and a prudency to capture all SSCS, which although GS, contribute to a Defense-In-Depth (DID) approach to the design and use of equipment in potentially flammable gas environments. This Revision 0 of the LDUA SEL has been created to capture these SSCs and they are designated as GS-DID in this document. The specific reasons for

Undetected latent failures of safety-related systems. Preliminary survey of events in nuclear power plants 1980-1997

Energy Technology Data Exchange (ETDEWEB)

Lydell, B. [RSA Technologies, Vista, CA (United States)

1998-03-01

This report summarizes results and insights from a preliminary survey of events involving undetected, latent failures of safety-related systems. The survey was limited to events where mispositioned equipment (e.g., valves, switches) remained undetected, thus rendering standby equipment or systems unavailable for short or long time periods. Typically, these events were symptoms of underlying latent errors (e.g., design errors, procedure errors, unanalyzed safety conditions) and programmatic errors. The preliminary survey identified well over 300 events. Of these, 95 events are documented in this report. Events involving mispositioned equipment are commonplace. Most events are discovered soon after occurrence, however. But as evidenced by the survey results, some events remained undetected beyond several shift changes. The recommendations developed by the survey emphasize the importance of applying modern root cause analysis techniques to the event analysis to ensure that the causes and implications of occurred events are fully understood. 7 refs, 4 tabs, 3 figs. Also available at the SKI Home page: //www.ski.se.

Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-Category I systems for Turkey Point Nuclear Power Plant, Units 3 and 4

International Nuclear Information System (INIS)

Collins, E.K.

1979-08-01

Three separate reviews of the Turkey Point Units 3 and 4 were conducted by the FPLCO between 1972 and 1975. Initially, at the request of NBC in 1972, the FPLCO reviewed several water systems as sources of flooding. Subsequently, as a result of an abnormal occurrence, the drainage system was reviewed. Finally, the facilities were again reviewed at NRC's request and both the potential sources of flooding and safety-related equipment which could be damaged by flooding were identified. The sources of flooding and the appropriate safety equipment are discussed. An evaluation is presented of measures that were taken by FPLCO to minimize the danger of flooding and to protect safety-related equipment

Strategy for conformity of non-standard cryogenic equipment

CERN Multimedia

CERN. Geneva

2016-01-01

CERN as an intergovernmental organization establishes its own Safety Rules as necessary for its proper functioning. In particular, the CERN General Safety Instruction for cryogenic equipment requires that cryogenic pressure equipment at CERN shall comply with the European Pressure Equipment Directive (PED). However, due to the particular features of some of the cryogenic equipment required for the accelerators, as well as the existence of international collaborations with in-kind contributions from non-EU countries, full compliance with the PED may not always be achieved. This situation is foreseen in the Safety Rules, where CERN HSE will define the Safety requirements applicable to such equipment as well as any eventual additional compensatory measure as to ensure a commensurate level of Safety for our pressure equipment. Where compliance with PED may not be achieved, CERN HSE will become the de facto Notified Body and therefore be in charge of the assessment of the conformity of the equipment to the applica...

A Study of Cyber Security Activities for Development of Safety-related Controller

Energy Technology Data Exchange (ETDEWEB)

Lee, Myeongkyun; Song, Seunghwan; Yoo, Kwanwoo; Yun, Donghwa [Korea Univ., Seoul (Korea, Republic of)

2014-05-15

Nuclear Power Plant Regulatory guide describes the regulatory requirements to implement cyber security activities to ensure that design and operate to respond to cyber threats that exploited to vulnerability of digital-based technologies associated with safety-related digital instrumentation and control systems at nuclear power plants. Cyber security activities coverage is instrumentation and control systems to perform safety functions and digital-based equipment to use development, test, analysis and asset for instrumentation and control systems. Regulatory guidance is required to the cyber security activities that should be performed in each development phase of safety-related controller. Development organization should establish and implement to cyber security plans for responding to cyber threats throughout each lifecycle phase and the result of the cyber security activities should be generated to the documents. In addition, the independent verification and validation organization should perform simulated penetration test for enhancing response capabilities to cyber security threats and development organization should establish and implement response hardening solutions for the cyber security vulnerabilities identified in the simulated penetration test.

A Study of Cyber Security Activities for Development of Safety-related Controller

International Nuclear Information System (INIS)

Lee, Myeongkyun; Song, Seunghwan; Yoo, Kwanwoo; Yun, Donghwa

2014-01-01

Nuclear Power Plant Regulatory guide describes the regulatory requirements to implement cyber security activities to ensure that design and operate to respond to cyber threats that exploited to vulnerability of digital-based technologies associated with safety-related digital instrumentation and control systems at nuclear power plants. Cyber security activities coverage is instrumentation and control systems to perform safety functions and digital-based equipment to use development, test, analysis and asset for instrumentation and control systems. Regulatory guidance is required to the cyber security activities that should be performed in each development phase of safety-related controller. Development organization should establish and implement to cyber security plans for responding to cyber threats throughout each lifecycle phase and the result of the cyber security activities should be generated to the documents. In addition, the independent verification and validation organization should perform simulated penetration test for enhancing response capabilities to cyber security threats and development organization should establish and implement response hardening solutions for the cyber security vulnerabilities identified in the simulated penetration test

Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category 1 systems for the Yankee Rowe Nuclear Power Station

International Nuclear Information System (INIS)

Epps, R.C.

1980-11-01

This report documents the technical evaluation of the Maine Yankee Atomic Power Station. The purpose of this evaluation was to determine whether the failure of any non-Class I (seismic) equipment could result in a condition, such as flooding, that might adversely affect the performance of the safety-related equipment required for the safe shutdown of the facility, or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection system as well as measures taken by Maine Yankee Atomic Power Company (MYAPC) to minimize the danger of flooding and to protect safety-related equipment

Equipment Qualification and Environment Establishment for COTS Dedication of Safety Grade PC

International Nuclear Information System (INIS)

Ahn, Kwang Chul; Bin, Chang Sun; Kwon, Yoon Kwang; Kang, Shin Woo; Koh, Sung Won; Shon, Eui Chan; Jang, Hyun Doo; Song, Il Sup; Lee, Hyun Noh

2010-08-01

Safety grade PCs are required for protection systems of SKN 3 and 4 nuclear power plant and subsequent plants. Commercial grade item (CGI) dedication should be carried out to utilize a commercial PC as a safety grade PC of nuclear power plants. This project is aimed to perform the equipment qualification of the commercial PC, and review and improve the quality system of the PC supplier. As a result of the EQ for the CGI dedication, selected military PCs have passed the environment test, the seismic test, and the EMI test required for the digital controllers of nuclear plants. In addition, a walk-through of the quality system of the PC supplier, ISO 9001, was carried out and the quality system was improved. History data for the PC was gathered. As the analysis of the history data showed that operating experience time of the PC is longer than the plant life time, the history data could be used as an evidence of acceptance. After dedicated according to the CGI dedication process, military rugged PCs could be used for safety grade PCs

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

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1998-02-01

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

Qualification of class 1e equipment: regulation, technological margins and test experience

International Nuclear Information System (INIS)

Pasco, Y.; Le Meur, M.; Henry, J.Y.; Droger, J.P.; Morange, E.; Roubault, J.

1986-10-01

French regulation requires licensee to qualify electrical equipment important to safety for service in nuclear power plants to ensure that the equipment can perform its safety function under the set of plausible operating conditions. The French regulatory texts entitled Fundamental safety rules have classified safety related electrical equipment in three main categories: k1, k2, k3, according to their location and operating conditions. The definition of a design basis accident test profile must account for margins applied to thermal hydraulic code outputs. Specific safety margins was added to cover uncertainties in qualification test representativity. Up to now, accidental sequence studies have shown the validity of such a qualification test profile. On the other hand, the results from post accident simulation tests have shown that it is useful not only to validate post accident operating life but also to reveal failures initiated during previous tests [fr

Personal Protective Equipment

National Research Council Canada - National Science Library

1998-01-01

... of personal protective equipment A safety program for new employees is a necessary part of any orientation program An on-going safety program should be used to motivate employees to continue to use...

Reliability analysis of PLC safety equipment

Energy Technology Data Exchange (ETDEWEB)

Yu, J.; Kim, J. Y. [Chungnam Nat. Univ., Daejeon (Korea, Republic of)

2006-06-15

FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system.

Reliability analysis of PLC safety equipment

International Nuclear Information System (INIS)

Yu, J.; Kim, J. Y.

2006-06-01

FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system

School Chemistry Laboratory Safety Guide

Science.gov (United States)

Brundage, Patricia; Palassis, John

2006-01-01

The guide presents information about ordering, using, storing, and maintaining chemicals in the high school laboratory. The guide also provides information about chemical waste, safety and emergency equipment, assessing chemical hazards, common safety symbols and signs, and fundamental resources relating to chemical safety, such as Material…

New Safety rules

CERN Multimedia

Safety Commission

2008-01-01

The revision of CERN Safety rules is in progress and the following new Safety rules have been issued on 15-04-2008: Safety Procedure SP-R1 Establishing, Updating and Publishing CERN Safety rules: http://cern.ch/safety-rules/SP-R1.htm; Safety Regulation SR-S Smoking at CERN: http://cern.ch/safety-rules/SR-S.htm; Safety Regulation SR-M Mechanical Equipment: http://cern.ch/safety-rules/SR-M.htm; General Safety Instruction GSI-M1 Standard Lifting Equipment: http://cern.ch/safety-rules/GSI-M1.htm; General Safety Instruction GSI-M2 Standard Pressure Equipment: http://cern.ch/safety-rules/GSI-M2.htm; General Safety Instruction GSI-M3 Special Mechanical Equipment: http://cern.ch/safety-rules/GSI-M3.htm. These documents apply to all persons under the Director General’s authority. All Safety rules are available at the web page: http://www.cern.ch/safety-rules The Safety Commission

Anchorage of equipment - requirements and verification methods with emphasis on equipment of existing and constructed VVER-type nuclear power plants

International Nuclear Information System (INIS)

Masopust, R.

1999-01-01

Criteria and verification methods which are recommended for use in the capacity evaluation of anchorage of safety-related equipment at WWER-type nuclear power plants are presented. Developed in compliance with the relevant basic standards documents specifically for anchorage of WWER-type equipment components, the criteria and methods cover different types of anchor bolts and other anchorage elements which are typical of existing, constructed, or reconstructed WWER-type nuclear power plants

30 CFR 56.4230 - Self-propelled equipment.

Science.gov (United States)

2010-07-01

....4230 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-SURFACE METAL AND NONMETAL MINES Fire Prevention and Control Firefighting Equipment § 56.4230 Self-propelled equipment. (a)(1) Whenever a fire or its effects...

Safety of information in electronic equipment influenced by the charged space particles

Directory of Open Access Journals (Sweden)

Ksenia Gennad’evna Sizova

2016-10-01

Full Text Available A version of the existing evaluation method of electronic equipment to the influence of charged space particles causing single event effects for the purpose of improving the accuracy of calculation in the field of information safety is suggested. On the basis of the existing and modified methods radiation tolerance of real payload spacecraft responsible for the security of transmitted information are defined. The results of comparison are introduced. Significant differences not only in quantitative but also in qualitative character of tolerance indicators are revealed. It is demonstrated that the modified method allows to take into account the functional complexity of the hardware and the application efficiency of the sophisticated single event effects protection tools. To confirm the applicability of the modified method of equipment tolerance evaluation method to the influence of charged space particles causing single event effects proposals to the procedure of ground tests of the payload and the space experiment are developed.

An appraisal of current regulations and recommendations in relation to proposals for the use of X-ray optics equipment

International Nuclear Information System (INIS)

Weatherley, E.G.

1984-01-01

Current regulatory requirements for the use of x-ray optics equipment in factories are compared with the 19-year-old Guidance Notes for their use in research and teaching establishments. The difficulties in drafting new legislation to cater for both areas of use are discussed and the proposed regulations and approved code of practice having a direct bearing on x-ray optics equipment are reviewed. Comment is made concerning the proposed Health and Safety Executive document, ''Radiation Safety in the use of X-ray Optics Equipment, which will give practical advice and guidance on cost effective methods of achieving the regulatory objectives. The costs involved are unlikely to be significant for the majority of users and implementation of the proposed legislation should not restrict the use of x-ray optics equipment. (author)

Renewal of radiological equipment.

Science.gov (United States)

2014-10-01

In this century, medical imaging is at the heart of medical practice. Besides providing fast and accurate diagnosis, advances in radiology equipment offer new and previously non-existing options for treatment guidance with quite low morbidity, resulting in the improvement of health outcomes and quality of life for the patients. Although rapid technological development created new medical imaging modalities and methods, the same progress speed resulted in accelerated technical and functional obsolescence of the same medical imaging equipment, consequently creating a need for renewal. Older equipment has a high risk of failures and breakdowns, which might cause delays in diagnosis and treatment of the patient, and safety problems both for the patient and the medical staff. The European Society of Radiology is promoting the use of up-to-date equipment, especially in the context of the EuroSafe Imaging Campaign, as the use of up-to-date equipment will improve quality and safety in medical imaging. Every healthcare institution or authority should have a plan for medical imaging equipment upgrade or renewal. This plan should look forward a minimum of 5 years, with annual updates. Teaching points • Radiological equipment has a definite life cycle span, resulting in unavoidable breakdown and decrease or loss of image quality which renders equipment useless after a certain time period.• Equipment older than 10 years is no longer state-of-the art equipment and replacement is essential. Operating costs of older equipment will be high when compared with new equipment, and sometimes maintenance will be impossible if no spare parts are available.• Older equipment has a high risk of failure and breakdown, causing delays in diagnosis and treatment of the patient and safety problems both for the patient and the medical staff.• Every healthcare institution or authority should have a plan for medical imaging equipment upgrade or replacement. This plan should look forward a

Electrical safety in health care area

International Nuclear Information System (INIS)

Amer, G.M.

2011-01-01

An electrical safety in health care area is necessary to protect patients and staff from potential electrical hazards.Functional, accurate and safe clinical equipment is an essential requirement in the provision of health services. Well-maintained equipment will give clinicians greater confidence in the reliability of its performance and contribute to a high standard of client care. Clinical equipment, like all health services, requires annual or periodic servicing of medical equipment. In addition to planned servicing and preventative maintenance, there may be the unexpected failure of medical (and other) equipment, necessitating repair. In general, clinical equipment that has an electrical power source and has direct contact with the client must be serviced as a first priority. In this presentation, a review of the main concepts related to the electrical safety in health area,theinternational standard, the distribution of electric power in hospital and protection against shockwill be introduced. Protection system in hospital will be presented in its two ways: inpower distribution in hospitaland inbiomedical equipment design,finally the optimum maintenance technology and safety tests in health care areawill presented also.

Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category 1 systems for the San Onofre Nuclear Power Plant, Unit 1

International Nuclear Information System (INIS)

Latorre, V.R.; Victor, R.A.

1980-11-01

This report documents the technical evaluation of Southern California Edison Company's San Onofre Nuclear Power Plant, Unit 1, to determine whether the failure of any non-Category 1 (seismic) equipment could result in a condition, such as flooding, that might potentially adversely affect the performance of safety-related equipment required for the safe shutdown of the facility or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection as well as measures taken by Southern California Edison Company to minimize the danger of flooding and to protect safety-related equipment

The specific tasks of RF TSO - FSUE VO 'Safety', related with Implementation of obligations under the Convention on Nuclear Safety

International Nuclear Information System (INIS)

Potapov, V.; Kuznetsov, M.; Kapralov, E.

2010-01-01

It was more than 20 years ago that IAEA discussed the issue pertaining to the need in scientific and engineering support to the regulatory body. The Convention on Nuclear Safety being the keystone in assurance of the global nuclear safety and security regime was adopted in 1994. It is pointed out that two independent organizations supervised by Rostechnadzor have been established within the Russian TSO system, FSUE VO 'Safety' being one of them. The tasks of the organization comprise obligatory certification of equipment as well as acceptance of equipment before its delivery to the NPP both in Russia and in the countries constructing the power units based on the Russian designs. The acceptance procedure has been set forth in the new Russian document at the level of the federal rules and regulations for nuclear safety assurance. As far as its implementation decision is concerned, a task for selection and training of personnel has been set and allocated on the Training and Methodological Center of Nuclear and Radiation Safety established with the support of FSUE VO 'Safety', which provides training programmes and specific lecture courses in the wide range of the relevant topics. (author)

Equipment design guidance document for flammable gas waste storage tank new equipment

International Nuclear Information System (INIS)

Smet, D.B.

1996-01-01

This document is intended to be used as guidance for design engineers who are involved in design of new equipment slated for use in Flammable Gas Waste Storage Tanks. The purpose of this document is to provide design guidance for all new equipment intended for application into those Hanford storage tanks in which flammable gas controls are required to be addressed as part of the equipment design. These design criteria are to be used as guidance. The design of each specific piece of new equipment shall be required, as a minimum to be reviewed by qualified Unreviewed Safety Question evaluators as an integral part of the final design approval. Further Safety Assessment may be also needed. This guidance is intended to be used in conjunction with the Operating Specifications Documents (OSDs) established for defining work controls in the waste storage tanks. The criteria set forth should be reviewed for applicability if the equipment will be required to operate in locations containing unacceptable concentrations of flammable gas

A Study on Performance Evaluation of Safety-Related Protective Coating for Yonggwang Unit 1 and 2

International Nuclear Information System (INIS)

Kim, Sung Young; Kim, Young Bum; Lee, Won Sang

2010-01-01

Protective coating inside nuclear power plants could find its origin from NRC Reg. Guide 1.82(Rev. 3) regarding current issue for the regulation of foreign materials inside containment building. The current issue for the regulation of foreign materials inside containment considered/determined the current issues only regarding the blockage of sump screen by foreign materials such as coating material, insulator, and other materials, while safety-related coating is separately managed by NRC Reg. Guide 1.54(Rev. 1). In this study, we performed field walk-down to evaluate the as-is condition of protective coating inside containment building which was classified as for structure and for equipment with applying the requirement for safety-related coating

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

29 CFR 1926.105 - Safety nets.

Science.gov (United States)

2010-07-01

... 29 Labor 8 2010-07-01 2010-07-01 false Safety nets. 1926.105 Section 1926.105 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR (CONTINUED) SAFETY AND HEALTH REGULATIONS FOR CONSTRUCTION Personal Protective and Life Saving Equipment § 1926.105...

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

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

International Nuclear Information System (INIS)

Anon.

1992-01-01

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

Preliminary safety equipment list for Tank 241-C-106 Manipulator Retrieval System, Project W-340

International Nuclear Information System (INIS)

Guthrie, R.L.

1994-01-01

This document identifies the anticipated safety classification of the estimated major subsystems, based on the projected major functions, that will be used as guidance for the development of the conceptual design of the Manipulator Retrieval System for Tank 241-C-106. This document is intended to be updated as the design of the Manipulator Retrieval System evolves through the conceptual and definitive design phases. The Manipulator Retrieval System is to be capable of removing the hardened sludge heel at the bottom of single shell Tank 241-C-106 and to perform an overall clean out of the tank that leaves a maximum of 360 ft 3 (TPA milestone M-45-00). The thickness of the heel prior to initiation of waste retrieval with the Manipulator Retrieval System is estimated to be 1- to 2-ft. The Manipulator Retrieval System is currently in the pre-conceptual phase with no definitive systems or subsystems. The anticipated retrieval functions for the Manipulator Retrieval System is based on Table 6-2 of WHC-SD-W340-ES-001, Rev. 1. Projected equipment to accomplish these functions were based on the following systems and equipment: Rotary Mode Core Sampling Equipment (WHC-SD-WM-SEL-032); Light Duty Utility Arm System Equipment (WHC-SD-WM-SEL-034); Single Shell Tanks Equipment (WHC-SD-WM-SEL-020)

Qualification of Electrical Equipment in Nuclear Power Plants - Management of ageing

International Nuclear Information System (INIS)

Spaang, Kjell; Staahl, Gunnar

2013-02-01

The purpose of this report is to describe programs and tools for assessment of accomplished and documented qualification with respect to ageing of electrical equipment and for development of complimentary ageing management programs. In addition to description of complete programs for management of ageing, tools for validation of the status with regard to ageing of installed ('old') equipment and, where needed, for complementation of their qualification are also included. The report is restricted to safety related equipment containing ageing sensitive parts, mainly organic materials. To this category belong cables and cable joints and a number of equipment containing oils, seals (o-rings), etc. For equipment located in the containment, the possibilities of continuous supervision are limited. The accessibility for regular inspections is also limited in many cases. The main part of this report deals with the qualification of such equipment. Some safety related equipment outside the containment can be located in areas where they are subjected to high temperature and other excessive environmental stresses during normal operation and in areas affected by an accident. Therefore, some material is given also on qualification of equipment located outside containment with better possibilities for frequent inspection and supervision. Part 1 of the report is an executive summary with a general review of the methodologies and their application. The more detailed description of the programs and underlying material, useful data, etc. is given in Part 2

49 CFR 193.2511 - Personnel safety.

Science.gov (United States)

2010-10-01

... Transportation Other Regulations Relating to Transportation (Continued) PIPELINE AND HAZARDOUS MATERIALS SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) PIPELINE SAFETY LIQUEFIED NATURAL GAS FACILITIES... or a means of escape. (c) Each LNG plant must be equipped with suitable first-aid material, the...

30 CFR 75.811 - High-voltage underground equipment; grounding.

Science.gov (United States)

2010-07-01

...-voltage equipment supplying power to such equipment receiving power from resistance grounded systems shall... 30 Mineral Resources 1 2010-07-01 2010-07-01 false High-voltage underground equipment; grounding... COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Underground High-Voltage...

Qualification of electrical equipment. A United States nuclear system supplier perspective

International Nuclear Information System (INIS)

Jordan, W.G.

1978-01-01

At Westinghouse Pressurized Water Reactor Systems Division (PWR-SD) qualification. of safety related electrical equipment can be segregated into three distinct generations: (1) the initial seismic and environmental qualification programs for electrical equipment (1969-1972); (2) the supplemental sismic and environmental qualification programs (1975-1977); and (3) the seismic and environmental qualification programs to meet IEEE-323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations''(1975). The latter two programs (in a licensing framework), as they are most current, with emphasis on electrical equipment (e.g. transmitters, rack mounted equipment) as opposed to electro-mechanical equipment (valve operators, pump motors) are reviewed. (author)

Normal people working in normal organizations with normal equipment: system safety and cognition in a mid-air collision.

Science.gov (United States)

de Carvalho, Paulo Victor Rodrigues; Gomes, José Orlando; Huber, Gilbert Jacob; Vidal, Mario Cesar

2009-05-01

A fundamental challenge in improving the safety of complex systems is to understand how accidents emerge in normal working situations, with equipment functioning normally in normally structured organizations. We present a field study of the en route mid-air collision between a commercial carrier and an executive jet, in the clear afternoon Amazon sky in which 154 people lost their lives, that illustrates one response to this challenge. Our focus was on how and why the several safety barriers of a well structured air traffic system melted down enabling the occurrence of this tragedy, without any catastrophic component failure, and in a situation where everything was functioning normally. We identify strong consistencies and feedbacks regarding factors of system day-to-day functioning that made monitoring and awareness difficult, and the cognitive strategies that operators have developed to deal with overall system behavior. These findings emphasize the active problem-solving behavior needed in air traffic control work, and highlight how the day-to-day functioning of the system can jeopardize such behavior. An immediate consequence is that safety managers and engineers should review their traditional safety approach and accident models based on equipment failure probability, linear combinations of failures, rules and procedures, and human errors, to deal with complex patterns of coincidence possibilities, unexpected links, resonance among system functions and activities, and system cognition.

Management of radiological related equipments. Creating the equipment management database and analysis of the repair and maintenance records

International Nuclear Information System (INIS)

Eguchi, Megumu; Taguchi, Keiichi; Oota, Takashi; Kajiwara, Hiroki; Ono, Kiyotune; Hagio, Kiyofumi; Uesugi, Ekizo; Kajishima, Tetuo; Ueda, Kenji

2002-01-01

In 1997, we established the committee of equipments maintenance and management in our department. We designed the database in order to classify and register all the radiological related equipments using Microsoft Access. The management of conditions and cost of each equipment has become easier, by keeping and recording the database in the equipments management ledger and by filing the history of repairs or maintenances occurred to modalities. We then accounted numbers, cost of repairs and downtimes from the data of the repair and maintenance records for four years, and we reexamined the causal analysis of failures and the contents of the regular maintenance for CT and MRI equipments that had shown the higher numbers of repairs. Consequently, we have found the improvement of registration method of the data and the more economical way to use of the cost of repair. (author)

Seismic qualification of equipment by means of probabilistic risk assessment

International Nuclear Information System (INIS)

Azarm, M.A.; Farahzad, P.; Boccio, J.L.

1982-01-01

Upon the sponsorship of the Equipment Qualification Branch (EQB) of NRC, Brookhaven National Laboratory (BNL) has utilized a risk-based approach for identifying, in a generic fashion, seismically risk-sensitive equipment. It is anticipated that the conclusions drawn therefrom and the methodology employed will, in part, reconcile some of the concerns dealing with the seismic qualification of equipment in operating plants. The approach taken augments an existing sensitivity analysis, based upon the WASH-1400 Reactor Safety Study (RSS), by accounting for seismicity and component fragility with the Kennedy model and by essentially including the requisite seismic data presented in the Zion Probabilistic Safety Study (ZPSS). Parametrically adjusting the seismic-related variables and ascertaining their effects on overall plant risk, core-melt probability, accident sequence probability, etc., allows one to identify those seismically risk-sensitive systems and equipment. This paper describes the approach taken and highlights the results obtained thus far for a hypothetical pressurized water reactor

Nonfatal tool- or equipment-related injuries treated in US emergency departments among workers in the construction industry, 1998-2005.

Science.gov (United States)

Lipscomb, Hester J; Schoenfisch, Ashley L; Shishlov, Kirill S; Myers, Douglas J

2010-06-01

Individuals in the construction industry are exposed to a variety of tools and pieces of equipment as they work. Data from the National Institute for Occupational Safety and Health (NIOSH) occupational supplement to the National Electronic Injury Surveillance System (NEISS-Work) were used to characterize tool- and equipment-related injuries among workers in the construction industry that were treated in US emergency departments between 1998 and 2005. Based on a national stratified probability sample of US hospitals with 24 hr emergency services, NEISS-Work allows calculation of national injury estimates. Over the 8-year period between 1998 and 2005, we estimated 786,900 (95% CI 546,600-1,027,200) ED-treated tool- or equipment-related injuries identified by the primary or secondary source of injury code. These injuries accounted for a quarter of all ED-treated construction industry injuries. Although over 100 different tools or pieces of equipment were responsible for these injuries, seven were responsible for over 65% of the injury burden: ladders, nail guns, power saws, hammers, knives, power drills, and welding tools in decreasing order. Current injury estimates and their severity, marked by the proportion of cases that were not released after ED treatment, indicate interventions are particularly needed to prevent injuries associated with use of ladders as well as nail guns and power saws. Attention should focus on design and guarding to more efficiently prevent these injuries rather than simply calling for the training of workers in how to safely use a dangerous tool or piece of equipment. 2010 Wiley-Liss, Inc.

Seismic qualification of equipment in operating nuclear power plants. Unresolved safety issue A-46, draft report for comment

International Nuclear Information System (INIS)

Chang, T.Y.

1985-08-01

The margin of safety provided in existing nuclear power plant equipment to resist seismically induced loads and perform their intended safety functions may vary considerably, because of significant changes in design criteria and methods for the seismic qualification of equipment over the years. Therefore, the seismic qualification of equipment in operating plants should be reassessed to determine whether requalification is necessary. The objective of technical studies performed under the Task Action Plan A-46 was to establish an explicit set of guidelines and acceptance criteria to judge the adequacy of equipment under seismic loading at all operating plants, in lieu of requiring qualification to the current criteria that are applied to new plants. This report summarizes the work accomplished on USI A-46 by the Nuclear Regulatory Commission staff and its contractors, Idaho National Engineering Laboratory, Southwest Research Institute, Brookhaven National Laboratory, and Lawrence Livermore National Laboratory. In addition, the collection and review of seismic experience data by the Seismic Qualification Utility Group and the review and recommendations of a group of seismic consultants, the Senior Seismic Review Advisory Panel, are presented. Staff assessment of work accomplished under USI A-46 leads to the conclusion that the use of seismic experience data provides the most reasonable alternative to current qualification criteria. Consideration of seismic qualification by use of experience data was a specific task in USI A-46. Several other A-46 tasks serve to support the use of an experience data base

Upgrading Planning and Executive Strategy for Reactor Protection System and Relative Equipment in Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Jiang Zuyue

2010-01-01

Qinshan Nuclear Power Plant (QNPP) is the first nuclear power plant in China which completed the reactor protection system (RPS) upgrading with new digital safety instrumentation and control (I and C) platform instead of original analog system. At the same time,the nuclear instrumentation system (NIS) was upgraded with the same digital I and C platform. For adapting QNPP's actual engineering situation,the upgrading planning was taken by comprehensively investigating current development and application of digital safety I and C platform in the worldwide scope and by reviewing plant's original systems operation history. The project executive strategy-QNPP's leading role with necessary overseas cooperation and internal technical supports as great as possible, was determined. Some significant factors might influence and restrict the RPS and relative equipment upgrading executive actions in an operating NPP were analyzed.Finally, the engineering feasibility was briefly assessed to recognize the anticipated issues and difficulties and to prepare the relative solutions in advance for the purpose of ensuring the RPS upgrading objectives completely realized. (authors)

Sample-related peripheral equipment at IPNS

International Nuclear Information System (INIS)

Bohringer, D.E.; Crawford, R.K.

1985-01-01

This paper describes samples environment equipment provided by IPNS to visiting users and staff scientists. Of the twelve horizontal neutron beam stations, (ten now operational, two under construction) all use one or more form of such support equipment. An in-house support group devotes a significant fraction of its time to development, calibration, and maintenance of this equipment

Hoisting appliances and fuel handling equipment at nuclear facilities

International Nuclear Information System (INIS)

1987-01-01

The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection

46 CFR 109.347 - Pilot boarding equipment.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Pilot boarding equipment. 109.347 Section 109.347... OPERATIONS Operation and Stowage of Safety Equipment § 109.347 Pilot boarding equipment. (a) The master or person in charge shall ensure that pilot boarding equipment is maintained as follows: (1) The equipment...

Enhancement of nuclear safety culture

International Nuclear Information System (INIS)

Anderson, Stanley J.

1996-01-01

Throughout the 40-year history of the commercial nuclear power industry, improvements have continually been made in the design of nuclear power plants and the equipment in them. In one sense, we have reached an enviable point -- in most plants, equipment failures have become relatively rare. Yet events continue to occur. Regardless of how much the plants are improved, that equipment is operated by people -- highly motivated, well-trained people -- but people nonetheless. And people occasionally make mistakes. By setting the right climate and by setting high standards, good plant management can reduce the number of mistakes made ? and also reduce their potential consequences. Another way to say this is that the proper safety culture must be established and continually improved upon in our nuclear plants. Safety culture is defined by the International Atomic Energy Agency as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance.' In short, we must make safety our top priority

A model for managing cold-related health and safety risks at workplaces.

Science.gov (United States)

Risikko, Tanja; Mäkinen, Tiina M; Påsche, Arvid; Toivonen, Liisa; Hassi, Juhani

2003-05-01

Cold conditions increase health and safety risks at work in several ways. The effects of cold have not been sufficiently taken into consideration in occupational safety and health practices. A systematic model and methods were developed for managing cold-related health and safety risks at workplaces. The development work was performed, in a context-bound manner, in pilot industries and workplaces. The model can be integrated into the company's occupational health and safety management system, such as OHSAS 18001. The cold risks are identified and assessed by using a checklist. The preventive measures are systematically planned in a written form specifically produced for cold workplaces. It includes the organisational and technical preventive measures, protective clothing and personal protective equipment, as well as training and information of the personnel. According to the model, all the workers, foremen, occupational safety personnel and occupational health care personnel are trained to recognise the cold risks and to conduct preventive actions. The developed model was evaluated in the context of cold outdoor (construction) and indoor work (fish processing), and by occupational health and safety professionals. According to the feedback, the model and methods were easy to use after a one-day introduction session. The continuum between the cold risk assessment and management worked well, although there was some overlap in the documentation. The cold risk management model and its methods form an essential part of ISO CD 15743 Strategy for risk assessment, management and work practice in cold environments.

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.

Height, fun and safety in the design of children's playground equipment.

Science.gov (United States)

Wakes, Sarah; Beukes, Amanda

2012-01-01

The study reported in this paper adopted a holistic design approach to investigate issues associated with height related playground injuries from a users' perspective. The main objective was to gain an understanding of the relationship between height and fun so as to establish practical guidelines for addressing the causes of height related injuries whilst maintaining the attributes of playground equipment that children find fun and challenging. Results show that, on the one hand, the risk of injury increases when height is coupled with the use of upper body strength and, on the other hand, that coordination is a greater source of fun and challenge than height for children. Accordingly, it is suggested that the level of risk of injury attached to children's playground equipment can be reduced when the use of lower body strength and coordination are combined with lower free fall heights.

Seismic and environmental qualification of class IE equipment manufactured in Spain

International Nuclear Information System (INIS)

Gerini, P.; Lumbreras, A.; Naredo, F.

1978-01-01

Nuclear power plant instrumentation and control design is affected by several factors such as various plant operating conditions, transient response capability, safety requirements and changes in IEEE standards. Recent upgraded IEEE standards that call for Qualification of all Safety related I anc C equipment, namely IEEE 323 (Qualifying Class IE Electric Equipment for nuclear power generating stations) and its daughter Standard IEEE 344 (Recommended Practices for Seismic Qualification of Class IE Equipment for nuclear power generating stations) have been endorsed by the United States Nuclear Regulatory Commission through the issuance of corresponding Regulatory Guides. The author describes the Qualification requirements applicable to the Class IE I and C Components made in Spain for the different vintages of plants, and the programs implemented or plans established by Westinghouse to fulfill those requirements. (author)

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

Equipment inspection function; A funcao inspecao de equipamentos

Energy Technology Data Exchange (ETDEWEB)

Bressan, Edemir [PETROBRAS (Brazil). Refinaria Alberto Pasqualini

1994-04-01

Development of the safety engineering and equipment inspection policy in petroleum industry, with emphasis to the refineries equipment, are discussed. Presentation of the accidents percentages in petroleum plants under international researches, with details of involved equipment, kind of accident, plant and causes are reported. The start and actual stage of the safety inspection program in the PETROBRAS are also presented 5 refs., 3 figs.

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

International Nuclear Information System (INIS)

1985-07-01

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

Review of equipment aging theory and technology. Final report

Energy Technology Data Exchange (ETDEWEB)

Carfagno, S.P.; Gibson, R.J.

1980-09-01

The theory and technology of equipment aging is reviewed, particularly as they relate to the qualification of safety-system equipment for nuclear power generating stations. A fundamental degradation model is developed, and its relation to more restricted models (e.g., Arrhenius and inverse-stress models) is shown. The most common theoretical and empirical models of aging are introduced, and limitations on their practical application are analyzed. Reliability theory and its application to the acceleration of aging are also discussed. A compendium of aging data for materials and components, including degradation mechanisms, failure modes and activation energies, is included.

Review of equipment aging theory and technology. Final report

International Nuclear Information System (INIS)

Carfagno, S.P.; Gibson, R.J.

1980-09-01

The theory and technology of equipment aging is reviewed, particularly as they relate to the qualification of safety-system equipment for nuclear power generating stations. A fundamental degradation model is developed, and its relation to more restricted models (e.g., Arrhenius and inverse-stress models) is shown. The most common theoretical and empirical models of aging are introduced, and limitations on their practical application are analyzed. Reliability theory and its application to the acceleration of aging are also discussed. A compendium of aging data for materials and components, including degradation mechanisms, failure modes and activation energies, is included

Qualification of Programmable Electronic System (PES) equipment based on international nuclear I and C standards

International Nuclear Information System (INIS)

De Grosbois, J.; Hepburn, G. A.; Olmstead, R.; Goble, W.; Kumar, V.

2006-01-01

Nuclear power plants (NPPs) are increasingly faced with the challenge of qualifying procured equipment, sub-components, and systems that contain digital programmed electronics for use in safety-related applications. Referred to as a 'programmable electronic system' (PES), such equipment typically contains both complex logic that is vulnerable to systematic design faults, and low voltage electronics hardware that is subject to random faults. Procured PES products or components are often only commercial grade, yet can offer reliable cost effective alternatives to custom-designed or nuclear qualified equipment, provided they can be shown to meet the quality assurance, functional safety, environmental, and reliability requirements of a particular application. The process of confirming this is referred to as application-specific product qualification (ASPQ) and can be challenging and costly. This paper provides an overview of an approach that has been developed at Atomic Energy Canada Limited (AECL) and successfully applied to PES equipment intended for use in domestic Candu R 6 nuclear power plants and special purpose reactors at Chalk River Laboratories. The approach has evolved over the past decade and has recently been adapted to be consistent with, and take advantage of new standards that are applicable to nuclear safety-related I and C systems. Also discussed are how recognized third-party safety-certifications of PES equipment to International Electrotechnical Commission (IEC) standards, and the assessment methods employed, may be used to reduce ASPQ effort. (authors)

Survey on requirements for independent reviews and inspections of electrical and I and C equipment

International Nuclear Information System (INIS)

Byman, Karin; Groenkvist, Stefan; Egerbo, Anders; Nilsson, Thomas

2009-03-01

Agencies are involved in almost all technical issues related to the assessment of safety of installations and their operation, including licensing procedures, tests, evaluations, reviews, supervisory procedures, and inspections. The expert organisations are part of the approval processes through their evaluations provided to the authorities in, for instance, expert analysis reports. But the experts are also authorised to give their approval to certain steps of the licensing procedure. The type approval certificate is granted by an accredited inspector. The Swiss nuclear regulatory framework is domestic. General requirements for the design as well as specific criteria for the licensing procedure are provided in the Nuclear Energy Ordinace (KEV). Detailed requirements of, for example, the design, review, and inspection of different kinds of safety-related electrical and IandC equipment are found in regulatory guidelines, which are provided by the Swiss Federal Nuclear Safety Inspectorate (HSK). In Switzerland, there are no general differences between the safety licensing procedure for electrical and IandC equipment and the procedure for mechanical equipment. The licensing procedure follows the four different phases: concept, design, realisation, and installation/commissioning. HSK may, as the federal supervisory authority, seek support from an expert organisation during different steps of the safety licensing procedure for equipment and systems in Swiss nuclear power plants. There are also other examples of the involvement of independent organisations in the safety assurance process for different kinds of equipment in Swiss nuclear power plants. The accredited organisation Electrosuisse inspects electrical installations with regard to general Swiss electrical regulations for personal safety and the Swiss Association for Technical Inspections, SVTI, inspects mechanical equipment

Certification of boiler safety equipment with ALSPA P320 system, to IEC 61508 standard; Certification des equipements de securite chaudiere avec le systeme ALSPA P320 selon la norme IEC 61508

Energy Technology Data Exchange (ETDEWEB)

Dalzon, J.P. [Alstom Power (France)

2002-02-01

Safety and environmental protection are major priorities governing Alstom work on control systems (Alspa P320) for power plant applications, especially when potentially dangerous processes like high-power flame boilers (typically rated at 100 to 600 MW) are involved. Here, automatic protection chains are implemented to guard against the risk of serious incident such as explosion and equipment destruction. In this kind of situation, the IEC 61508 standard provides baseline specifications for certification of safety systems. (authors)

Preserving EQ as per CSA N290.13: EQ of equipment for CANDU NPP

International Nuclear Information System (INIS)

Bordeleau, H.

2006-01-01

EQ is basically a Quality Assurance process that ensures that required safety related equipment will performed their safety functions during accident generating harsh environmental condition. As any good Quality Assurance Program, EQ activities have to be imbedded in the normal tasks of the people at the plant in such a way that it is not an extra concern but part of the things to be assessed and done on a regular basis so that the EQ Status of the equipment will not be compromised. The experience shows us that even with brand new plants there are discrepancies between equipment designed to be installed and that installed (wrong model or wrong instrument). Improper installation was also found. To eliminate these mistakes it is recommended to have a thorough initial inspection during construction or commissioning and later at regular intervals during operation. To avoid new mistakes during maintenance, crews have to be properly qualified and trained on completed procedure with appropriate work report that will required to be scrutinized for possible common cause failure and OPEX report. Procurement and handling of equipment and spare parts require a thorough revision of procurement and storage practices to guarantee proper storage, availability and identification of equipment for EQ replacement. A shelf life program for degradable material shall be implemented. Condition monitoring program could be use to extend life of equipment or as a qualification methods, caution should be use to look very precisely to the critical parameters that need to be monitored and to carefully tailor the monitoring program. EQ program should also consider barriers, they are not EQ equipment; it is usually made out of non-ageing material like metal or concrete. But there could be sealing material used in conjunction with the barriers and these may need special care. The impact of EQ parameters changes related to services conditions, new technical data or condition monitoring has to be

Rolls-Royce successful modernization of safety-critical Instrumentation and Control (I and C) equipment at the Dukovany VVER 440/213 Nuclear Power Plant, based on SPINLINE 3 platform

International Nuclear Information System (INIS)

Rebreyend, P.; Burel, J.P.; Spoc, J.; Karasek, A.

2010-01-01

Rolls-Royce has provided on-time delivery of a substantial safety-critical I and C overhaul for four Nuclear reactors operated by Czech Republic utility, CEZ a.s. This nine-year project is considered to be one of the largest I and C modernization projects in the world. The Dukovany VVER 440 I and C modernization project and its key success factors are profiled in this paper. The project is in the final stages with the last unit to be completed in 2009. Beginning in September 2000, the project is in compliance with the initial schedule. Rolls-Royce has been designing and manufacturing I and C solutions dedicated to the implementation of safety and safety-related functions in nuclear power plants (NPPs) for more than 30 years. Though the early solutions were non-software-based, since 1984 software-based solutions for safety I and C functions have been deployed in operating NPPs across France and 15 other countries. The Rolls-Royce platform is suitable for implementation of safety I and C functions in new NPPs, as well as in the modernization of safety equipment in existing plants. CEZ a.s. is a major electricity supplier for the national grid. At Dukovany, CEZ a.s. operates four units of VVER-440/213-type reactors producing one quarter of CEZ a.s. electricity production. The first of these units was connected to the grid in 1985. Since the year 2000, the nine-year modernization program has been underway at Dukovany, at a cost of more than 200 million Euros. The equipment replacement was implemented during regular, planned outages of the original equipment and systems. After an international bidding phase, CEZ a.s. awarded a contract to Skoda JS for general engineering and project management. Individual subcontracts were then signed between Skoda JS and a consortium between Rolls-Royce and Areva for modernization of the safety systems, including the Reactor Protection System (RPS), the Reactor Control System (RCS), and the Post-Accident Monitoring System (PAMS). Two

Shields calculations for teletherapy equipment. Regulatory approach of the National Center of Nuclear Safety

International Nuclear Information System (INIS)

Fuente P, A. de la; Dumenigo G, C.; Quevedo G, J.R.; Lopez F, Y.

2006-01-01

The evaluation of applications of construction licenses for the new services of radiotherapy has occupied a significant space in the activity developed by the National Center of Nuclear Safety (CNSN) in the last 2 years. Presently work the experiences of the authors in the evaluation of the required shield for the local where cobalt therapy equipment and lineal accelerators of medical use are used its are exposed, the practical problems detected are approached during the application of the methodologies recommended in both cases and its are discussed which have been the suppositions of items accepted by the Regulatory Authority for the realization of these shield calculations. The accumulated experience allows to assure that the realistic application of the item data and the rational use of the engineering logic makes possible to design local for radiotherapy equipment that fulfill the established dose restrictions in the in use legislation in Cuba, without it implies an excessive expense of construction materials. (Author)

Safety classification of items in Tianwan Nuclear Power Plant

International Nuclear Information System (INIS)

Sun Yongbin

2005-01-01

The principle of integrality, moderation and equilibrium should be considered in the safety classification of items in nuclear power plant. The basic ways for safety classification of items is to classify the safety function based on the effect of the outside enclosure damage of the items (parts) on the safety. Tianwan Nuclear Power Plant adopts Russian VVER-1000/428 type reactor, it safety classification mainly refers to Russian Guidelines and standards. The safety classification of the electric equipment refers to IEEE-308(80) standard, including 1E and Non 1E classification. The safety classification of the instrumentation and control equipment refers to GB/T 15474-1995 standard, including safety 1E, safety-related SR and NC non-safety classification. The safety classification of Tianwan Nuclear Power Plant has to be approved by NNSA and satisfy Chinese Nuclear Safety Guidelines. (authors)

A Study on Performance and Safety Tests of Defibrillator Equipment.

Science.gov (United States)

Tavakoli Golpaygani, A; Movahedi, M M; Reza, M

2017-12-01

Nowadays, more than 10,000 different types of medical devices can be found in hospitals. This way, medical electrical equipment is being employed in a wide variety of fields in medical sciences with different physiological effects and measurements. Hospitals and medical centers must ensure that their critical medical devices are safe, accurate, reliable and operational at the required level of performance. Defibrillators are critical resuscitation devices. The use of reliable defibirillators has led to more effective treatments and improved patient safety through better control and management of complications during Cardiopulmonary Resuscitation (CPR). The metrological reliability of twenty frequent use, manual defibrillators in use ten hospitals (4 private and 6 public) in one of the provinces of Iran according to international and national standards was evaluated. Quantitative analysis of control and instrument accuracy showed the amount of the obtained results in many units are critical which had less value over the standard limitations especially in devices with poor battery. For the accuracy of delivered energy analysis, only twelve units delivered acceptable output values and the precision in the output energy measurements especialy in weak battry condition, after activation of discharge alarm, were low. Obtained results indicate a need for new and severe regulations on periodic performance verifications and medical equipment quality control program especially for high risk instruments. It is also necessary to provide training courses on the fundumentals of operation and performane parameters for medical staff in the field of meterology in medicine and how one can get good accuracy results especially in high risk medical devices.

Equipment Obsolescence Management Program

Energy Technology Data Exchange (ETDEWEB)

Redmond, J.

2014-07-01

Nuclear Power Plant (NPP) Operators are challenged with securing reliable supply channels for safety related equipment due to equipment obsolescence. Many Original Equipment Manufacturers (OEMs) have terminated production of spare parts and product life-cycle support. The average component life cycles are much shorter than the NPP design life, which means that replacement components and parts for the original NPP systems are not available for the complete design life of the NPPs. The lack or scarcity of replacement parts adversely affects plant reliability and ultimately the profitability of the affected NPPs. This problem is further compounded when NPPs pursue license renewal and approval for plant-life extension. A reliable and predictable supply of replacement co components is necessary for NPPs to remain economically competitive and meet regulatory requirements and guidelines. Electrical and I and C components, in particular, have short product life cycles and obsolescence issues must be managed pro actively and not reactively in order to mitigate the risk to the NPP to ensure reliable and economic NPP operation. (Author)

Maintenance and test strategies to optimize NPP equipment performance

International Nuclear Information System (INIS)

Mayer, S.; Tomic, B.

2000-01-01

This paper proposes an approach to maintenance optimization of nuclear power plant components, which can help to increase both safety and availability. In order to evaluate the benefits of preventive maintenance on a quantitative basis, a software code has been developed for component performance and reliability simulation of safety related nuclear power plant equipment. A three state Markov model will be introduced, considering a degraded state in addition to an operational state and a failed state. (author)

Safety evaluation report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446)

International Nuclear Information System (INIS)

1988-11-01

Supplement 19 to the Safety Evaluation Report related to the operation of the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 (NUREG-0797), has been prepared by the Office of Special Projects of the US Nuclear Regulatory Commission (NRC). The facility is located in Somervell County, Texas, approximately 40 miles southwest of Fort Worth, Texas. This supplement presents the staff's evaluation of the Texas Utilities Electric Company's (lead applicant's) corrective action program (CAP) related to equipment qualification. The scope and methodology for the CAP workscope, as summarized in Revision 0 to the Equipment Qualification Project Status Report and as detailed in related documents, were developed to resolve various issues raised by the Comanche Peak Response Team (CPRT) and the NRC staff to ensure that plant equipment is appropriately environmentally and/or seismically and dynamically qualified and documented in accordance with the validated plant design resulting from other CAP scopes of work for Unit 1 and areas common to Units 1 and 2. The staff concludes that the CAP workscope for equipment qualification provides a comprehensive program for resolving the concerns identified by the CPRT and the NRC staff, including issues raised in the Comanche Peak Safety Evaluation Report and its supplements, and its implementation will ensure that the environmental and/or seismic and dynamic qualification of equipment at CPSES satisfies the validated plant design and the applicable requirements of 10 CFR Part 50. As is routine staff practice, the NRC staff will verify the adequacy of implementation of the environmental and seismic and dynamic equipment qualification program at CPSES during inspections that will take place before fuel loading. 97 refs

Working Safely at Some Times and Unsafely at Others: A Typology and Within-Person Process Model of Safety-Related Work Behaviors.

Science.gov (United States)

Beus, Jeremy M; Taylor, William D

2017-06-22

Why do individuals choose to work safely in some instances and unsafely in others? Though this inherently within-person question is straightforward, the preponderance of between-person theory and research in the workplace safety literature is not equipped to answer it. Additionally, the limited way in which safety-related behaviors tend to be conceptualized further restricts understanding of why individuals vary in their safety-related actions. We use a goal-focused approach to conceptually address this question of behavioral variability and contribute to workplace safety research in 2 key ways. First, we establish an updated typology of safety-related behaviors that differentiates behaviors based on goal choice (i.e., safe vs. unsafe behaviors), goal-directedness (i.e., intentional vs. unintentional behaviors), and the means of goal pursuit (i.e., commission vs. omission and promotion vs. prevention-focused behaviors). Second, using an expectancy-value theoretical framework to explain variance in goal choice, we establish within-person propositions stating that safety-related goal choice and subsequent behaviors are a function of the target of safety-related behaviors, the instrumentality and resource requirement of behaviors, and the perceived severity, likelihood, and immediacy of the threats associated with behaviors. Taken together, we define what safety-related behaviors are, explain how they differ, and offer propositions concerning when and why they may vary within-persons. We explore potential between-person moderators of our theoretical propositions and discuss the practical implications of our typology and process model of safety-related behavior. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

Statistical analysis of the behaviour of the mechanical equipment of EDFs power plants - evaluation of the availability and safety of thermal and nuclear units

International Nuclear Information System (INIS)

Procaccia, H.; Brillon, A.; Cravero, M.; Lucenet, G.

1975-01-01

The investigation and research directorate of EDF has undertaken a statistical analysis of the behaviour of large mechanical equipment at conventional power stations during the ten years following the operating reports of these stations. It has thus been possible to determine the intrinsic reliability, the failure rate, the mean repair time, and the mean good operating time of feed water reheating points, power turbines, pumps and boilers of the various EDF plants (125 and 250 MW) leading to a consideration of the feasibility of an extrapolation to present and future plants. Based on these elementary investigation two methods of calculation have been developed. One is used to assess the overall availability of a thermal or nuclear power station based on the knowledge of the failure rates of the equipment, each piece of equipment being associated with an idea of its technical importance in the functioning of the equipment. A numerical application is given for 125 and 250 MW conventional plants. The purpose of the other method is to estimate the operational safety of the safety equipment of nuclear power stations, based on the development of tree diagrams for faults in basic equipment. A numerical example is given for the cooling systems for Phenix and for one of the Super Phenix versions. (author)

Usability analysis of industrial cooking equipment.

Science.gov (United States)

Calado, Alexana Vilar Soares; Soares, Marcelo Márcio

2012-01-01

This paper refers to the comparative study of the equipment used for cooking in commercial of kitchens restaurants that use the system of traditional cooking and those ones which use the system called smart cooking (combination oven). The study investigates the usability issues concerning to the two systems, analyzing comparatively the aspects related to anthropometry, dimensional variables, the use of the product and also the product safety, as well as issues of information related to operation of the new concepts of cooking in intelligent systems.

Modernization of Unit 2 at Oskarshamn NPP- Main Objectives, Experience from Design, Separation of Operational and Nuclear Safety Equipment - Lessons Learned

International Nuclear Information System (INIS)

Kanaan, Salah K.

2015-01-01

This paper aims to give a picture of Oskarshamn Nuclear Power Plant (OKG) experience from design for one of the biggest modernization project in the world and focuses on what was learned that is specific to robustness of electrical power systems, especially through Fukushima Station Blackout (SBO). The planning for unit 2 at OKG was initiated in 1967 and the plant was completed on time and was synchronized to the grid October 2, 1974 and is of type BWR. Unit 2 was originally on 580 MW. In 1982 a thermal power up-rate was performed, from 1700 MWh to 1800 MWh (106% reactor output). A decision was made to perform a modernization and a new power up-rate to 850 MW and there were several reasons for this decision; New safety regulations from Swedish Radiation Safety Authority (SSM), Ageing of important components and the initial focus was on safety and availability - Project Plant Life Extension (Plex) was established and became the largest nuclear power modernization in the world. The modernization will lead to: - New safety concept with 4 divisions instead for existing 2 with 2 new buildings South Electrical Building (SEB) and North Electrical Building (NEB); - Completely new software - based equipments for monitoring, control and I and C; - New Low Pressure Turbine, new generator and main transformer; - New MCR and simulator; - Compliance with modern reactor safety requirements; - Redundancy, Separation, Diversification, Earthquake; - Reinforcement of existing safety functions; - New Electricity - I and C (electric power incl. reinforced emergency power and control systems); - New buildings for Electricity - I and C; - Reinforcement of existing process systems as well as installation of new ones. Based on studies and good experiences on how to separate the operational and the safety equipment, the project led to a completely new safety concept. The safety concept is based on fully understanding the safety system that shall encompass all of the elements required to

Systems analysis of radiation safety during dismantling of power-plant equipment at a nuclear power station

International Nuclear Information System (INIS)

Bylkin, B.K.; Shpitser, V.Ya.

1993-01-01

A systems analysis of the radiation safety makes possible an ad hoc determination of the elements forming the system, as well as the establishment of the characteristics of their interaction with radiation-effect factors. Here the authors will present part of the hierarchical analysis procedure, consisting in general of four separate procedures. The purpose is to investigate and analyze the mean and stable (on the average) indices of radiation safety, within the framework of alternative mathematical models of dismantling the power-plant equipment of a nuclear power station. The following three of the four procedures are discussed: (1) simulated projection, of the processing of radioactive waste; (2) analysis of the redistribution of radionuclides during the industrial cycle of waste treatment; (3) planning the collective dose load during the dismantling operation. Within the framework of the first of these procedures, the solutions to the problem of simulating a waste-treatment operation of maximum efficiency are analyzed. This analysis is based on the use of a data base for the parameters of the installations, assemblies, and equipment, enabling the integration of these in a simulation of a complex automated facility. The results were visualized in an AUTOCAD-10 medium using a graphical data base containing an explanation of the rooms

Days individual equipment of protection and professional risks; Equipements de protection individuelle et risques professionnelles

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

The personal protection equipment is studied in the legal way (legal liabilities, certification, European texts), technical way (ergonomics, painfulness of ventilated equipment wearing, reliability of a respirable air line, protection gloves against the chemical risk, exposure to nano particulates, working in hot area), human factors (hostile area and emotion management), studies on personal equipment such evaluation, efficiency, conception of new equipment, physiological tolerance, limit of use, and some general safety studies on the working places. (N.C.)

Survey on requirements for independent reviews and inspections of electrical and I and C equipment

Energy Technology Data Exchange (ETDEWEB)

Byman, Karin; Groenkvist, Stefan; Egerbo, Anders; Nilsson, Thomas (AaF Consult AB, Stockholm (Sweden))

2009-03-15

are involved in almost all technical issues related to the assessment of safety of installations and their operation, including licensing procedures, tests, evaluations, reviews, supervisory procedures, and inspections. The expert organisations are part of the approval processes through their evaluations provided to the authorities in, for instance, expert analysis reports. But the experts are also authorised to give their approval to certain steps of the licensing procedure. The type approval certificate is granted by an accredited inspector. The Swiss nuclear regulatory framework is domestic. General requirements for the design as well as specific criteria for the licensing procedure are provided in the Nuclear Energy Ordinace (KEV). Detailed requirements of the design, review, and inspection of different kinds of safety-related electrical and IandC equipment are found in regulatory guidelines, which are provided by the Swiss Federal Nuclear Safety Inspectorate (HSK).

Terms standardization between the rules of diagnosis radiation equipment safety management and atomic energy law: problems and suggestions

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwa Gon; Kang, Se Sik; Kim, Chang Soo; Park, Cheol Seo [Catholic University of Pusan, Busan (Korea, Republic of)

2006-03-15

The rules and terms are described different meaning in this results the research is accomplished for preventing practical workers from confusion. Atomic law are kept up modification and development in our situation by the ICRP's recommendation, on the other hand, the rules of diagnosis radiation equipment safety managements are modified partial, then resulted in confusion. The study was comparison between the rules of diagnosis radiation equipment safety management and atomic energy law, and the modification items obtained were as follows. With each other different the terms and units are used. With the exception of special terms for affairs usage, it is needless to say that common term uniformity is standardized. The standardization of rules and guidance have not need to confusion radiological practical workers. The following is omitted. The radiation protection against the patient and the hospital visitor. Radiation dose limit of the woman patient who is in the process of becoming pregnant. Radiation dose limit of the person who is not regarded as medical exposure. The control of the exposure of pregnant of women at work.

Terms standardization between the rules of diagnosis radiation equipment safety management and atomic energy law: problems and suggestions

International Nuclear Information System (INIS)

Kim, Hwa Gon; Kang, Se Sik; Kim, Chang Soo; Park, Cheol Seo

2006-01-01

The rules and terms are described different meaning in this results the research is accomplished for preventing practical workers from confusion. Atomic law are kept up modification and development in our situation by the ICRP's recommendation, on the other hand, the rules of diagnosis radiation equipment safety managements are modified partial, then resulted in confusion. The study was comparison between the rules of diagnosis radiation equipment safety management and atomic energy law, and the modification items obtained were as follows. With each other different the terms and units are used. With the exception of special terms for affairs usage, it is needless to say that common term uniformity is standardized. The standardization of rules and guidance have not need to confusion radiological practical workers. The following is omitted. The radiation protection against the patient and the hospital visitor. Radiation dose limit of the woman patient who is in the process of becoming pregnant. Radiation dose limit of the person who is not regarded as medical exposure. The control of the exposure of pregnant of women at work

Interactive effects of relay and circuit breaker aging in a safety-related system

International Nuclear Information System (INIS)

Toman, G.J.; Bacanskas, V.P.; Shook, T.A.; Ladlow, C.C.; Gunther, W.

1987-01-01

This paper provides an overview of the results of a program to evaluate the aging of circuit breakers and relays and the effects of that aging on the function of a safety system used in nuclear power plants. The program was performed under the Nuclear Plant Aging Research (NPAR) Program of the US Nuclear Regulatory Commission under subcontract to Brookhaven National Laboratory. There were two primary aspects to the program. In the first, the aging and failure modes of relays and circuit breakers were determined by evaluating the construction, design, and materials and the failure data related to nuclear power plant service. In the second, the interactions between a safety system and its relays and circuit breakers were evaluated to determine the effects of relay and circuit breaker aging on the function of the safety system. The aging of relays and circuit breakers was assessed through evaluation of failure data bases, discussions with utility personnel, and evaluation of equipment operating and maintenance manuals. The interaction study was based on an analysis of the safety injection system of a pressurized water reactor. The effects of stresses from the system were analyzed for the tendency to cause deterioration of the relays and circuit breakers in the system. Then the effect of the deterioration of relays and circuit breakers on the functional capability of the safety system was evaluated

Inquiry relating to safety due to modification of usage of nuclear fuel material (establishment of waste safety testing facility) in Tokai Laboratory, Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

1979-01-01

Application was made to the director of the Science and Technology Agency (STA) for the license relating to the modification of usage of nuclear fuel material (the establishment of waste safety testing facility) from the director of the Japan Atomic Energy Research Institute on November 30, 1978. After passing through the safety evaluation in the Nuclear Safety Bureau of STA, inquiry was conducted to the head of the Atomic Energy Safety Commission (AESC) on June 6, 1979, from the director of the STA. The head of AESC directed to conduct the safety examination to the head of the Nuclear Fuel Safety Examination Specialist Committee on June 7, 1979. The content of the modification of usage of nuclear fuel material is the establishment of waste safety testing facility to study and test the safety relating to the treatment and disposal of high level radioactive liquid wastes due to the reprocessing of spent fuel. As for the results of the safety examination, the siting of the waste safety testing facility which is located in the Tokai Laboratory, Japan Atomic Energy Research Institute (JAERI), and the test plan of the glass solidification of high level radioactive liquid are presented as the outline of the study plan. The building, main equipments including six cells, the isolation room and the glove box, the storage, and the disposal facilities for gas, liquid and solid wastes are explained as the outline of the facilities. Concerning the items from the viewpoint of safety, aseismatic design, slightly vacuum operation, shielding, decay heat removal, fire protection, explosion protection, criticality management, radiation management and environmental effect were evaluated, and the safety was confirmed. (Nakai, Y.)

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

30 CFR 75.509 - Electric power circuit and electric equipment; deenergization.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Electric power circuit and electric equipment... LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Electrical Equipment-General § 75.509 Electric power circuit and electric equipment; deenergization. [Statutory Provisions] All...

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…

Regulatory process for material handling equipment

International Nuclear Information System (INIS)

Rajendran, S.; Agarwal, Kailash

2017-01-01

Atomic Energy (Factories) Rules (AEFR) 1996, Rule 35 states, 'Thorough inspection and load testing of a Crane shall be done by a Competent Person at least once every 12 months'. To adhere to this rule, BARC Safety Council constituted 'Material Handling Equipment Committee (MHEC)' under the aegis of Conventional Fire and Safety Review Committee (CFSRC) to carry out periodical inspection and certification of Material Handling Equipment (MHE), tools and tackles used in BARC Facilities at Trombay, Tarapur and Kalpakkam

Hoisting appliances and fuel handling equipment at nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-31

The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection. 36 refs. Translation. The original text is published under the same guide number. The guide is valid from 5 January 1987 and will be in force until further notice.

Maintenance program guidelines for programmatic equipment

International Nuclear Information System (INIS)

1994-11-01

The Division Directors at Lawrence Berkeley Laboratory are responsible for implementing a maintenance program for research equipment (also referred to as programmatic equipment) assigned to them. The program must allow maintenance to be accomplished in a manner that promotes operational safety, environmental protection and compliance, and cost effectiveness; that preserves the intended functions of the facilities and equipment; and that supports the programmatic mission of the Laboratory. Programmatic equipment -- such as accelerators, lasers, radiation detection equipment, and glove boxes -- is dedicated specifically to research. Installed equipment, by contrast, includes the mechanical and electrical systems installed as part of basic building construction, equipment essential to the normal functioning of the facility and its intended use. Examples of installed equipment are heating, ventilating, and air conditioning systems; elevators; and communications systems. The LBL Operating and Assurance Program Plan (PUB-3111, Revision 4) requires that a maintenance program be prepared for programmatic equipment and defines the basic maintenance program elements. Such a program of regular, documented maintenance is vital to the safety and quality of research activities. As a part of that support, this document offers guidance to Laboratory organizations for developing their maintenance programs. It clarifies the maintenance requirements of the Operating and Assurance Program (OAP) and presents an approach that, while not the only possibility, can be expected to produce an effective maintenance program for research equipment belonging to the Laboratory's organizations

Auditing of Monitoring and Respiratory Support Equipment in a Level III-C Neonatal Intensive Care Unit

Directory of Open Access Journals (Sweden)

Elena Bergon-Sendin

2015-01-01

Full Text Available Background. Random safety audits (RSAs are a safety tool but have not been widely used in hospitals. Objectives. To determine the frequency of proper use of equipment safety mechanisms in relation to monitoring and mechanical ventilation by performing RSAs. The study also determined whether factors related to the patient, time period, or characteristics of the area of admission influenced how the device safety systems were used. Methods. A prospective observational study was conducted in a level III-C Neonatal Intensive Care Unit (NICU during 2012. 87 days were randomly selected. Appropriate overall use was defined when all evaluated variables were correctly programmed in the audited device. Results. A total of 383 monitor and ventilator audits were performed. The Kappa coefficient of interobserver agreement was 0.93. The rate of appropriate overall use of the monitors and respiratory support equipment was 33.68%. Significant differences were found with improved usage during weekends, OR 1.85 (1.12–3.06, p=0.01, and during the late shift (3 pm to 10 pm, OR 1.59 (1.03–2.4, p=0.03. Conclusions. Equipment safety systems of monitors and ventilators are not properly used. To improve patient safety, we should identify which alarms are really needed and where the difficulties lie for the correct alarm programming.

A Study on Performance and Safety Tests of Defibrillator Equipment

Directory of Open Access Journals (Sweden)

Tavakoli Golpaygani A.

2017-12-01

Full Text Available Introduction: Nowadays, more than 10,000 different types of medical devices can be found in hospitals. This way, medical electrical equipment is being employed in a wide variety of fields in medical sciences with different physiological effects and measurements. Hospitals and medical centers must ensure that their critical medical devices are safe, accurate, reliable and operational at the required level of performance. Defibrillators are critical resuscitation devices. The use of reliable defibirillators has led to more effective treatments and improved patient safety through better control and management of complications during Cardiopulmonary Resuscitation (CPR. Materials and Methods: The metrological reliability of twenty frequent use, manual defibrillators in use ten hospitals (4 private and 6 public in one of the provinces of Iran according to international and national standards was evaluated. Results: Quantitative analysis of control and instrument accuracy showed the amount of the obtained results in many units are critical which had less value over the standard limitations especially in devices with poor battery. For the accuracy of delivered energy analysis, only twelve units delivered acceptable output values and the precision in the output energy measurements especialy in weak battry condition, after activation of discharge alarm, were low. Conclusion: Obtained results indicate a need for new and severe regulations on periodic performance verifications and medical equipment quality control program especially for high risk instruments. It is also necessary to provide training courses on the fundumentals of operation and performane parameters for medical staff in the field of meterology in medicine and how one can get good accuracy results especially in high risk medical devices.

Safety status system for operating room devices.

Science.gov (United States)

Guédon, Annetje C P; Wauben, Linda S G L; Overvelde, Marlies; Blok, Joleen H; van der Elst, Maarten; Dankelman, Jenny; van den Dobbelsteen, John J

2014-01-01

Since the increase of the number of technological aids in the operating room (OR), equipment-related incidents have come to be a common kind of adverse events. This underlines the importance of adequate equipment management to improve the safety in the OR. A system was developed to monitor the safety status (periodic maintenance and registered malfunctions) of OR devices and to facilitate the notification of malfunctions. The objective was to assess whether the system is suitable for use in an busy OR setting and to analyse its effect on the notification of malfunctions. The system checks automatically the safety status of OR devices through constant communication with the technical facility management system, informs the OR staff real-time and facilitates notification of malfunctions. The system was tested for a pilot period of six months in four ORs of a Dutch teaching hospital and 17 users were interviewed on the usability of the system. The users provided positive feedback on the usability. For 86.6% of total time, the localisation of OR devices was accurate. 62 malfunctions of OR devices were reported, an increase of 12 notifications compared to the previous year. The safety status system was suitable for an OR complex, both from a usability and technical point of view, and an increase of reported malfunctions was observed. The system eases monitoring the safety status of equipment and is a promising tool to improve the safety related to OR devices.

Hydrogen Safety Issues Compared to Safety Issues with Methane and Propane

International Nuclear Information System (INIS)

Green, Michael A.

2005-01-01

The hydrogen economy is not possible if the safety standards currently applied to liquid hydrogen and hydrogen gas by many laboratories are applied to devices that use either liquid or gaseous hydrogen. Methane and propane are commonly used by ordinary people without the special training. This report asks, 'How is hydrogen different from flammable gasses that are commonly being used all over the world?' This report compares the properties of hydrogen, methane and propane and how these properties may relate to safety when they are used in both the liquid and gaseous state. Through such an analysis, sensible safety standards for the large-scale (or even small-scale) use of liquid and gaseous hydrogen systems can be developed. This paper is meant to promote discussion of issues related to hydrogen safety so that engineers designing equipment can factor sensible safety standards into their designs

Hydrogen Safety Issues Compared to Safety Issues with Methane andPropane

Energy Technology Data Exchange (ETDEWEB)

Green, Michael A.

2005-08-20

The hydrogen economy is not possible if the safety standards currently applied to liquid hydrogen and hydrogen gas by many laboratories are applied to devices that use either liquid or gaseous hydrogen. Methane and propane are commonly used by ordinary people without the special training. This report asks, 'How is hydrogen different from flammable gasses that are commonly being used all over the world?' This report compares the properties of hydrogen, methane and propane and how these properties may relate to safety when they are used in both the liquid and gaseous state. Through such an analysis, sensible safety standards for the large-scale (or even small-scale) use of liquid and gaseous hydrogen systems can be developed. This paper is meant to promote discussion of issues related to hydrogen safety so that engineers designing equipment can factor sensible safety standards into their designs.

Intermediate probabilistic safety assessment approach for safety critical digital systems

International Nuclear Information System (INIS)

Taeyong, Sung; Hyun Gook, Kang

2001-01-01

Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)

Radiation therapy sources, equipment and installations

International Nuclear Information System (INIS)

2011-03-01

The safety code for Telegamma Therapy Equipment and Installations, (AERB/SC/MED-1) and safety code for Brachytherapy Sources, Equipment and Installations, (AERB/SC/MED-3) were issued by AERB in 1986 and 1988 respectively. These codes specified mandatory requirements for radiation therapy facilities, covering the entire spectrum of operations ranging from the setting up of a facility to its ultimate decommissioning, including procedures to be followed during emergency situations. The codes also stipulated requirements of personnel and their responsibilities. With the advent of new techniques and equipment such as 3D-conformal radiation therapy, intensity modulated radiation therapy, image guided radiation therapy, treatment planning system, stereotactic radiosurgery, stereotactic radiotherapy, portal imaging, integrated brachytherapy and endovascular brachytherapy during the last two decades, AERB desires that these codes be revised and merged into a single code titled Radiation Therapy Sources, Equipment, and Installations

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

AERB information booklet: personal protective equipment- safety footwear

International Nuclear Information System (INIS)

1992-01-01

The main classes of safety footwear required for industrial operations in the units of Department of Atomic Energy are the following; leather safety boots and shoes, firemen's leather boots - Wellington type, electrical safety shoes, chemical safety shoes, shoes suitable for mining operations. The criteria to be adopted for selection of safety shoes for nuclear installations are given. (M.K.V.). 5 annexures, 1 appendix

Modification and backfitting at the Oskarshamn Nuclear Power Plant Unit 2 in safety related systems

International Nuclear Information System (INIS)

Karlsson, Leif; Nilsson, Ove; Lidh, B.

1995-05-01

This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Oskarshamn-2, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 3 refs

Management methodology for pressure equipment

Science.gov (United States)

Bletchly, P. J.

Pressure equipment constitutes a significant investment in capital and a major proportion of potential high-risk plant in many operations and this is particularly so in an alumina refinery. In many jurisdictions pressure equipment is also subject to statutory regulation that imposes obligations on Owners of the equipment with respect to workplace safety. Most modern technical standards and industry codes of practice employ a risk-based approach to support better decision making with respect to pressure equipment. For a management system to be effective it must demonstrate that risk is being managed within acceptable limits.

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.

Elevator and Escalator Safety Education for Older Adults.

Science.gov (United States)

Hanks, Roma Stovall

1996-01-01

In eight focus groups in five cities, older adults identified their concerns about safety on elevators and escalators, often related to misunderstanding of the equipment. Their preferences for delivery of safety information included video/television, pamphlets, discussions, and posters. Educational interventions and modifications for disabilities…

Nuclear EQ sourcebook: A compilation of documents for nuclear equipment qualification

International Nuclear Information System (INIS)

Anon.

1992-01-01

In the nuclear power industry, environmental and seismic qualification of safety-related electrical and instrumentation equipment is collectively known as Equipment Qualification (EQ). Related technology, as well as regulatory requirements, have evolved rapidly during the last 15 years. For environmental qualification, what began in 1971 with one trial-use guide (IEEE Std 323-1971), now stands as a full complement of Nuclear Regulatory Commission (NRC) rules, guides and industry standards. In addition to the Institute of Electrical and Electronics Engineers (IEEE), the American Society of Mechanical Engineers (ASME) has also undertaken development of its own set of standards for use in qualifying safety-related mechanical equipment. To ensure that the original design and qualification is preserved, engineers need to select and use the correct set of NRC regulations, regulatory guides, industry standards, and generic correspondence. Given that the total number of these documents exceed 200, this task becomes resource intensive. This compilation is the first known publication available to serve the user-community need for a complete and exhaustive single source. Approximately 180 items (Bulletins, Federal Rules, Generic Letters, Notices, Regulatory Guides, IEEE Standards, IEEE Recommended Practices, and IEEE Guides) have been processed separately for inclusion on the data base

Cryogenic Safety Rules and Guidelines at CERN

CERN Multimedia

CERN. Geneva

2016-01-01

CERN defines and implements a Safety Policy that sets out the general principles governing safety at CERN. As an intergovernmental organisation, CERN further establishes its own Safety Rules as necessary for its proper functioning. In this process, it takes into account the laws and regulation of the Host States (France and Switzerland), EU regulations and directives, as well as international regulations, standards and directives. For the safety of cryogenic equipment, this is primarily covered by the Safety Regulation for Mechanical Equipment and the General Safety Instruction for Cryogenic Equipment. In addition, CERN has also developed Safety Guidelines to support the implementation of these safety rules, covering cryogenic equipment and oxygen deficiency hazard assessment and mitigation. An overview of the cryogenic safety rules and these safety guidelines will be presented.

Dukovany nuclear power plant safety

International Nuclear Information System (INIS)

1999-01-01

Presentation covers recommended safety issues for the Dukovany NPP which have been solved with satisfactory conclusions. Safety issues concerned include: radiation safety; nuclear safety; security; emergency preparedness; health protection at work; fire protection; environmental protection; chemical safety; technical safety. Quality assurance programs at all stages on NPP life time is described. Report includes description of NPP staff training provision, training simulator, emergency operating procedures, emergency preparedness, Year 2000 problem, inspections and life time management. Description of Dukovany Plant Safety Analysis Projects including integrity of the equipment, modernisation, equipment innovation and safety upgrading program show that this approach corresponds to the actual practice applied in EU countries, and fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of MORAWA Equipment Upgrading program

Seismic qualification of multiple interconnected safety-related cabinets in a high seismic zone

International Nuclear Information System (INIS)

Khan, M.R.; Chen, W.H.W.; Wang, T.Y.

1993-01-01

Certain safety-related multiple, interconnected electrical cabinets and the devices contained therein are required to perform their intended safety functions during and after a design basis seismic event. In general, seismic testing is performed to ensure the structural integrity of the cabinets and the functionality of their associated devices. Constrained by the shake table capacity, seismic testing is usually performed only for a limited number of interconnected cabinets. Also, original shake table tests performed usually did not provide detailed response information at various locations inside the cabinets. For operational and maintenance purposes, doors and panels of some cabinets may need to be opened while the adjacent cabinets are required to remain functional. In addition, in-cabinet response spectra need to be generated for the seismic qualification of new devices and the replacement parts. Consequently, seismic analysis of safety-related multiple, interconnected cabinets is frequently required for configurations which are different from the original tested conditions. This paper presents results of seismic tests of three interconnected safety-related cabinets and finite element analyses performed to compare the analytical results with those obtained from the cabinet seismic tests. Parametric analyses are performed to determine how many panels and doors can be opened while the adjacent cabinets still remain functional. The study indicates that for cabinets located in a high seismic zone, the critical damping of the cabinet is significantly higher than 5% to 7% typically used in qualifying electrical equipment. For devices mounted on the cabinet doors to performed their intended safety function, it requires stiffening of doors and that these doors be properly bolted to the cabinet frame. It also shows that even though doors and panels bolted to the cabinet frame are the primary seismic resistant element of the cabinet, opening of a limited number of them

Transport of bundles and equipment which contain radioactive material

International Nuclear Information System (INIS)

1987-01-01

This norm settles down: 1) The requirements that should be completed in relation to safety precautions and protection against ionizing radiations during the transport radioactive material and/or equipment containing it, in order to avoid risks to the collective and the environment. 2) The basic information on procedures that will be completed in the event of happening accidents during the transport or the transit storage of radioactive material and/or equipment that contain it. 3) The measures of security and physical protection during the transport of radioactive material and/or equipment containing it. This norm is applied: 1) To all the ways of transport (by air, by ground and by ship, fluvial and marine) of radioactive material and/or equipment that contain it. 2) To all natural or legal, public or private person, devoted to install, produce, trade, market, import or export radioactive materials and/or equipment containing it, and that needs to transport them as main or secondary activity [es

33 CFR 142.24 - Use of equipment.

Science.gov (United States)

2010-07-01

... CONTINENTAL SHELF ACTIVITIES WORKPLACE SAFETY AND HEALTH Personal Protective Equipment § 142.24 Use of equipment. (a) Each holder of a lease or permit issued under the Act shall ensure that all personnel who are required by this subpart to use or wear personal protective equipment do so when within the lease area or...

14 CFR 27.1413 - Safety belts.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety belts. 27.1413 Section 27.1413 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: NORMAL CATEGORY ROTORCRAFT Equipment Safety Equipment § 27.1413 Safety belts. Each safety belt...

30 CFR 57.12027 - Grounding mobile equipment.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Grounding mobile equipment. 57.12027 Section 57.12027 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND NONMETAL MINES Electricity...

Implications of NUREG-1150 for equipment qualification

International Nuclear Information System (INIS)

Siu, R.P.; Sarram, M.

1988-01-01

Current USNRC regulations in 10CFR50 requires that structures, systems and components important to safety in a nuclear power plant are to be designed to accommodate the effects of harsh environments associated with potential accidents. Detailed regulatory requirements are as discussed in Regulatory Guide 1.89. Some of these regulatory requirements have been based upon rather outdated understanding of the phenomenology of severe accidents -- e.g., the radiation source terms for equipment qualification can be traced to TID-14844 which was originally published in 1962. Furthermore, the definition of important-to-safety in these regulatory requirements is rather broad, usually resulting in a long Q-list of equipment to be qualified for the design-basis Large LOCA. In the past few years, extensive research/analysis on severe accidents have been conducted -- from the IREP/NREP activities, the IDCOR programs, to the NUREG-1150 efforts. This paper reports that as a result of all these PRA-related activities, a much-improved understanding of the phenomenology and environmental conditions associated with severe accidents has been gained

Effects of safety climate on safety norm violations: exploring the mediating role of attitudinal ambivalence toward personal protective equipment.

Science.gov (United States)

Cavazza, Nicoletta; Serpe, Alessandra

2009-01-01

Research on the role of organizational and psychosocial factors in influencing risk behaviors and the likelihood of injury at work showed that safety climate also has great impact on workers' behavior. However, the mechanisms through which this impact operates are still partially unclear. In order to explore the role that attitudinal ambivalence toward wearing PPE might play in mediating the impact of safety climate on safety norm violations, a questionnaire was administered to 345 Italian workers. Three dimensions of safety climate (i.e., company safety concern, senior managers' safety concern, supervisors' attitudes towards safety) were found to be positively associated with the individual ambivalence level, whereas the fourth one (i.e., work pressure) was negatively correlated with it. In turn, low levels of ambivalence were associated with a lower tendency to break the safety norms, even though the perception of a good safety climate also maintained a direct effect on unsafe behaviors. Designers of training program for the prevention of work related injuries must pay great attention to the psycho-social factors (such as the effects of the safety climate perception by employees on their attitudes and behaviors), and include specific contents into the prevention programs in order to improve workers compliance with safety norms.

Seismic fragility levels of nuclear power plant equipment

International Nuclear Information System (INIS)

Bandyopadhyay, K.K.; Hofmayer, C.H.

1987-01-01

Seismic fragility levels of safety-related electrical and mechanical equipment used in nuclear power plants are discussed. The fragility level is defined as the vibration level corresponding to initiation of equipment malfunctions. The test response spectrum is used as a measure of this vibration level. The fragility phenomenon of an equipment is represented by a number of response spectra corresponding to various failure modes. Analysis methods are described for determination of the fragility level by use of existing test data. Useful conversion factors are tabulated to transform test response spectra from one damping value to another. Results are presented for switch-gears and motor control centers. The capacity levels of these equipment assemblies are observed to be limited by malfunctioning of contactors, motor starters, relays and/or switches. The applicability of the fragility levels, determined in terms of test response spectra, to Seismic Margin Studies and Probabilistic Risk Assessments is discussed and specific recommendations are provided

Methodology of assessment of the impact on the risk of changes in peak times of unavailability of safety equipment integrating treatment and analysis of uncertainties

International Nuclear Information System (INIS)

Martorell, S.; Villamizar, M.; Marton, I.; Villanueva, J. f.; Carlos, S.; Sanchez, A. I.

2013-01-01

Surveillance of safety equipment requirements form parts of the technical specifications of Performance included in the bases of operation of nuclear power plants license, so they are center of attention in the study of improvements to the safety of the operation of the plant using different methods, both traditional deterministic as probabilistic.

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)

29 CFR 1926.1000 - Rollover protective structures (ROPS) for material handling equipment.

Science.gov (United States)

2010-07-01

... 29 Labor 8 2010-07-01 2010-07-01 false Rollover protective structures (ROPS) for material handling equipment. 1926.1000 Section 1926.1000 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY... CONSTRUCTION Rollover Protective Structures; Overhead Protection § 1926.1000 Rollover protective structures...

49 CFR 232.609 - Handling of defective equipment with ECP brake systems.

Science.gov (United States)

2010-10-01

... (ECP) Braking Systems § 232.609 Handling of defective equipment with ECP brake systems. (a) Ninety-five... systems. 232.609 Section 232.609 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION BRAKE SYSTEM SAFETY STANDARDS FOR FREIGHT...

Environmental tests of a digital safety channel: An investigation of stress-related vulnerabilities of computer-based safety system

International Nuclear Information System (INIS)

Korsah, K.; Wilson, T.L.; Wood, R.; Tanaka, T.

1997-01-01

This article presents the results of environmental stress tests performed on an experimental digital safety channel (EDSC) assembled at the Oak Ridge National Laboratory as part of the Qualification of Advanced Instrumentation and Controls Systems Research program, which was sponsored by the US Nuclear Regulatory Commission. The program is expected to provide recommendations for environmental qualification of digital safety systems. The purpose of the study was to investigate potential vulnerabilities of distributed computer systems used in safety applications when subjected to environmental stressors. The EDSC assembled for the tests employs technologies and digital subsystems representative of those proposed for use in advanced light-water reactors or as retrofits in existing plants. Subsystems include computers, electrical and optical serial communication links, fiber-optic network links, analog-to-digital and digital-to-analog converters, and multiplexers. The EDSC was subjected to selected stressors that are a potential risk to digital equipment in a mild environment. The selected stressors were electromagnetic and radiofrequency interferences (EMI-RFI), temperature, humidity, and smoke exposure. The stressors were applied at levels of intensity considerably higher than the safety channel is likely to experience in a normal nuclear power plant environment. Ranges of stress were selected at a sufficiently high level to induce errors so that failure modes that are characteristic of the technologies employed could be identified. On the basis of the incidence of functional errors observed during testing, EMI-RFI, smoke exposure, and high temperature coupled with high relative humidity, in that order, were found to have the greatest impact of the stressors investigated. The most prevalent stressor-induced upsets, as well as the most severe, were found to occur during the EMI-RFI tests

The evolution of cryogenic safety at Fermilab

International Nuclear Information System (INIS)

Stanek, R.; Kilmer, J.

1992-12-01

Over the past twenty-five years, Fermilab has been involved in cryogenic technology as it relates to pursuing experimentation in high energy physics. The Laboratory has instituted a strong cryogenic safety program and has maintained a very positive safety record. The solid commitment of management and the cryogenic community to incorporating safety into the system life cycle has led to policies that set requirements and help establish consistency for the purchase and installation of equipment and the safety analysis and documentation

30 CFR 56.12027 - Grounding mobile equipment.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Grounding mobile equipment. 56.12027 Section 56.12027 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-SURFACE METAL AND NONMETAL MINES Electricity § 56...

The application of GIS equipment in nuclear power plant

International Nuclear Information System (INIS)

Ji Lin; Huang Pengbo; Chang Xin'ai

2012-01-01

In this paper, the advantage and disadvantage of gas insulated switchgear (GIS) in environmental adaptability, operation safety and economic benefit are analyzed. Issues concerning the manufacture, transportation, on-site installation, operation, maintenance and extension of GIS equipment are discussed. Comparing those characteristics with air insulated switchgear (AIS), GIS is characterized by better aseismic ability, less occupied area and installation process, lower fault rate, longer maintenance period, easier for extension and higher economic benefit, SF6 gas insures the operation safety and reliability of GIS equipment, modular transport and re-assembling improves the installation flexibility. Therefore, GIS equipment may be the first choice for the primary equipment of nuclear power plant. (authors)

Preliminary studies on the impact of smoke on digital equipment

International Nuclear Information System (INIS)

Tanaka, T.J.; Korsah, K.; Antonescu, C.

1995-01-01

Last year the USNRC initiated a program at Sandia National Laboratories to determine the potential impact of smoke on advanced safety-related digitial instrumentation. In recognition of the fact that the reliability of safety-related equipment during or shortly after a fire in a nuclear power plant is more risk significant than long-term effects, we are concentrating on short-term failures. We exposed a multiplexer module board to three different types of smoke to determine whether the smoke would affect its operation. The operation of the multiplexer board was halted by one out of the three smoke exposures. In coordination with Oak Ridge National Laboratory, an experimental digital safety system was also smoke tested. The series of tests showed that smoke can cause potentially serious failures of a safety system. Most of these failures were intermittent and showed that smoke can temporarily interrupt communication between digital systems

Environmental, safety, and health engineering

International Nuclear Information System (INIS)

Woodside, G.; Kocurek, D.

1997-01-01

A complete guide to environmental, safety, and health engineering, including an overview of EPA and OSHA regulations; principles of environmental engineering, including pollution prevention, waste and wastewater treatment and disposal, environmental statistics, air emissions and abatement engineering, and hazardous waste storage and containment; principles of safety engineering, including safety management, equipment safety, fire and life safety, process and system safety, confined space safety, and construction safety; and principles of industrial hygiene/occupational health engineering including chemical hazard assessment, personal protective equipment, industrial ventilation, ionizing and nonionizing radiation, noise, and ergonomics

Criticality safety evaluation in Tokai Reprocessing Plant

International Nuclear Information System (INIS)

Shirai, Nobutoshi; Nakajima, Masayoshi; Takaya, Akikazu; Ohnuma, Hideyuki; Shirouzu, Hidetomo; Hayashi, Shinichiro; Yoshikawa, Koji; Suto, Toshiyuki

2000-04-01

Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 'Criticality safety of single unit' in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units. (author)

Safety Management at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Ligam, A.S.; Zarina Masood; Ahmad Nabil Abdul Rahim

2011-01-01

Adequate safety measures and precautions, which follow relevant safety standards and procedures, should be in place so that personnel safety is assured. Nevertheless, the public, visitor, contractor or anyone who wishes to enter or be in the reactor building should be well informed with the safety measures applied. Furthermore, these same elements of safety are also applied to other irradiation facilities within the premises of Nuclear Malaysia. This paper will describes and explains current safety management system being enforced especially in the TRIGA PUSPATI Reactor (RTP) namely radiation monitoring system, safety equipment, safe work instruction, and interconnected internal and external health, safety and security related departments. (author)

Equipment nonconformance and degradation: Promptly determining operability and establishing corrective action plans

International Nuclear Information System (INIS)

Hoxie, C.L.; Cotton, K.R.; Emch, R.L. Jr.

1992-01-01

Nine principles for dealing with degraded and nonconforming equipment are presented and some examples are discussed. The distinction between equipment operability (i.e., capability to perform the safety function) and equipment qualification (conformance to all aspects of the current licensing basis, including codes and standards, design criteria, and commitments) is discussed. The concept of finding reasonable assurance of safety for continued plant operation for equipment not covered by technical specifications is also presented. Degraded or nonconforming equipment must be evaluated for its safety impact and for operability. In all cases, degraded or nonconforming conditions must ultimately be resolved, either through prompt corrective action or through some process of showing that the changed state of the plant is acceptable for continued operation, based on 10 CFR 50.59

76 FR 6088 - Installed Systems and Equipment for Use by the Flightcrew

Science.gov (United States)

2011-02-03

... comments we receive, without change, to http://www.regulations.gov , including any personal information you... often result from a sequence, or combination, of flightcrew errors and safety related events. The design... enables new kinds of flightcrew interfaces that previous requirements do not address. Equipment Behavior...

Load Bearing Equipment for Bone and Muscle

Science.gov (United States)

Shackelford, Linda; Griffith, Bryan

2015-01-01

Resistance exercise on ISS has proven effective in maintaining bone mineral density and muscle mass. Exploration missions require exercise with similar high loads using equipment with less mass and volume and greater safety and reliability than resistance exercise equipment used on ISS (iRED, ARED, FWED). Load Bearing Equipment (LBE) uses each exercising person to create and control the load to the partner.

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

Diagnostic x-ray equipment and its use

International Nuclear Information System (INIS)

1993-01-01

The guide contains the requirements for the use, structure and operation of diagnostic x-ray equipment necessary to ensure the radiation safety of workers and patients. The guide is intended for the party responsible for operating the x-ray units, the staff operating, installing and maintaining x-ray examination equipment, and companies manufacturing and selling such equipment. Special requirements for mammographic and dental x-ray equipments are given in separate ST guides. (5 refs., 1 figs., 2 tabs.)

Fire safety regulations and licensing

International Nuclear Information System (INIS)

Berg, H.P.

1998-01-01

Experience of the past tow decades of nuclear power plant operation and results obtained from modern analytical techniques confirm that fires may be a real threat to nuclear safety and should receive adequate attention from the design phase throughout the life of the plant. Fire events, in particular influence significantly plant safety due to the fact that fires have the potential to simultaneously damage components of redundant safety-related equipment. Hence, the importance of fire protection for the overall safety of a nuclear power plant has to be reflected by the fire safety regulations and to be checked during the licensing process of a plant as well as during the continuous supervision of the operating plant

Challenges of equipment qualification using today's standards with emphasis on a class 1E motor program

International Nuclear Information System (INIS)

Deaton, K.

1995-01-01

This paper describes qualification of new equipment for safety related service in the nuclear power industry in accordance with current industry standards. This topic is presented from the perspective of an original equipment manufacturer J(OEM). Equipment qualification is first discussed in a general way then an example is provided of an electric motor qualification. A review of alternative qualification methods including commercial dedication is included. Potential difficulties with alternative/expedited qualification methods are also discussed

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)

Equipment reliability process improvement and preventive maintenance optimization

International Nuclear Information System (INIS)

Darragi, M.; Georges, A.; Vaillancourt, R.; Komljenovic, D.; Croteau, M.

2004-01-01

The Gentilly-2 Nuclear Power Plant wants to optimize its preventive maintenance program through an Integrated Equipment Reliability Process. All equipment reliability related activities should be reviewed and optimized in a systematic approach especially for aging plants such as G2. This new approach has to be founded on best practices methods with the purpose of the rationalization of the preventive maintenance program and the performance monitoring of on-site systems, structures and components (SSC). A rational preventive maintenance strategy is based on optimized task scopes and frequencies depending on their applicability, critical effects on system safety and plant availability as well as cost-effectiveness. Preventive maintenance strategy efficiency is systematically monitored through degradation indicators. (author)

Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

Energy Technology Data Exchange (ETDEWEB)

Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

2005-03-15

The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments.

Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

International Nuclear Information System (INIS)

Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

2005-03-01

The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments

15 CFR Notes Applicable to State... - Notes applicable to State of Understanding related to Medical Equipment:

Science.gov (United States)

2010-01-01

... 15 Commerce and Foreign Trade 2 2010-01-01 2010-01-01 false Notes applicable to State of Understanding related to Medical Equipment: applicable Notes applicable to State of Understanding related to Medical Equipment: Commerce and Foreign Trade Regulations Relating to Commerce and Foreign Trade (Continued) BUREAU OF INDUSTRY AND SECURITY,...

Quantitative techniques for medical equipment maintenance management

NARCIS (Netherlands)

Ben Houria, Zeineb; Masmoudi, Malek; Al Hanbali, Ahmad; Khatrouch, Ikram; Masmoudi, Faouzi

2016-01-01

The maintenance department in a hospital is responsible for ensuring the safety of medical equipment and their availability while keeping the operation costs minimal. The selection of the best maintenance strategy is a key decision to reduce the equipment downtime, increase the availability, and

Manual on safety aspects of the design and equipment of hot laboratories. 1981 ed

International Nuclear Information System (INIS)

1981-01-01

This manual covers the general principles of planning and design of areas inside laboratories according to the varying potential radiation and contamination hazards; enclosures for radioactive material containment; viewing and lighting systems and various types of manipulators; transfer and transport of radioactive materials within the laboratories; air cleaning and ventilation systems, with particular reference to IAEA Safety Series No.17; techniques for controlling air pollution from the operation of nuclear facilities; various radioactive waste disposal systems; criticality control; fire protection; personnel monitoring, including changing-room monitoring and protective clothing; standardization and automation; and administrative controls. Although alpha, beta, gamma technologies have developed separately, equipment used in radioactive work is common to many operations. There is a step change in technology between work with uranium and plutonium and between work with plutonium and other transuranics; with plutonium one enters the field of alpha, beta, gamma technology. This manual reports the basic requirements and gives reference to more sophisticated techniques available. It is not concerned with work on a commercial scale. Other publications of interest in this context, for instance IAEA Safety Series No.39, are referenced for more detailed information.

Reliability prediction for the vehicles equipped with advanced driver assistance systems (ADAS and passive safety systems (PSS

Directory of Open Access Journals (Sweden)

Balbir S. Dhillon

2012-10-01

Full Text Available The human error has been reported as a major root cause in road accidents in today’s world. The human as a driver in road vehicles composed of human, mechanical and electrical components is constantly exposed to changing surroundings (e.g., road conditions, environmentwhich deteriorate the driver’s capacities leading to a potential accident. The auto industries and transportation authorities have realized that similar to other complex and safety sensitive transportation systems, the road vehicles need to rely on both advanced technologies (i.e., Advanced Driver Assistance Systems (ADAS and Passive Safety Systems (PSS (e.g.,, seatbelts, airbags in order to mitigate the risk of accidents and casualties. In this study, the advantages and disadvantages of ADAS as active safety systems as well as passive safety systems in road vehicles have been discussed. Also, this study proposes models that analyze the interactions between human as a driver and ADAS Warning and Crash Avoidance Systems and PSS in the design of vehicles. Thereafter, the mathematical models have been developed to make reliability prediction at any given time on the road transportation for vehicles equipped with ADAS and PSS. Finally, the implications of this study in the improvement of vehicle designs and prevention of casualties are discussed.

Modification and backfitting at the Barsebaeck Nuclear Power Plant Unit 1 and 2 in safety related systems

International Nuclear Information System (INIS)

Karlsson, Leif; Nilsson, Ove; Lidh, B.

1995-05-01

This report is intended for use by the Swedish Nuclear Power Inspectorate. It has been published to enable comparison of modification and backfitting implemented at Barsebaeck, with those implemented at other plants, both domestic and abroad. The report summarizes the more notable modifications and backfitting carried out on any safety-related equipment, or software, at Barsebaeck, and covers the decade 1984 to 1994. Modifications to hardware, and to some extent to software, are catalogued, but not described in any detail. No general procedures (operational or maintenance) are dealt with. 3 refs

Earthquake protection of essential civil and industrial equipments

International Nuclear Information System (INIS)

Bourrier, P.; Le Breton, F.; Thevenot, A.

1986-01-01

This document presents the principal reflexions concerning seismic engineering applications for equipment and the difference of the non-employment towards these structures. The notion of essential equipment is then pointed out as well as the main particularities of equipment considered as structures. Finally, this document illustrates a few pathological examples encountered after an earthquake, and presents some equipments of a nuclear power plant which to resist an increased safety seism [fr

30 CFR 57.9330 - Clearance for surface equipment.

Science.gov (United States)

2010-07-01

... NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND NONMETAL MINES Loading, Hauling, and Dumping Safety Devices, Provisions, and Procedures for Roadways, Railroads, and Loading and... the farthest projection of moving railroad equipment shall be provided on at least one side of the...

47 CFR 80.1085 - Ship radio equipment-General.

Science.gov (United States)

2010-10-01

... SERVICES STATIONS IN THE MARITIME SERVICES Global Maritime Distress and Safety System (GMDSS) Equipment... maritime safety information service, as identified by the IMO GMDSS Master Plan Publication, is provided... Publication 117, or the Admiralty List of Radio Signals Volume 5 Global Maritime Distress and Safety System...

30 CFR 57.4230 - Surface self-propelled equipment.

Science.gov (United States)

2010-07-01

... Section 57.4230 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND NONMETAL MINES Fire... fire or its effects could impede escape from self-propelled equipment, a fire extinguisher shall be on...

46 CFR 197.326 - Oxygen safety.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Oxygen safety. 197.326 Section 197.326 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE OCCUPATIONAL SAFETY AND HEALTH STANDARDS GENERAL PROVISIONS Commercial Diving Operations Equipment § 197.326 Oxygen safety. (a) Equipment used with...

Resolution of Unresolved Safety Issue A-48, ''Hydrogen control measures and effects of hydrogen burns on safety equipment''

International Nuclear Information System (INIS)

Ferrell, C.M.; Soffer, L.

1989-09-01

Unresolved Safety Issue (USI) A-48 arose as a result of the large amount of hydrogen generated and burned within containment during the Three Mile Island accident. This issue covers hydrogen control measures for recoverable degraded-core accidents for all boiling-water reactors (BWRs) and those pressurized-water reactors (PWRs) with ice-condenser containments. The Commission and the nuclear industry have sponsored extensive research in this area, which has led to significant revision of the Commission's hydrogen control regulations, given in Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Section 50.44. BWRs having Mark I and II containments are presently required to operate with inerted containment atmospheres that effectively prevent hydrogen combustion. BWRs with Mark III containments and PWRs with ice-condenser containments are now required to be equipped with hydrogen control systems to protect containment integrity and safety systems inside containment. Industry has chosen to use hydrogen igniter systems to burn hydrogen produced in a controlled fashion to prevent damage. An independent review by a Committee of the National Research Council concluded that, for most accident scenarios, current regulatory requirements make it highly unlikely that hydrogen detonation would be the cause of containment failure. On the basis of the extensive research effort conducted and current regulatory requirements, including their implementation, the staff concludes that no new regulatory guidance on hydrogen control for recoverable degraded-core accidents for these types of plants is necessary and that USI A-48 is resolved

Allowed outage time for test and maintenance - Optimization of safety

International Nuclear Information System (INIS)

Cepin, M.; Mavko, B.

1997-01-01

The main objective of the project is the development and application of methodologies for improvement and optimization of test and maintenance activities for safety related equipment in NPPs on basis of their enhanced safety. The probabilistic safety assessment serves as a base, which does not mean the replacement of the deterministic analyses but the consideration of probabilistic safety assessment results as complement to deterministic results. 15 refs, 2 figs

Microbiological safety cabinets, cytotoxic safety cabinets. Choice and use

International Nuclear Information System (INIS)

Balty, I.; Belhanini, B.; Clermont, H.; Cornu, J.C.; Jacquet, M.A.; Texte, J.C.

2003-01-01

Drawn up by a working group composed of prevention professionals, manufacturers and control bodies, this guide is intended to help those responsible for safety in laboratories choose and acquire materials responding to the intended protection objectives. It provides recommendations for the commissioning and control of these materials as well as for their use. After a description of the operational characteristics of safety cabinets, this guide looks at the important points to be taken into account when ordering, accepting and commissioning equipment adapted to precise needs. It also covers verification of correct operation and provides a number of common sense rules relative to precautions for use. Material cleaning and decontamination is described briefly on account of the very specialized character of this activity. Detailed information relative to this subject should be sought in the literature. (authors)

Procuring, maintaining and using equipment: the key issues in controlling exposure to patients

International Nuclear Information System (INIS)

Hendra, I.R.F.

1989-01-01

Radiological equipment is not purpose-built for each individual application. Thus, not only are acceptance criteria directly related to the design of equipment and the quality assurance exercised during its manufacture, but also the performance, safety and reliability of equipment must feature in the requirements for procuring new equipment. In the UK this issue is dealt with under the Department of Health and Social Security Manufacturer Registration Scheme for Medical Equipment in its application to radiological equipment. Similarly the continuing reliability of equipment in service is directly related to the quality of maintenance services. This is determined largely by the manner in which such services are procured. In the UK this issue has been considered by one of four Working Parties commissioned by the South-East Thames Regional Health Authority who are the ''Centre of Responsibility'' for diagnostic imaging equipment. The report prepared from the first phase of activity of this working group describes features of a procurement specification for maintenance services and prescribes the complementary duties of contractors and customers. The use in X-ray departments of the kind of manufacturing quality system that is used by most reputable equipment manufacturers has the potential for a dramatic reduction in exposure to patients. (author)

The Impact Of Occupational Hazards In Workplaces - Maintenance, A Main Target For Ensuring The Safety Of Working Equipment

Science.gov (United States)

Antonov, Anca Elena; Buica, Georgeta; Darabont, Doru Costin; Beiu, Constantin

2015-07-01

For use of work equipment having the economic performance and the highest level of safety, it must be ensured that it complies with the conditions set by the manufacturer in terms of putting into service, use and maintenance operations, ensuring appropriate technical and environmental requirements, including appropriate measures and means of protection. The research aimed to identify and analyze the occupational hazards associated to maintenance operations, in terms of the history of the adjustments, maintenance, cleaning and repair, including the case that occur after the incidents, capital repairs and upgrades. The results of the research consisted in the development of recommendations on the effective management of maintenance activities of work equipment and a software model to enable an efficient management of maintenance, as a tool for occupational hazards in companies - premise for increasing the competitiveness of employers in the market economy.

Planning of maintenance of electrical equipment in nuclear plants/laboratories [Paper No.: VB-3

International Nuclear Information System (INIS)

Narasinga Rao, S.N.; Bhattacharyya, A.K.

1981-01-01

Satisfactory operating performance of electrical systems ensures continuous availability of power to the various plants and machinery in nuclear plant and laboratories. For effective optimal functioning of the electrical equipment and to reduce their down time, scheduled planning of maintenance to the equipment is essential. Maintenance of power plant, nuclear or fossil, and industrial plant and research laboratories demands essential ingredients such as right type of trained and motivated technical personnel, adoption of standard procedures for maintenance, adequate safety and protection for equipment, safety procedures adopted in the installation to prevent hazards to the workers, provision of adequate stores and inventories, facilities for quick repairs and testing of equipment and effective planning of procedures for their maintenance. While breakdown maintenance allows equipment to operate before it is repaired or replaced, preventive maintenance makes use of scheduled inspection and periodical equipment overhaul and has little value for predicting future continuous performances of equipment. The engineered maintenance is most advantageous and offers maximum operating time to reduce down time of the equipment while adding predictive testing technique to aid in determining the frequency of overhaul of equipment. The important checks to be conducted and preventive maintenance programme to be scheduled are discussed in this paper. The safety and reliable functioning of the electrical equipment depend on proper optimal design, selection of equipment, their installation, subsequent maintenance and strict compliance with safety regulations. (author)

Challenges of equipment qualification using today`s standards with emphasis on a class 1E motor program

Energy Technology Data Exchange (ETDEWEB)

Deaton, K.

1995-02-01

This paper describes qualification of new equipment for safety related service in the nuclear power industry in accordance with current industry standards. This topic is presented from the perspective of an original equipment manufacturer J(OEM). Equipment qualification is first discussed in a general way then an example is provided of an electric motor qualification. A review of alternative qualification methods including commercial dedication is included. Potential difficulties with alternative/expedited qualification methods are also discussed.

46 CFR 197.454 - First aid and treatment equipment.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false First aid and treatment equipment. 197.454 Section 197.454 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE OCCUPATIONAL SAFETY AND... Equipment § 197.454 First aid and treatment equipment. The diving supervisor shall ensure that medical kits...

Evaluation of safety requirements of erbium laser equipment used in dentistry; Avaliacao de requisitos de seguranca de um equipamento a laser de erbio para fins odontologicos

Energy Technology Data Exchange (ETDEWEB)

Braga, Flavio Hamilton

2002-07-01

The erbium laser (Er:YAG) has been used in several therapeutic processes. Erbium lasers, however, operate with energies capable to produce lesions in biological tissues. Aiming the safe use, the commercialization of therapeutic laser equipment is controlled in Brazil, where the equipment should comply with quality and safety requirement prescribed in technical regulations. The objective of this work is to evaluate the quality and safety requirements of a commercial therapeutic erbium laser according to Brazilian regulations, and to discuss a risk control program intended to minimize the accidental exposition at dangerous laser radiation levels. It was verified that the analyzed laser can produce lesions in the skin and eyes, when exposed to laser radiation at distances smaller than 80 cm by 10 s or more. In these conditions, the use of protection glasses is recommended to the personnel that have access to the laser operation ambient. It was verified that the user's training and the presence of a target indicator are fundamental to avoid damages in the skin and buccal cavity. It was also verified that the knowledge and the correct use of the equipment safety devices, and the application of technical and administrative measures is efficient to minimize the risk of dangerous expositions to the laser radiation. (author)

Identification of new unresolved safety issues relating to nuclear power plants - special report to Congress. Congressional report

International Nuclear Information System (INIS)

1981-03-01

As a result of NRC staff review and extended collegial consultations and investigations within the NRC, the Commission has designated four new Unresolved Safety Issues (USIs). This report describes the process used to evaluate the large number of concerns and recommendations which resulted from the major investigations of the Three Mile Island-2 accident as well as other events and investigations of the past year, and the report identifies the four new USIs selected as follows: (1) Shutdown decay heat removal requirements (Task A-45); (2) Seismic qualification of equipment in operating plants (Task A-46); (3) Safety implications of control systems (Task A-47); and (4) Hydrogen control measures and effects of hydrogen burns on safety equipment (Task A-48). Appendix A of the report presents an expanded discussion of each new USI including issue definition, a preliminary discussion of the action plan and a basis for continued plant operations and licensing. Appendix B of the report provides a brief discussion of each of the candidate safety issues not designated as an USI

Managing electrical safety

CERN Document Server

Wiggins, James H, Jr

2001-01-01

Managing Electrical Safety provides an overview of electric basics, hazards, and established standards that enables you to understand the hazards you are likely to encounter in your workplace. Focusing on typical industrial environments-which utilize voltages much higher than household or office circuits-the author identifies the eight key components of an electrical safety program and examines each using a model safety management process. You'll learn how to identify electrical hazards, how to prescribe necessary electrical Personal Protective Equipment, how to ensure that equipment is de-ene

DIFFICULTIES IN WORK SAFETY MANAGEMENT IN A COMPANY PRODUCING STEEL FLAT BARS

Directory of Open Access Journals (Sweden)

Marta Niciejewska

2018-01-01

Full Text Available According to research, metal processing industry has the highest risk of accidents at work, due to the necessity for personal protective equipment use and strict execution of safety rules and procedures. The paper presents the individual characteristics of health and safety management in the company producing steel flat bars. Special attention has been paid to the safety aspect of the use of automated logistical equipment (gantries, automatics that increase the risk of work-related accidents. The results of research on the difficulties resulting from the specifics of production (metallurgical industry has been presented.

Laboratory safety handbook

Science.gov (United States)

Skinner, E.L.; Watterson, C.A.; Chemerys, J.C.

1983-01-01

Safety, defined as 'freedom from danger, risk, or injury,' is difficult to achieve in a laboratory environment. Inherent dangers, associated with water analysis and research laboratories where hazardous samples, materials, and equipment are used, must be minimized to protect workers, buildings, and equipment. Managers, supervisors, analysts, and laboratory support personnel each have specific responsibilities to reduce hazards by maintaining a safe work environment. General rules of conduct and safety practices that involve personal protection, laboratory practices, chemical handling, compressed gases handling, use of equipment, and overall security must be practiced by everyone at all levels. Routine and extensive inspections of all laboratories must be made regularly by qualified people. Personnel should be trained thoroughly and repetitively. Special hazards that may involve exposure to carcinogens, cryogenics, or radiation must be given special attention, and specific rules and operational procedures must be established to deal with them. Safety data, reference materials, and texts must be kept available if prudent safety is to be practiced and accidents prevented or minimized.

Study of industry safety management

International Nuclear Information System (INIS)

Park, Pil Su

1987-06-01

This book deals with general remarks, industrial accidents, statistics of industrial accidents, unsafe actions, making machinery and facilities safe, safe activities, having working environment safe, survey of industrial accidents and analysis of causes, system of safety management and operations, safety management planning, safety education, human engineering such as human-machines system, system safety, and costs of disaster losses. It lastly adds individual protective equipment and working clothes including protect equipment for eyes, face, hands, arms and feet.

Advances in industrial ergonomics and safety II

Energy Technology Data Exchange (ETDEWEB)

Das, B [ed.; Technical University of Nova Scotia, Halifax, NS (Canada). Dept. of Industrial Engineering

1990-01-01

135 papers were presented at the conference in 20 sessions with the following headings: aging and industrial performance; back injury and rehabilitation; bioinstrumentation and electromyography; cumulative trauma disorders; engineering anthropometry; equipment design and ergonomics; human computer interaction; human performance and worker satisfaction; human strength and testing; industrial accidents and prevention; industrial biomechanics; injuries in health care; manual materials handling; noise and vibration effects; occupational health and safety; robotics and agricultural machinery safety; statistics and modelling in ergonomics; work environment; workplace safety analysis; and workstation design. Papers are included entitled: A model for analyzing mining machine illumination systems' by R.L. Unger, A.F. Glowacki and E.W. Rossi, 'Ergonomic design guidelines for underground coal mining equipment by E.J. Conway and R. Unger, and Hot work environment and human strain - a relation proposed by K. Bhattacharya and S. Raja.

Advances in industrial ergonomics and safety II

Energy Technology Data Exchange (ETDEWEB)

Das, B. (ed.) (Technical University of Nova Scotia, Halifax, NS (Canada). Dept. of Industrial Engineering)

1990-01-01

135 papers were presented at the conference in 20 sessions with the following headings: aging and industrial performance; back injury and rehabilitation; bioinstrumentation and electromyography; cumulative trauma disorders; engineering anthropometry; equipment design and ergonomics; human computer interaction; human performance and worker satisfaction; human strength and testing; industrial accidents and prevention; industrial biomechanics; injuries in health care; manual materials handling; noise and vibration effects; occupational health and safety; robotics and agricultural machinery safety; statistics and modelling in ergonomics; work environment; workplace safety analysis; and workstation design. Papers are included entitled: A model for analyzing mining machine illumination systems' by R.L. Unger, A.F. Glowacki and E.W. Rossi, 'Ergonomic design guidelines for underground coal mining equipment by E.J. Conway and R. Unger, and Hot work environment and human strain - a relation proposed by K. Bhattacharya and S. Raja.

Cryogenic safety organisation at CERN

CERN Multimedia

CERN. Geneva

2016-01-01

With Safety being a top priority of CERN’s general policy, the Organisation defines and implements a Policy that sets out the general principles governing Safety at CERN. To the end of the attainment of said Safety objectives, the organic units (owners/users of the equipment) are assigned the responsibility for the implementation of the CERN Safety Policy at all levels of the organization, whereas the Health and Safety and Environmental Protection Unit (HSE) has the role of providing assistance for the implementation of the Safety Policy, and a monitoring role related to the implementation of continuous improvement of Safety, compliance with the Safety Rules and the handling of emergency situations. This talk will elaborate on the roles, responsibilities and organisational structure of the different stakeholders within the Organization with regards to Safety, and in particular to cryogenic safety. The roles of actors of particular importance such as the Cryogenic Safety Officers (CSOs) and the Cryogenic Sa...

Effects of organizational safety practices and perceived safety climate on PPE usage, engineering controls, and adverse events involving liquid antineoplastic drugs among nurses.

Science.gov (United States)

DeJoy, David M; Smith, Todd D; Woldu, Henok; Dyal, Mari-Amanda; Steege, Andrea L; Boiano, James M

2017-07-01

Antineoplastic drugs pose risks to the healthcare workers who handle them. This fact notwithstanding, adherence to safe handling guidelines remains inconsistent and often poor. This study examined the effects of pertinent organizational safety practices and perceived safety climate on the use of personal protective equipment, engineering controls, and adverse events (spill/leak or skin contact) involving liquid antineoplastic drugs. Data for this study came from the 2011 National Institute for Occupational Safety and Health (NIOSH) Health and Safety Practices Survey of Healthcare Workers which included a sample of approximately 1,800 nurses who had administered liquid antineoplastic drugs during the past seven days. Regression modeling was used to examine predictors of personal protective equipment use, engineering controls, and adverse events involving antineoplastic drugs. Approximately 14% of nurses reported experiencing an adverse event while administering antineoplastic drugs during the previous week. Usage of recommended engineering controls and personal protective equipment was quite variable. Usage of both was better in non-profit and government settings, when workers were more familiar with safe handling guidelines, and when perceived management commitment to safety was higher. Usage was poorer in the absence of specific safety handling procedures. The odds of adverse events increased with number of antineoplastic drugs treatments and when antineoplastic drugs were administered more days of the week. The odds of such events were significantly lower when the use of engineering controls and personal protective equipment was greater and when more precautionary measures were in place. Greater levels of management commitment to safety and perceived risk were also related to lower odds of adverse events. These results point to the value of implementing a comprehensive health and safety program that utilizes available hazard controls and effectively communicates

30 CFR 70.305 - Respiratory equipment; gas, dusts, fumes, or mists.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Respiratory equipment; gas, dusts, fumes, or... LABOR COAL MINE SAFETY AND HEALTH MANDATORY HEALTH STANDARDS-UNDERGROUND COAL MINES Respiratory Equipment § 70.305 Respiratory equipment; gas, dusts, fumes, or mists. Respiratory equipment approved by...

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

30 CFR 75.1907 - Diesel-powered equipment intended for use in underground coal mines.

Science.gov (United States)

2010-07-01

... underground coal mines. 75.1907 Section 75.1907 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Diesel-Powered Equipment § 75.1907 Diesel-powered equipment intended for use in underground coal mines. (a) As of...

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

Days individual equipment of protection and professional risks

International Nuclear Information System (INIS)

2007-01-01

The personal protection equipment is studied in the legal way (legal liabilities, certification, European texts), technical way (ergonomics, painfulness of ventilated equipment wearing, reliability of a respirable air line, protection gloves against the chemical risk, exposure to nano particulates, working in hot area), human factors (hostile area and emotion management), studies on personal equipment such evaluation, efficiency, conception of new equipment, physiological tolerance, limit of use, and some general safety studies on the working places. (N.C.)

Construction Site Workers’ Awareness on Using Safety Equipment: Case Study

Directory of Open Access Journals (Sweden)

Ulang N. Md

2014-01-01

Full Text Available Construction sector is an important sector and contributed significantly to national development. However, this sector poses higher risk to accident. This is due to fact that construction site can be considered as a dangerous zone to workers and to the public. Due to the variety of cases occurs on site, the contractor will usually have to pay the cost related to accidents in the form of higher insurance premium. Despite various measures, accidents still occur at construction sites. Personal Protective Equipment (PPE is one of the important means to protect the wearer from hazards in the workplace. Since this equipment is the last frontier of the wearer from worksite hazards, it is important to select it based on the job scope and the intended protection. Therefore, this study was formulated to find out the level of knowledge and awareness of construction workers on PPE usage. It was also important to know what make the workers would want or do not want to use the PPE. It was found in this study that the level of awareness and knowledge among workers on the proper use of PPE is moderate. Construction sites accident can be further be reduced with proper implementation of PPE voluntarily by all workers.

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

A DDC Bibliography on Microfiche, Microfilm and Related Equipment, Volume 1.

Science.gov (United States)

Defense Documentation Center, Alexandria, VA.

This bibliography contains abstracts of 40 unclassified-unlimited reports on microfiche, microfilm and related equipment acquired by the Defense Documentation Center since 1953. Citations are topically arranged in AD sequence, ascending order. The following computer produced indexes are provided: subject, corporate author, personal author,…

Equipment Reliability Program in NPP Krsko

International Nuclear Information System (INIS)

Skaler, F.; Djetelic, N.

2006-01-01

Operation that is safe, reliable, effective and acceptable to public is the common message in a mission statement of commercial nuclear power plants (NPPs). To fulfill these goals, nuclear industry, among other areas, has to focus on: 1 Human Performance (HU) and 2 Equipment Reliability (EQ). The performance objective of HU is as follows: The behaviors of all personnel result in safe and reliable station operation. While unwanted human behaviors in operations mostly result directly in the event, the behavior flaws either in the area of maintenance or engineering usually cause decreased equipment reliability. Unsatisfied Human performance leads even the best designed power plants into significant operating events, which can be found as well-known examples in nuclear industry. Equipment reliability is today recognized as the key to success. While the human performance at most NPPs has been improving since the start of WANO / INPO / IAEA evaluations, the open energy market has forced the nuclear plants to reduce production costs and operate more reliably and effectively. The balance between these two (opposite) goals has made equipment reliability even more important for safe, reliable and efficient production. Insisting on on-line operation by ignoring some principles of safety could nowadays in a well-developed safety culture and human performance environment exceed the cost of electricity losses. In last decade the leading USA nuclear companies put a lot of effort to improve equipment reliability primarily based on INPO Equipment Reliability Program AP-913 at their NPP stations. The Equipment Reliability Program is the key program not only for safe and reliable operation, but also for the Life Cycle Management and Aging Management on the way to the nuclear power plant life extension. The purpose of Equipment Reliability process is to identify, organize, integrate and coordinate equipment reliability activities (preventive and predictive maintenance, maintenance

The operator's role and safety functions

International Nuclear Information System (INIS)

Corcoran, W.R.; Finnicum, D.J.; Hubbard, F.R.; Musick, C.R.; Walzer, R.F.

1980-01-01

A nuclear power plant can be thought of as a single system with two major subsystems: equipment and people. Both play important roles in nuclear safety. Whereas, in the past, the role of equipment had been emphasized in nuclear safety, the accident at Three Mile Island and its subsequent investigations point out the vital role of the operator. This paper outlines the operator's roles in nuclear safety and suggests how the concept of safety functions can be used to reduce economic losses and increase safety margins. (auth)

Two NextGen Air Safety Tools: An ADS-B Equipped UAV and a Wake Turbulence Estimator

Science.gov (United States)

Handley, Ward A.

Two air safety tools are developed in the context of the FAA's NextGen program. The first tool addresses the alarming increase in the frequency of near-collisions between manned and unmanned aircraft by equipping a common hobby class UAV with an ADS-B transponder that broadcasts its position, speed, heading and unique identification number to all local air traffic. The second tool estimates and outputs the location of dangerous wake vortex corridors in real time based on the ADS-B data collected and processed using a custom software package developed for this project. The TRansponder based Position Information System (TRAPIS) consists of data packet decoders, an aircraft database, Graphical User Interface (GUI) and the wake vortex extension application. Output from TRAPIS can be visualized in Google Earth and alleviates the problem of pilots being left to imagine where invisible wake vortex corridors are based solely on intuition or verbal warnings from ATC. The result of these two tools is the increased situational awareness, and hence safety, of human pilots in the National Airspace System (NAS).

ACED devices and SECAF supports for the control of structure, pipe network and equipment behaviour at seismic movements in order to enhance the safety margin

International Nuclear Information System (INIS)

Serban, Viorel; Prisecaru, I.; Cretu, D.; Moldoveanu, T.

2002-01-01

In order to enhance the safety margin of structure, pipe networks and equipment associated to the existing NPPs, the classic consolidation solutions are very expensive and many times, impossible to be implemented. Structures, pipe networks, systems and equipment have geometries imposed by the basic construction requirements, operating and safety requirements and their modifications is not always possible. In order to enhance the strength capacity of (new or old) structures, systems and equipment mechanical devices with controlled elasticity and damping (ACED) have been designed, constructed and experimented. These devices are capable to support very large static loads over which dynamic loads (shock, vibration and seismic movements) overlap (which are damped). To increase the strength capacity of (new or existing) pipe networks and equipment connecting with pipes, SECAF supports that allow displacements from thermal expansions with low reaction force have been designed, constructed and experimented. SECAF supports are capable elastically to take permanent loads over which shocks, vibrations and seismic movements (which are damp) overlap. ACED devices and SECAF supports can be used to rehabilitate the existing NPPs with law financial costs and an increase of their strength capacity up to 100% under seismic movements, shocks and vibrations. ACED devices and SECAF supports do not require maintenance, are not affected by presence of a radiation field and their estimated service-life is similar to the NPPs

Conduct of Operations at Nuclear Power Plants. Safety Guide (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide identifies the main responsibilities and practices of nuclear power plant (NPP) operations departments in relation to their responsibility for the safe functioning of the plant. The guide presents the factors to be considered in structuring the operations department of an NPP; setting high standards of performance; making safety related decisions in an effective manner; conducting control room and field activities in a thorough and professional manner; and maintaining an NPP within established operational limits and conditions. Contents: 1. Introduction; 2. Management and organization of plant operations; 3. Shift complement and functions; 4. Shift routines and operating practices; 5. Control of equipment and plant status; 6. Operations equipment and operator aids; 7. Work control and authorization.

Impact of power uprate on environmental qualification of equipment in nuclear power plants

International Nuclear Information System (INIS)

Raheja, R.D.; Mohiuddin, A.; Alsammarae, A.

1996-01-01

Many nuclear power facilities are finding it economically beneficial to increase reactor output, from operating plants, by resorting to power uprates. A power uprate implies that a utility can increase the reactor output, or the megawatts generated, by increasing steam pressure without adding or changing any plant systems. This is perhaps one of the least expensive options for increasing the generating capacity of a power plant. However, a nuclear plant requires a comprehensive review of the plant systems, structures and components to assure their capability to withstand the resulting increased normal and accident plant conditions. A power uprate will typically result in a plant operating at higher than the originally designed environmental conditions. Safety related equipment in nuclear plants is presently qualified to the UFSAR Chapter 15 accident events and the resulting temperatures, pressures, radiation levels etc. These values will increase when the reactor is producing a higher MWe output. Components that are sensitive to the environment must be re-evaluated and assessed to determine their acceptability and operability under the revised environmental conditions. Most safety-related mechanical and electrical equipment will require an assessment from an environmental qualification standpoint. Utilities must perform this task in a systematic, auditable and cost effective manner to optimize their resources and minimize plant costs associated with modifications, replacements or equipment testing. This paper discusses various approaches and provides recommendations to achieve equipment qualification while satisfying the plant's objective of a power uprate

New Paradigm in Nuclear Safety from Quality Assurance to Safety Management System

International Nuclear Information System (INIS)

Lim, Nam-Jin; Park, Chan-Gook; Nam, Ji-Hee; Kim, Kwan-Hyun; Kwon, Hyuk-il; Lee, Young-Gun Lee

2006-01-01

The initial concept of Quality Control (QC) controlling the quality of products is now evolving toward the Management System (MS) achieving safety, through Quality Assurance (QA) ensuring the quality of products and Quality Management (QM) managing the quality by a systematic approach. Nuclear safety can be achieved through an integrated MS that ensures the health, environmental, security, quality and economic requirements being considered together with nuclear safety requirements. MS approach is developed through realizing that most of nuclear accidents had occurred not by the malfunction of hardware or equipment, but by the human error. The MS is a set of inter-related or interacting elements (system) that establishes policies and objectives and which enables those objectives to be achieved in an efficient and effective way

30 CFR 75.1107-11 - Extinguishing agents; requirements on mining equipment employed in low coal.

Science.gov (United States)

2010-07-01

... equipment employed in low coal. 75.1107-11 Section 75.1107-11 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES... § 75.1107-11 Extinguishing agents; requirements on mining equipment employed in low coal. On mining...

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

Hydrogen and Fuel Cells for IT Equipment

Energy Technology Data Exchange (ETDEWEB)

Kurtz, Jennifer

2016-03-09

With the increased push for carbon-free and sustainable data centers, data center operators are increasingly looking to renewable energy as a means to approach carbon-free status and be more sustainable. The National Renewable Energy Laboratory (NREL) is a world leader in hydrogen research and already has an elaborate hydrogen infrastructure in place at the Golden, Colorado, state-of-the-art data center and facility. This presentation will discuss hydrogen generation, storage considerations, and safety issues as they relate to hydrogen delivery to fuel cells powering IT equipment.

Intensive care unit nurses' perceptions of safety after a highly specific safety intervention.

Science.gov (United States)

Elder, N C; Brungs, S M; Nagy, M; Kudel, I; Render, M L

2008-02-01

It is unknown if successful changes in specific safety practices in the intensive care unit (ICU) generalize to broader concepts of patient safety by staff nurses. To explore perceptions of patient safety among nursing staff in ICUs following participation in a safety project that decreased hospital acquired infections. After implementation of practices that reduced catheter-related bloodstream infections in ICUs at four community hospitals, ICU nurses participated in focus groups to discuss patient safety. Audiotapes from the focus groups were transcribed, and two independent reviewers categorised the data which were triangulated with responses from selected questions of safety climate surveys and with the safety checklists used by management leadership on walk rounds. Thirty-three nurses attended eight focus groups; 92 nurses and managers completed safety climate surveys, and three separate leadership checklists were reviewed. In focus groups, nurses predominantly related patient safety to dangers in the physical environment (eg, bed rails, alarms, restraints, equipment, etc.) and to medication administration. These areas also represented 47% of checklist items from leadership walk rounds. Nurses most frequently mentioned self-initiated "double checking" as their main safety task. Focus-group participants and survey responses both noted inconsistency between management's verbal and written commitment compared with their day-to-day support of patient safety issues. ICU nurses who participated in a project to decrease hospital acquired infections did not generalize their experience to other aspects of patient safety or relate it to management's interest in patient safety. These findings are consistent with many adult learning theories, where self-initiated tasks, combined with immediate, but temporary problem-solving, are stronger learning forces than management-led activities with delayed feedback.

Standard practice for qualification of journeyman painters for application of coatings to steel surfaces of safety-related areas in nuclear facilities

International Nuclear Information System (INIS)

Anon.

1984-01-01

This practice provides a standard qualifying method for journeyman painters to verify their proficiency and ability to attain the required quality for application of specified coatings to steel surfaces in safety-related areas in a nuclear facility. Variations or simplifications of the practice set forth herein may be appropriate for special coating work such as maintenance or qualifications of equipment suppliers shop personnel. It is not the intent of this practice to mandate a singular basis for all qualifications. Evaluation of the journeyman painter being qualified in accordance with this practice, shall be by qualified agents as specified in 4.1. Reports shall be prepared as specified in Section 5, and certification as specified in Section 6. It is the intent of this practice to judge only the ability of the journeyman painter to apply specified coatings with the proper tools and equipment. This standard may involve hazardous materials, operations, and equipment

49 CFR 238.229 - Safety appliances-general.

Science.gov (United States)

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false Safety appliances-general. 238.229 Section 238.229... Equipment § 238.229 Safety appliances—general. (a) Except as provided in this part, all passenger equipment continues to be subject to the safety appliance requirements contained in Federal statute at 49 U.S.C...

Fire, safety and ventilation

Energy Technology Data Exchange (ETDEWEB)

Hindle, D.

1999-02-01

Correct ventilation in tunnel environments is vital for the comfort and safety of the people passing through. This article gives details of products from several manufacturers of safety rescue and fire fighting equipment, fire and fume detection equipment, special fire resistant materials, fire resistant hydraulic oils and fire dampers, and ventilation systems. Company addresses and fax numbers are supplied. 4 refs., 5 tabs., 10 photos.

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

Directory of Open Access Journals (Sweden)

Kirrilly Thompson

2015-07-01

Full Text Available 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 in horse-related safety initiatives.

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

2015-07-22

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 in horse-related safety initiatives.

40 CFR 68.48 - Safety information.

Science.gov (United States)

2010-07-01

...) CHEMICAL ACCIDENT PREVENTION PROVISIONS Program 2 Prevention Program § 68.48 Safety information. (a) The... regulated substances, processes, and equipment: (1) Material Safety Data Sheets that meet the requirements...) Equipment specifications; and (5) Codes and standards used to design, build, and operate the process. (b...

The process control and management on equipment qualification of nuclear power plant

International Nuclear Information System (INIS)

Liu Dong; Wang Hongyin; Zhang Yong

2013-01-01

The equipment qualification (EQ) to the safety class equipment is an important safety measure for the nuclear power plants (NPP), and also reflects the nuclear safety culture. Along with the continuous constructions of NPP in China, it has become an important issue for NPP engineering company and equipment suppliers how to effectively establish standard EQ process control and management, and provide sufficient technical arrangements to maintain this EQ management system. This paper summarizes three process of EQ including Design Input, EQ Establishment and EQ Maintenance, proposes the measures and key points for EQ process control and management in phase of NPP construction, and introduces the documents management during the whole process of EQ. (authors)

Design Modification of Kori Unit 1 for the Equipment Qualification

International Nuclear Information System (INIS)

Park, J. W.; Kim, M. Y.; Han, K. T.; Park, J. D.

2007-01-01

There has not been a strict regulatory requirements for the Equipment Qualification(EQ) in 1970's when Kori Unit 1 had begun the construction and the commercial operation. The Korean regulatory body requested the EQ on the various safety-related components, as a result of Periodic Safety Review. However, the EQ itself is impossible in some areas, due to the high pressure/temperature and flooding environment conditions from the pipe breaks. Design modification is being considered in the Auxiliary Building, the Intermediate Building, the Component Cooling Water Heat Exchanger Building and the Turbine Building, in order to mitigate the environmental conditions for the EQ

IE Bulletin No. 79-01: Environmental qualification of Class IE equipment

International Nuclear Information System (INIS)

Anon.

1993-01-01

The intent of IE Circular 78-08 was to highlight to all licensees important lessons learned from environmental qualification deficiencies reported by individual licensees. In this regard, licensees were requested to examine installed safety-related electrical equipment and determine that proper documentation existed which provided assurance that this equipment would function under postulated accident conditions. The intent of this Bulletin is to raise the threshold of IE Circular 78-08 to the level of a Bulletin; i.e., action requiring a licensee response. Inspections conducted to date by the NRC of licensees' activities in response to IE Circular 78-08 have identified one component which licensees have found to be unqualified for service within the LOCA environment. Specifically, unqualified stem mounted limit switches (SMLS), other than those identified in previously issued IE Bulletin 78-04, were found to be installed on safety-related valves inside containment at both Duane Arnold and Quad Cities 1 and 2 Nuclear Generating Stations. According to the manufacturer, these switches are designed only for general purpose applications and are not considered suitable devices for service in the LOCA environment. Consequently, switches are being replaced at the above power plants with qualified components

SOLUTION FOR AN EQUIPMENT OF RAILWAY BRIDGE MONITORING

Directory of Open Access Journals (Sweden)

Liviu Mihai SIMA

2009-11-01

Full Text Available This paper presents a solution for the railway bridges monitoring based on the use of already existing equipments in the railway installations and some new equipments based on photo-elements, as well. This kind of equipments can function in continue and discontinuous regime, means of them tracking down if the bridge piles suffered movements by different reasons. This aspect can put in danger the safety of the circulation on that bridge.

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

Aging of turbine drives for safety-related pumps in nuclear power plants

International Nuclear Information System (INIS)

Cox, D.F.

1995-06-01

This study was 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 plants and in the Reactor Core Isolation Cooling and High-Pressure Coolant Injection systems for boiling-water reactor plants. This research has been conducted by examination of failure data in the Nuclear Plant Reliability Data System, review of Licensee Event Reports, discussion of problems with operating plant personnel, and personal observation. The reported failure data were reviewed to determine the cause of the event and the method of discovery. Based on the research results, attempts have been made to determine the predictability of failures and possible preventive measures that may be implemented. Findings in a recent study of AFW systems indicate that the turbine drive is the single largest contributor to AFW system degradation. However, examination of the data shows that the turbine itself is a reliable piece of equipment with a good service record. Most of the problems documented are the result of problems with the turbine controls and the mechanical overspeed trip mechanism; these apparently stem from three major causes which are discussed in the text. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design, have led to improved performance resulting in a reliable safety-related component. However, these improvements have not been universally implemented

Design guides for radioactive-material-handling facilities and equipment

International Nuclear Information System (INIS)

Doman, D.R.; Barker, R.E.

1980-01-01

Fourteen key areas relating to facilities and equipment for handling radioactive materials involved in examination, reprocessing, fusion fuel handling and remote maintenance have been defined and writing groups established to prepare design guides for each areas. The guides will give guidance applicable to design, construction, operation, maintenance and safety, together with examples and checklists. Each guide will be reviewed by an independent review group. The guides are expected to be compiled and published as a single document

The Safety Prevention in the Theater Management and Operation

Institute of Scientific and Technical Information of China (English)

WU Sheng

2015-01-01

Take the operation and management experience as examples, the author discussed how to formulate a set of complete and effective equipment management system, operating rules, procedures and standards, as well as the safety prevention and control measures, according to the national or trade related laws and regulations and combining the operation and performance characteristics of theatre management, in order to ensure the safe operation of theatre and stage equipment.

49 CFR 374.313 - Equipment.

Science.gov (United States)

2010-10-01

.... Each bus (except in commuter service) seating more than 14 passengers (not including the driver) shall..., DEPARTMENT OF TRANSPORTATION FEDERAL MOTOR CARRIER SAFETY REGULATIONS PASSENGER CARRIER REGULATIONS Adequacy of Intercity Motor Common Carrier Passenger Service § 374.313 Equipment. (a) Temperature control. A...

Procuring, maintaining and using equipment: the key issues in controlling exposure to patients

International Nuclear Information System (INIS)

Hendra, I.R.F.

1989-01-01

Not only are acceptance criteria directly related to the design of medical x-ray equipment and quality assurance exercised during its manufacture, but also its performance, safety and reliability must feature in requirements for procuring new equipment. In the UK this issue is dealt with under the Department of Health and Social Security Manufacturer Registration Scheme for Medical Equipment in its application to radiological equipment. In the UK quality of maintenance services has been considered by one of our Working Parties commissioned by the South-East Thames Regional Health Authority who are the ''Centre of Responsibility'' for diagnostic imaging equipment in the National Health Service. The report prepared from the first phase of activity of this working group describes features of a procurement specification for maintenance services and prescribes complementary duties of contractors and customers. Radiograph reject analysis has indicated human error outweighs unsatisfactory equipment performance as the reason for rejecting poor radiographs. The use in X-ray departments of the kind of manufacturing quality system used by most reputable equipment manufacturers has the potential for a dramatic reduction in exposure. (author)

Similarity principles for equipment qualification by experience

International Nuclear Information System (INIS)

Kana, D.D.; Pomerening, D.J.

1988-07-01

A methodology is developed for seismic qualification of nuclear plant equipment by applying similarity principles to existing experience data. Experience data are available from previous qualifications by analysis or testing, or from actual earthquake events. Similarity principles are defined in terms of excitation, equipment physical characteristics, and equipment response. Physical similarity is further defined in terms of a critical transfer function for response at a location on a primary structure, whose response can be assumed directly related to ultimate fragility of the item under elevated levels of excitation. Procedures are developed for combining experience data into composite specifications for qualification of equipment that can be shown to be physically similar to the reference equipment. Other procedures are developed for extending qualifications beyond the original specifications under certain conditions. Some examples for application of the procedures and verification of them are given for certain cases that can be approximated by a two degree of freedom simple primary/secondary system. Other examples are based on use of actual test data available from previous qualifications. Relationships of the developments with other previously-published methods are discussed. The developments are intended to elaborate on the rather broad revised guidelines developed by the IEEE 344 Standards Committee for equipment qualification in new nuclear plants. However, the results also contribute to filling a gap that exists between the IEEE 344 methodology and that previously developed by the Seismic Qualification Utilities Group. The relationship of the results to safety margin methodology is also discussed. (author)

Occupational Safety and Health Concerns in Logging: A Cross-Sectional Assessment in Virginia

Directory of Open Access Journals (Sweden)

Sunwook Kim

2017-11-01

Full Text Available Increased logging mechanization has helped improve logging safety and health, yet related safety risks and concerns are not well understood. A cross-sectional study was completed among Virginia loggers. Participants (n = 122 completed a self-administered questionnaire focusing on aspects of safety and health related to logging equipment. Respondents were at a high risk of workplace injuries, with reported career and 12-month injury prevalences of 51% and 14%, respectively. Further, nearly all (98% respondents reported experiencing musculoskeletal symptoms. Over half (57.4% of respondents reported symptoms related to diesel exhaust exposure in their career. Few (15.6%, however, perceived their jobs to be dangerous. Based on the opinions and suggestions of respondents, three priority areas were identified for interventions: struck-by/against hazards, situational awareness (SA during logging operations, and visibility hazards. To address these hazards, and to have a broader and more substantial positive impact on safety and health, we discuss the need for proactive approaches such as incorporating proximity technologies in a logging machine or personal equipment, and enhancing logging machine design to enhance safety, ergonomics, and SA.

Safety technology: Emergency equipment and sea rescue techniques

Energy Technology Data Exchange (ETDEWEB)

1986-06-01

In June 1986, a two day workshop was convened to review mobile offshore drilling unit (MODU) evacuation research and development (R D) requirements for programs currently sponsored by the Marine Engineering Committee for PERD (Panel on Energy Research and Development) task 6.2. The proceedings of the workshop are presented in terms of: evacuation technology needs; a review of current R D projects for their relevance to the technological needs; recommended changes to current projects where warranted; and recommended priority R D projects that are not presently being undertaken to meet the evacuation technology needs. Current R D projects reviewed include: preferred orientation and displacement (PROD) lifeboat launching, personal transfer baskets, arctic escape system (AES), fast rescue craft (FRC), helicopter survival suits, diving bell emergency heaters, working immersion suits, diving bell emergency ascent system, undersea robotics systems and DCIEM decompression tables. It was concluded that gaps in offshore safety/evacuation capability cannot be adequately addressed by one or two major projects. Rather, a wider range of problems exist that require attention through a greater financial commitment than is presently the case. Government-industry cooperation in this field is required. Overall, the projects were judged to have been well planned, managed, and relevant to offshore safety needs. The only exception was a feeling that diving related projects consummed a disproportionate share of the available funding. Further, there was some reservation that the robotics study was outside the 6.2 PERD R D mandate. A list of recommendations is presented. 1 tab.

Method and procedure of fatigue analysis for nuclear equipment

International Nuclear Information System (INIS)

Wen Jing; Fang Yonggang; Lu Yan; Zhang Yue; Sun Zaozhan; Zou Mingzhong

2014-01-01

As an example, the fatigue analysis for the upper head of the pressurizer in one NPP was carried out by using ANSYS, a finite element method analysis software. According to RCC-M code, only two kinds of typical transients of temperature and pressure were considered in the fatigue analysis. Meanwhile, the influence of earthquake was taken into account. The method and procedure of fatigue analysis for nuclear safety equipment were described in detail. This paper provides a reference for fatigue analysis and assessment of nuclear safety grade equipment and pipe. (authors)

Improved Management of Part Safety Classification System for Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Young; Park, Youn Won; Park, Heung Gyu; Park, Hyo Chan [BEES Inc., Daejeon (Korea, Republic of)

2016-10-15

As, in recent years, many quality assurance (QA) related incidents, such as falsely-certified parts and forged documentation, etc., were reported in association with the supply of structures, systems, components and parts to nuclear power plants, a need for a better management of safety classification system was addressed so that it would be based more on the level of parts . Presently, the Korean nuclear power plants do not develop and apply relevant procedures for safety classifications, but rather the safety classes of parts are determined solely based on the experience of equipment designers. So proposed in this paper is a better management plan for safety equipment classification system with an aim to strengthen the quality management for parts. The plan was developed through the analysis of newly introduced technical criteria to be applied to parts of nuclear power plant.

30 CFR 75.1310 - Explosives and blasting equipment.

Science.gov (United States)

2010-07-01

... for use so long as the present approval is maintained. (e) Electric detonators shall be compatible... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Explosives and blasting equipment. 75.1310... SAFETY AND HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Explosives and Blasting § 75.1310...

Personal protective equipment

International Nuclear Information System (INIS)

2004-01-01

This Practical Radiation Technical Manual is one of a series that has been designed to provide guidance on radiological protection for employers, radiation protection officers, managers and other technically competent persons who have responsibility for ensuring the safety of employees working with ionizing radiation. The Manual may be used with the appropriate IAEA Practical Radiation Safety Manuals to provide training, instruction and information for all employees engaged in work with ionizing radiation. Personal protective equipment (PPE) includes clothing or other special equipment that is issued to individual workers to provide protection against actual or potential exposure to ionizing radiations. It is used to protect each worker against the prevailing risk of external or internal exposure in circumstances in which it is not reasonably practicable to provide complete protection by means of engineering controls or administrative methods. Adequate personal protection depends on PPE being correctly selected, fitted and maintained. Appropriate training for the users and arrangements to monitor usage are also necessary to ensure that PPE provides the intended degree of protection effectively. This Manual explains the principal types of PPE, including protective clothing and respiratory protective equipment (RPE). Examples of working procedures are also described to indicate how PPE should be used within a safe system of work. The Manual will be of most benefit if it forms part of a more comprehensive training programme or is supplemented by the advice of a qualified expert in radiation protection. Some of the RPE described in this Manual should be used under the guidance of a qualified expert

Pediatric interventional radiography equipment: safety considerations

International Nuclear Information System (INIS)

Strauss, Keith J.

2006-01-01

This paper discusses pediatric image quality and radiation dose considerations in state-of-the-art fluoroscopic imaging equipment. Although most fluoroscopes are capable of automatically providing good image quality on infants, toddlers, and small children, excessive radiation dose levels can result from design deficiencies of the imaging device or inappropriate configuration of the equipment's capabilities when imaging small body parts. Important design features and setup choices at installation and during the clinical use of the imaging device can improve image quality and reduce radiation exposure levels in pediatric patients. Pediatric radiologists and cardiologists, with the help of medical physicists, need to understand the issues involved in creating good image quality at reasonable pediatric patient doses. The control of radiographic technique factors by the generator of the imaging device must provide a large dynamic range of mAs values per exposure pulse during both fluoroscopy and image recording as a function of patient girth, which is the thickness of the patient in the posterior-anterior projection at the umbilicus (less than 10 cm to greater than 30 cm). The range of pulse widths must be limited to less than 10 ms in children to properly freeze patient motion. Variable rate pulsed fluoroscopy can be leveraged to reduce radiation dose to the patient and improve image quality. Three focal spots with nominal sizes of 0.3 mm to 1 mm are necessary on the pediatric unit. A second, lateral imaging plane might be necessary because of the child's limited tolerance of contrast medium. Spectral and spatial beam shaping can improve image quality while reducing the radiation dose. Finally, the level of entrance exposure to the image receptor of the fluoroscope as a function of operator choices, of added filter thickness, of selected pulse rate, of the selected field-of-view and of the patient girth all must be addressed at installation. (orig.)

Review of safety related control room function research based on experience from nuclear power plants in Finland

International Nuclear Information System (INIS)

Juslin, K.; Wahlstroem, B.; Rinttilae, E.

1985-01-01

A comprehensive human engineering research programme was established in the second half of the 1970's at the Technical Research Centre of Finland (VTT). The research is performed in cooperation with the utility companies Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (TVO) and includes topics such as Handling of alarm information, Disturbance analysis systems, Assessment of control rooms and Validation of safety parameter display systems. Reference is also made to the Finnish contribution to the OECD Halden Reactor Project (Halden) and the Nordic Liaison Committee for Atomic Energy (NKA) research projects. In this paper feasible realization alternatives of safety related control room functions are discussed on the basis of experience from the nuclear power plants in Finland, which at present are equipped with extensive process computer systems. A proposal for future power plant information systems is described. It is intended that this proposal will serve as the basis for future computer systems at nuclear power plants in Finland. (author)

Seismic qualification of equipment

International Nuclear Information System (INIS)

Heidebrecht, A.C.; Tso, W.K.

1983-03-01

This report describes the results of an investigation into the seismic qualification of equipment located in CANDU nuclear power plants. It is particularly concerned with the evaluation of current seismic qualification requirements, the development of a suitable methodology for the seismic qualification of safety systems, and the evaluation of seismic qualification analysis and testing procedures

Basic fracture toughness requirements for ferritic materials of nuclear class pressure retaining equipment in NPP

International Nuclear Information System (INIS)

Ning Dong; Yao Weida

2005-01-01

In this paper, theory basis on cold brittleness and anti-brittle fracture design of ferritic materials are introduced summarily and fracture toughness requirements for ferritic materials in ASME code for nuclear safety class pressure retaining equipment in NPP are summarized and evaluated. The results show that notch impact toughness requirements for materials relate to nuclear safety class of materials so as to ensure that brittle fracture of retaining pressure boundary in NPP can not occur. (authors)

Fast reactor safety program. Progress report, January-March 1980

International Nuclear Information System (INIS)

1980-05-01

The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development

Department of Energy Construction Safety Reference Guide

Energy Technology Data Exchange (ETDEWEB)

1993-09-01

DOE has adopted the Occupational Safety and Health Administration (OSHA) regulations Title 29 Code of Federal Regulations (CFR) 1926 ``Safety and Health Regulations for Construction,`` and related parts of 29 CFR 1910, ``Occupational Safety and Health Standards.`` This nonmandatory reference guide is based on these OSHA regulations and, where appropriate, incorporates additional standards, codes, directives, and work practices that are recognized and accepted by DOE and the construction industry. It covers excavation, scaffolding, electricity, fire, signs/barricades, cranes/hoists/conveyors, hand and power tools, concrete/masonry, stairways/ladders, welding/cutting, motor vehicles/mechanical equipment, demolition, materials, blasting, steel erection, etc.

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...

NASA safety standard for lifting devices and equipment

Science.gov (United States)

1990-09-01

NASA's minimum safety requirements are established for the design, testing, inspection, maintenance, certification, and use of overhead and gantry cranes (including top running monorail, underhung, and jib cranes), mobile cranes, derrick hoists, and special hoist supported personnel lifting devices (these do not include elevators, ground supported personnel lifts, or powered platforms). Minimum requirements are also addressed for the testing, inspection, and use of Hydra-sets, hooks, and slings. Safety standards are thoroughly detailed.

Safety aspects in radiology

International Nuclear Information System (INIS)

Silva, D.C. da.

1991-05-01

The development of a program for the evaluation of the physical installations and operational procedures in diagnostic radiology with respect to radiation-safety is described. In addition, a proposal for the quality analysis of X-ray equipment and film-processing is presented. The purpose is both to ensure quality and safety of the radiology service, as well as to aid in the initial and in-service training of the staff. Interviews with patients, staff practicing radiology at a wide range of levels and the controlling authorities were carried out in the State of Rio de Janeiro in order to investigate the existence and the effective use of personal radioprotection equipment as well as user's and staff's concern for radiation safety. Additionally physical measurements were carried out in University Hospitals in Rio de Janeiro to assess the quality of equipment in day-to-day use. It was found that in the locations which did not have routine maintenance the equipment was generally in a poor state which lead to a high incidence of repetition of examinations and the consequent financial loss. (author)

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 evaluation for packaging transportation of equipment for tank 241-C-106 waste sluicing system

International Nuclear Information System (INIS)

Calmus, D.B.

1994-01-01

A Waste Sluicing System (WSS) is scheduled for installation in nd waste storage tank 241-C-106 (106-C). The WSS will transfer high rating sludge from single shell tank 106-C to double shell waste tank 241-AY-102 (102-AY). Prior to installation of the WSS, a heel pump and a transfer pump will be removed from tank 106-C and an agitator pump will be removed from tank 102-AY. Special flexible receivers will be used to contain the pumps during removal from the tanks. After equipment removal, the flexible receivers will be placed in separate containers (packagings). The packaging and contents (packages) will be transferred from the Tank Farms to the Central Waste Complex (CWC) for interim storage and then to T Plant for evaluation and processing for final disposition. Two sizes of packagings will be provided for transferring the equipment from the Tank Farms to the interim storage facility. The packagings will be designated as the WSSP-1 and WSSP-2 packagings throughout the remainder of this Safety Evaluation for Packaging (SEP). The WSSP-1 packagings will transport the heel and transfer pumps from 106-C and the WSSP-2 packaging will transport the agitator pump from 102-AY. The WSSP-1 and WSSP-2 packagings are similar except for the length

Development of fragility descriptions of equipment for seismic risk assessment of nuclear power plants

International Nuclear Information System (INIS)

Hardy, G.S.; Campbell, R.D.

1983-01-01

Probabilistic risk assessment (PRA) of a nuclear power plant for postulated hazard requires the development of fragility relationships for the plants' safety related equipment. The objective of this paper is to present some general results and conclusions concerning the development of these seismic fragility levels. Participation in fragility-related research and experience gained from the completion of several PRA studies of a variety of nuclear power plants have provided much insight as to the most vulnerable equipment and the most efficient use of resources for development of fragilities. Plants studied had seismic design bases ranging from very simple equivalent static analysis for some of the earlier plants to state-of-the-art complex multimode dyanamic analyses for plants currently under construction. Increased sophistication and rigor in seismic qualification of equipment has resulted for the most part in increased seismic resistance. The majority of equipment has been found, however, to possess more than adequate resistance to seismic loading regardless of the degree of sophistication utilized in design as long as seismic loading was included in the design process. This paper presents conclusions of the authors as to which items of equipment typically require an individual ''plant-specific'' fragility analysis and which can be treated in a generic fashion. In addition, general conclusions on the relative seismic capacity levels and most frequent failure modes are summarized for generic equipment groups

Development of a Consensus Standard for School Equipment: NSF/NSSEA 380

Science.gov (United States)

Breitner, Ashlee

2011-01-01

For many years, the school supplies and equipment industry has investigated methods to ensure product safety and compliance across all its product categories. In early 2010, NSF International and the National School Supply and Equipment Association (NSSEA) came together to develop quality standards for products and equipment designed for use in…

Fire safety requirements for electrical cables towards nuclear reactor safety

International Nuclear Information System (INIS)

Raju, M.R.

2002-01-01

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

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....

Safety analysis of the present status of the research reactor 'RA' at 'Vinca' Institute

International Nuclear Information System (INIS)

Jovic, V.; Jovic, L.; Zivotic, Z.; Milovanovic, Dj.

1995-01-01

Safety analysis of the nuclear facility which has been out of work for a long time and whose future is not defined at the present moment, can not be connected to the usual, normatively regulated system analysis procedure in both operational and accidental regimes. Therefore, the safety analysis of the present status of the present status of the reactor RA is related to system and components analysis which, in present conditions maintain their nuclear functions operational. In the first place, it refers to components and equipment in which radioactive radiation generation still exists and to installations and equipment maintaining radiation level below permitted limit. in the context of the analysis the following areas are being covered: present status characteristics, accidental events while operating period from 1959. to 1984., nuclear fuels and radioactive waste inventory, basic characteristics and status of safety-related systems and equipment, radiation protection, potential accident analysis at present status of the reactor RA, potential accidental situations due to natural events (earthquakes, water flood) or man-induced events and security. 8 refs

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.

State Office for Nuclear Safety - New Regulatory Body in Croatia

International Nuclear Information System (INIS)

Novosel, N.; Prah, M.; Valcic, I.; Cizmek, A.

2006-01-01

The Act on Nuclear Safety was adopted by the Croatian Parliament on 15 October 2003, and it is published in the Official Gazette No. 173/03. This Act regulates safety and protective measures for using nuclear materials and specified equipment and performing nuclear activities, and establishes the State Office for Nuclear Safety. Provisions of this Act apply on nuclear activities, nuclear materials and specified equipment. Also, by accession to international conventions and agreements, Croatia took the responsibility of implementing the provisions of those international treaties. In the process of European and international integrations, Croatia has to make harmonization with European and international standards also in the field of nuclear safety. The State Office for Nuclear Safety as an independent regulatory authority started its work on 1st June 2005 by taking over responsibility for activities relating to nuclear safety and cooperation with the International Atomic Energy Agency from the Ministry of the Economy, Labour and Entrepreneurship. In this paper responsibilities, organization and projects of the State Office for Nuclear Safety will be presented, with the accent on development of regulations and international cooperation. (author)

Laboratory services series: an electrical outlet and corded equipment inspection program

International Nuclear Information System (INIS)

Davis, E.A.

1976-04-01

A research and development laboratory has thousands of electrical outlets providing power to laboratories, offices, shops, and service areas. These outlets provide power for a wide variety of portable equipment and tools that are equipped with cord and plug. Electric safety requires a periodic check of outlet grounding capability and continuing inspection and repair of corded equipment. Personnel, equipment, reports, procedures, and schedule requirements are reported

Methodology development for availability improvement of standby equipment

International Nuclear Information System (INIS)

Shin, Sung Min; Jeon, In Seop; Kang, Hyun Gook

2014-01-01

The core damage frequency (CDF) of operating and constructing pressurized nuclear plants are ranging on the order of 10 -5 and 10 -6 per year. The target CDF of new NPP design has been set at 10 -7 . In this context, although various systems are currently studied, availability improvement of standby equipment will be more efficient than the additional application of safety systems. It is obvious in every aspect, such as management and cost efficiency. Here, soundness can affect equipment unavailability, and the soundness degrades because of aging. However, some studies did not consider aging when calculating the unavailability. Standby equipment can age because of two important factors: standby stress which accumulates over time, and test stress which accumulates with the number of tests (or operations). Both factors should be considered together when aging is considered. However, some studies only considered standby stress or test stress. There are some previous studies which considered both factors. Besides equipment soundness related to aging effect, some process like bypass during test also can affect equipment unavailability because the original function of equipment cannot be performed immediately during this process. However, there are seldom studies dealing with above factors as a whole problem. This study investigated a general approach to calculate the unavailability of standby equipment which considers aging caused by standby and test stresses and bypass process. Based on this general approach, we propose two maintenance strategies which aim to reduce standby equipment unavailability. In section 2, the general approach is presented. As one of the strategies, the changing test interval method (CIM) is introduced in section 3, and its effectiveness is also analyzed. The online monitoring method (OMM) is investigated in section 4 as another method to reduce equipment unavailability. In section 5, a combination of these two methods is analyzed. A general

Communications received from certain Member States regarding guidelines for transfers of nuclear-related dual-use equipment, materials, software and related technology

International Nuclear Information System (INIS)

2005-01-01

The Director General of the International Atomic Energy Agency has received Notes Verbales dated 25 October 2004 from the Resident Representatives to the Agency of Argentina, Australia, Austria, Belgium, Brazil, Canada, Cyprus, Czech Republic, Estonia, Finland, France, Greece, Hungary, Italy, Japan, Malta, Netherlands, Norway, Republic of Korea, South Africa, Spain, Sweden, Turkey, Ukraine, the United Kingdom of Great Britain and Northern Ireland and the United States of America, relating to transfers of nuclear-related dual-use equipment, materials, software and related technology. The purpose of the Notes Verbales is to provide further information on those Governments' guidelines for transfers of nuclear-related dual-use equipment, materials, software and related technology. In the light of the wish expressed at the end of each Note Verbale, the text of the Notes Verbales is attached. The attachment to these Notes Verbales is also reproduced in full

Communications received from certain Member States regarding guidelines for transfers of nuclear-related dual-use equipment, materials, software and related technology

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-02-23

The Director General of the International Atomic Energy Agency has received Notes Verbales dated 25 October 2004 from the Resident Representatives to the Agency of Argentina, Australia, Austria, Belgium, Brazil, Canada, Cyprus, Czech Republic, Estonia, Finland, France, Greece, Hungary, Italy, Japan, Malta, Netherlands, Norway, Republic of Korea, South Africa, Spain, Sweden, Turkey, Ukraine, the United Kingdom of Great Britain and Northern Ireland and the United States of America, relating to transfers of nuclear-related dual-use equipment, materials, software and related technology. The purpose of the Notes Verbales is to provide further information on those Governments' guidelines for transfers of nuclear-related dual-use equipment, materials, software and related technology. In the light of the wish expressed at the end of each Note Verbale, the text of the Notes Verbales is attached. The attachment to these Notes Verbales is also reproduced in full.

Radiation safety in welding and testing

International Nuclear Information System (INIS)

King, B.E.; Malaxos, M.; Hartley, B.M.

1985-01-01

There are a number of ways of achieving radiation safety in the workplace. The first is by engineering radiation safety into the equipment, providing shielded rooms and safety interlocks. The second is by following safe working procedures. The National Health and Medical Research Council's Code of practice for the control and safe handling of sealed radioactive sources used in industrial radiography (1968) sets out the standards which must be met by equipment to be used in industrial radiography

International Aspects of Nuclear Safety

International Nuclear Information System (INIS)

Lash, T.R.

2000-01-01

Even though not all the world's nations have developed a nuclear power industry, nuclear safety is unquestionably an international issue. Perhaps the most compelling proof is the 1986 accident at Chornobyl nuclear power plant in what is now Ukraine. The U.S. Department of Energy conducts a comprehensive, cooperative effort to reduce risks at Soviet-designed nuclear power plants. In the host countries : Armenia, Ukraine, Russia, Bulgaria, the Czech Republic, Hungary, Lithuania, Slovakia, and Kazakhstan joint projects are correcting major safety deficiencies and establishing nuclear safety infrastructures that will be self-sustaining.The U.S. effort has six primary goals: 1. Operational Safety - Implement the basic elements of operational safety consistent with internationally accepted practices. 2. Training - Improve operator training to internationally accepted standards. 3. Safety Maintenance - Help establish technically effective maintenance programs that can ensure the reliability of safety-related equipment. 4. Safety Systems - Implement safety system improvements consistent with remaining plant lifetimes. 5. Safety Evaluations - Transfer the capability to conduct in-depth plant safety evaluations using internationally accepted methods. 6. Legal and Regulatory Capabilities - Facilitate host-country implementation of necessary laws and regulatory policies consistent with their international treaty obligations governing the safe use of nuclear power

49 CFR 214.521 - Flagging equipment for on-track roadway maintenance machines and hi-rail vehicles.

Science.gov (United States)

2010-10-01

... maintenance machines and hi-rail vehicles. 214.521 Section 214.521 Transportation Other Regulations Relating... WORKPLACE SAFETY On-Track Roadway Maintenance Machines and Hi-Rail Vehicles § 214.521 Flagging equipment for on-track roadway maintenance machines and hi-rail vehicles. Each on-track roadway maintenance machine...

Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

International Nuclear Information System (INIS)

Lukacevic, H.; Kopinc, D.; Ivanjko, M.

2016-01-01

Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

LMZ experience in refurbishment of hydroturbine equipment

Energy Technology Data Exchange (ETDEWEB)

Sotnikov, Anatoly A. [LMZ, St. Petersburg (Russian Federation). Div. of Hydraulic Machine

2000-07-01

AO LMZ experience in refurbishment of hydroturbine equipment is generalized. Hydraulic turbines of many power stations having been in service of more than 30 years need rehabilitation and modernization. As a rule, the following problems are solved in the process of refurbishment works: increase of turbine efficiency and output, ensuring of reliable operation of the equipment during the next length of life, ensuring the environmental safety of the equipment, furnishing of the power station with up to date automatic control systems. The process of refurbishment used by LMZ is described. The examples of refurbishment are given. (author)

Code of Practice for Protection Against Ionizing Radiation Emitted from X-ray Analysis Equipment (1984)

International Nuclear Information System (INIS)

1984-01-01

Appropriate working rules, safety features and monitoring requirements for general X-ray analysis units or equipment are laid down in this Code which is intended for users of such equipment. The Code advises that establishments draw up their own working procedures based on appropriate legislation and on the recommendations contained in this Code. The Code also describes the requirements for X-ray analysis equipment necessary to ensure safety. (NEA) [fr

Cookbooks in U.S. history: How do they reflect food safety from 1896 to 2014?

Science.gov (United States)

Almanza, Barbara A; Byrd, Karen S; Behnke, Carl; Ma, Jing; Ge, Li

2017-09-01

Historical cookbooks as a source of recipes and food preparation information would be expected to document advancements in food safety related to kitchen equipment, cleaning, foodborne illness knowledge, and consumer education materials. In turn, this food safety information might be expected to contribute to consumers' food safety behaviors. Using both quantitative and qualitative research methodology, this study assessed how food safety information in cookbooks changed and how quickly advancements were incorporated. Faster assimilation into cookbooks was associated with kitchen equipment, educational resources (hotlines and websites), and foodborne illness outbreaks. The rate of incorporation of education materials was moderate. Cleaning advances were the slowest to be incorporated. Modern cookbooks published after the 1980's rapidly evolved with advances in food safety knowledge. Copyright © 2017 Elsevier Ltd. All rights reserved.

Rules and procedures for the design and operation of hazardous research equipment

International Nuclear Information System (INIS)

1978-12-01

The manual has been prepared for use by research personnel involved in experiments at the Lawrence Berkeley Laboratory. It contains rules and procedures for the design, test, installation, and operation of hazardous research equipment. Sect. I contains such information as responsibility of experimenters for safety, descriptions of the various Laboratory safety organizations, and enumeration of various services available to experimenters at the Laboratory. Sect. II describes specific rules for the setup and operation of experimental equipment at the Laboratory. Sect. III gives detailed design criteria and procedures for equipment frequently encountered in the high energy physics laboratory

NFC based Equipment Qualification Management (NEQM) system preventing counterfeit and fraudulent item

Energy Technology Data Exchange (ETDEWEB)

Chang, C.K., E-mail: ckchang@kings.ac.kr [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Lee, K.J., E-mail: klee@khu.ac.kr [Kyung Hee Univ., Seoul (Korea, Republic of)

2014-07-01

Qualification of equipment essential to safety in nuclear power plants (NPPs) ensures its capability to perform designated safety functions on demand under postulated service conditions. However, a number of incidents identified by the NRC since 1980s catalysed the US nuclear industry to adopt standard precautions to guard against counterfeit items. The purpose of this paper is to suggest the NFC (Near Field Communication) based equipment qualification management system preventing counterfeit and fraudulent items. The NEQM (NFC based Equipment Qualification Management) system work with the support of legacy systems such as PMS (Procurement Management System) and FMS (Facility management System). (author)

NFC based Equipment Qualification Management (NEQM) system preventing counterfeit and fraudulent item

International Nuclear Information System (INIS)

Chang, C.K.; Lee, K.J.

2014-01-01

Qualification of equipment essential to safety in nuclear power plants (NPPs) ensures its capability to perform designated safety functions on demand under postulated service conditions. However, a number of incidents identified by the NRC since 1980s catalysed the US nuclear industry to adopt standard precautions to guard against counterfeit items. The purpose of this paper is to suggest the NFC (Near Field Communication) based equipment qualification management system preventing counterfeit and fraudulent items. The NEQM (NFC based Equipment Qualification Management) system work with the support of legacy systems such as PMS (Procurement Management System) and FMS (Facility management System). (author)

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...

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

1968 Listing of Swimming Pool Equipment.

Science.gov (United States)

National Sanitation Foundation, Ann Arbor, MI. Testing Lab.

An up-to-date listing of swimming pool equipment including--(1) companies authorized to display the National Sanitation Foundation seal of approval, (2) equipment listed as meeting NSF swimming pool equipment standards relating to diatomite type filters, (3) equipment listed as meeting NSF swimming pool equipment standard relating to sand type…

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.

30 CFR 77.403-1 - Mobile equipment; rollover protective structures (ROPS).

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Mobile equipment; rollover protective structures (ROPS). 77.403-1 Section 77.403-1 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS, SURFACE COAL MINES AND SURFACE...

30 CFR 57.12005 - Protection of power conductors from mobile equipment.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Protection of power conductors from mobile equipment. 57.12005 Section 57.12005 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND...

30 CFR 77.403 - Mobile equipment; falling object protective structures (FOPS).

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Mobile equipment; falling object protective structures (FOPS). 77.403 Section 77.403 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS, SURFACE COAL MINES AND SURFACE WORK...

30 CFR 77.203 - Use of material or equipment overhead; safeguards.

Science.gov (United States)

2010-07-01

...; safeguards. 77.203 Section 77.203 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR COAL MINE SAFETY AND HEALTH MANDATORY SAFETY STANDARDS, SURFACE COAL MINES AND SURFACE WORK AREAS...; safeguards. Where overhead repairs are being made at surface installations and equipment or material is taken...

National Chemistry Teacher Safety Survey

Science.gov (United States)

Plohocki, Barbra A.

This study evaluated the status of secondary school instructional chemistry laboratory safety using a survey instrument which focused on Teacher background Information, Laboratory Safety Equipment, Facility Safety, General Safety, and a Safety Content Knowledge Survey. A fifty question survey instrument based on recent research and questions developed by the researcher was mailed to 500 secondary school chemistry teachers who participated in the 1993 one-week Woodrow Wilson National Fellowship Foundation Chemistry Institute conducted at Princeton University, New Jersey. The data received from 303 respondents was analyzed by t tests and Analysis of Variance (ANOVA). The level of significance for the study was set at ~\\ performance on the Safety Content Knowledge Survey and secondary school chemistry teachers who have had undergraduate and/or graduate safety training and those who have not had undergraduate and/or graduate safety training. Secondary school chemistry teachers who attended school district sponsored safety inservices did not score higher on the Safety Content Knowledge Survey than teachers who did not attend school district sponsored safety inservice sessions. The type of school district (urban, suburban, or rural) had no significant correlation to the type of laboratory safety equipment found in the instructional chemistry laboratory. The certification area (chemistry or other type of certificate which may or may not include chemistry) of the secondary school teacher had no significant correlation to the type of laboratory equipment found in the instructional chemistry laboratory. Overall, this study indicated a majority of secondary school chemistry teachers were interested in attending safety workshops applicable to chemistry safety. Throughout this research project, many teachers indicated they were not adequately instructed on the collegiate level in science safety and had to rely on common sense and self-study in their future teaching careers.

Improving eye safety in citrus harvest crews through the acceptance of personal protective equipment, community-based participatory research, social marketing, and community health workers.

Science.gov (United States)

Tovar-Aguilar, J Antonio; Monaghan, Paul F; Bryant, Carol A; Esposito, Andrew; Wade, Mark; Ruiz, Omar; McDermott, Robert J

2014-01-01

For the last 10 years, the Partnership for Citrus Workers Health (PCWH) has been an evidence-based intervention program that promotes the adoption of protective eye safety equipment among Spanish-speaking farmworkers of Florida. At the root of this program is the systematic use of community-based preventive marketing (CBPM) and the training of community health workers (CHWs) among citrus harvester using popular education. CBPM is a model that combines the organizational system of community-based participatory research (CBPR) and the strategies of social marketing. This particular program relied on formative research data using a mixed-methods approach and a multilevel stakeholder analysis that allowed for rapid dissemination, effective increase of personal protective equipment (PPE) usage, and a subsequent impact on adoptive workers and companies. Focus groups, face-to-face interviews, surveys, participant observation, Greco-Latin square, and quasi-experimental tests were implemented. A 20-hour popular education training produced CHWs that translated results of the formative research to potential adopters and also provided first aid skills for eye injuries. Reduction of injuries is not limited to the use of safety glasses, but also to the adoption of timely intervention and regular eye hygiene. Limitations include adoption in only large companies, rapid decline of eye safety glasses without consistent intervention, technological limitations of glasses, and thorough cost-benefit analysis.

Experience with rotating equipment in FBTR

International Nuclear Information System (INIS)

Babu, A.; Subramanian, K.G.; Selvam, B.

2006-01-01

Fast breeder test reactor (FBTR) is a 40MWth sodium cooled PuC-UC fuelled fast breeder reactor located at Kalpakkam, Tamilnadu. FBTR has a variety of mechanical and electrical equipment comprising of single stage and multi stage water pumps, sodium pumps, compressors, Chiller units, Diesel Generator sets, vacuum pumps, ventilation fans, Turbo generator etc. The degree and extent of maintenance on these equipment are decided based on nuclear safety, station technical specification requirements, design criteria and operating philosophy. The variety and complexity of the equipment in FBTR demands various maintenance strategies (preventive, predictive, proactive etc.) for their upkeep. This paper discusses the performance of all mechanical and electrical equipment in FBTR, measures taken for their upkeep, major modifications done to prevent recurring failures, condition monitoring programme implemented, Maintenance Management and Information System developed for data acquisition and dissemination, etc. (author)

A multi-criteria decision making approach for the selection of a flexible packaging equipment

Directory of Open Access Journals (Sweden)

Cristea Ciprian

2017-01-01

Full Text Available Flexible packaging is one of the fastest growing segments of the packaging industry, combining paper, plastic film and aluminum foil to deliver a broad array of products for food and beverage, personal care products, and pharmaceutical industries. In order to preserve the quality and safety of the product contained in them, there are currently a variety of flexible packaging equipment options available. However their relative costs and performance differ. This study proposes an application of a ranking methodology to assess a selection of the most suitable flexible packaging equipment. The adequate criteria in the selection of equipment have been identified, and the considered options are assessed, considering the decision maker’s preferences and existing constraints. The options are ranked in terms of their suitability for selecting equipment by using the Electre III method. The results obtained from the simulation experiment highlight the effectiveness of the model in outranking different options in the process of equipment selection.

Farm Health and Safety

Science.gov (United States)

... the United States. Farms have many health and safety hazards, including Chemicals and pesticides Machinery, tools and ... inspection and maintenance can help prevent accidents. Using safety gloves, goggles and other protective equipment can also ...

Communications received from certain Member States regarding guidelines for transfers of nuclear-related dual-use equipment, materials, software and related technology

International Nuclear Information System (INIS)

2003-01-01

The Director General of the International Atomic Energy Agency has received Notes Verbales dated 28 February 2003 from the Resident Representatives to the Agency of Argentina, Austria, Belgium, Bulgaria, Canada, Cyprus, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Italy, Japan, Kazakhstan, Latvia, Netherlands, Norway, Portugal, Republic of Korea, Slovakia, Slovenia, South Africa, Spain, Sweden, Switzerland, Turkey, Ukraine, the United Kingdom of Great Britain and Northern Ireland, and the United States of America, relating to transfers of nuclear-related dual-use equipment, materials, software and related technology. The purpose of the Note Verbales is to provide further information on those Governments' Guidelines for Transfers of Nuclear-related Dual-use Equipment, Materials, Software and Related Technology. In the light of the wish expressed at the end of each Note Verbale, the text of the Notes Verbales is attached. The attachment to these Notes Verbales is also reproduced in full

Communications received from certain Member States regarding guidelines for transfers of nuclear-related dual-use equipment, materials, software and related technology

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-05-16

The Director General of the International Atomic Energy Agency has received Notes Verbales dated 28 February 2003 from the Resident Representatives to the Agency of Argentina, Austria, Belgium, Bulgaria, Canada, Cyprus, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Italy, Japan, Kazakhstan, Latvia, Netherlands, Norway, Portugal, Republic of Korea, Slovakia, Slovenia, South Africa, Spain, Sweden, Switzerland, Turkey, Ukraine, the United Kingdom of Great Britain and Northern Ireland, and the United States of America, relating to transfers of nuclear-related dual-use equipment, materials, software and related technology. The purpose of the Note Verbales is to provide further information on those Governments' Guidelines for Transfers of Nuclear-related Dual-use Equipment, Materials, Software and Related Technology. In the light of the wish expressed at the end of each Note Verbale, the text of the Notes Verbales is attached. The attachment to these Notes Verbales is also reproduced in full.

The development of synthetic test procedure for hot cell equipment systems in IMEF

International Nuclear Information System (INIS)

Ahn, Sang Bok; Lee, Key Soon; Park, Dae Kyu; Hong, Kwon Pyo; Choo, Yong Sun

1998-04-01

Hot cell facility should be confirmed to operation safety through pre-commissioning test after construction. In this report, the detailed procedure of hot cell equipment are described. The contents are as follows: 1. Entrance equipment of hot cell 2. Specimen transportation equipment between hot cells 3. Waste discharge equipment in hot cell 4. Specimen loading equipment to hot cell 5. Interlinking equipment in hot cell. (author). 4 tabs

Health and safety manual

International Nuclear Information System (INIS)

1980-02-01

The manual consists of the following chapters: general policies and administration; the Environmental Health and Safety Department; the Medical Services Department: biological hazards; chemical safety; confined space entry; cryogenic safety; electrical safety; emergency plans; engineering and construction; evacuations, trenching, and shoring; fire safety; gases, flammable and compressed; guarding, mechanical; ladders and scaffolds, work surfaces; laser safety; materials handling and storage; noise; personal protective equipment; pressure safety; radiation safety, ionizing and non-ionizing; sanitation; seismic safety; training, environmental health and safety; tools, power and hand-operated; traffic and transportation; and warning signs and devices

30 CFR 56.12005 - Protection of power conductors from mobile equipment.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Protection of power conductors from mobile equipment. 56.12005 Section 56.12005 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-SURFACE METAL AND NONMETAL...

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)

Statistical evaluation of information reported to ISI and ISKO systems from a safety point of view

International Nuclear Information System (INIS)

Alonso Pallares, C.

1993-01-01

This paper describes he event percentages made by the main systems or equipment groups being the cause of incidents or directly linked to the incident. Command and protection systems, first-circuit equipment (BPC, VPC, volume compensator) safety systems, reactor installation and electrical input systems are analyzed. More over the main causes of notifies events are stressed and those where operation experience obtained in WWER-type nuclear power plants shows that and important part of incidents related to safety are due to personnel errors

30 CFR 57.9102 - Movement of independently operating rail equipment.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Movement of independently operating rail... NONMETAL MINES Loading, Hauling, and Dumping Traffic Safety § 57.9102 Movement of independently operating rail equipment. Movement of two or more pieces of rail equipment operating independently on the same...