OSART guidelines - 2005 edition. Reference report for IAEA Operational Safety Review Teams (OSARTs)

International Nuclear Information System (INIS)

2008-01-01

when exchanging ideas, at the working level, for enhancing safety. The OSART programme is based on the IAEA's Nuclear Safety Standard Series (Fundamentals, Requirements and Safety Guides) for nuclear power plants and the Basic Safety Standards for Radiation Protection. The Nuclear Safety Standards reflect the consensus of Member States on nuclear safety matters. The reports of the International Nuclear Safety Advisory Group, identifying important current nuclear safety issues also serve as references during an OSART review. The OSART Guidelines provide overall guidance for the experts to ensure the consistency and comprehensiveness of the operational safety review. Additional guidance and reference material prepared by the IAEA and the expertise of the OSART members contribute to the bases of the review. OSART reviews are performance oriented in that they accept different approaches to commissioning and operational safety that represent good practices and may contribute to ensuring a good safety record on the part of the operating organization. Recommendations are made on items of direct relevance to safety, whereas suggestions made might enhance plant safety indirectly but would certainly improve performance. Commendable good practices identified at plants are communicated to other plants where relevant in order to effect improvements. This revision of the OSART guidelines supersedes the 1994 Edition (IAEA-TECDOC-744)

Optimum Safety Levels for Breakwaters

DEFF Research Database (Denmark)

Burcharth, H. F.; Sørensen, John Dalsgaard

2005-01-01

Optimum design safety levels for rock and cube armoured rubble mound breakwaters without superstructure are investigated by numerical simulations on the basis of minimization of the total costs over the service life of the structure, taking into account typical uncertainties related to wave...... statistics and structure response. The study comprises the influence of interest rate, service lifetime, downtime costs and damage accumulation. Design limit states and safety classes for breakwaters are discussed. The results indicate that optimum safety levels are somewhat higher than the safety levels...

Rural expressway intersection safety toolbox : desktop reference.

Science.gov (United States)

2011-06-01

This document is intended to be a guide for planning-level decisions concerning safety : issues and subsequent potential improvements at rural expressway intersections. It is : NOT a design guide. It simply presents the gamut of safety treatment opti...

Safety Assessment for LILW Near-Surface Disposal Facility Using the IAEA Reference Model and MASCOT Program

International Nuclear Information System (INIS)

Kim, Hyun Joo; Park, Joo Wan; Kim, Chang Lak

2002-01-01

A reference scenario of vault safety case prepared by the IAEA for the near-surface disposal facility of low-and intermediate-level radioactive wastes is assessed with the MASCOT program. The appropriate conceptual models for the MASCOT implementation is developed. An assessment of groundwater pathway through a drinking well as a geosphere-biosphere interface is performed first, then biosphere pathway is analysed to estimate the radiological consequences of the disposed radionuclides based on compartment modeling approach. The validity of conceptual modeling for the reference scenario is investigated where possible comparing to the results generated by the other assessment. The result of this study shows that the typical conceptual model for groundwater pathway represented by the compartment model can be satisfactorily used for safety assessment of the entire disposal system in a consistent way. It is also shown that safety assessment of a disposal facility considering complex and various pathways would be possible by the MASCOT program

Discussion on Implementation of ICRP Recommendations Concerning Reference Levels and Optimisation

International Nuclear Information System (INIS)

2013-02-01

national plans. The International Atomic Energy Agency (IAEA) has adopted the ICRP recommendations into its revised Basic Safety Standards (BSSs). In practice, the full implementation will take some time. The EGIRES undertook this work to produce a report that would hopefully contribute to the common understanding of these important concepts and that would therefore assist in the use of reference levels and optimisation in the development of protection strategies in national plans. This report provides emergency management authorities in the NEA's member countries with clear and concise information and recommendations on key issues, possible approaches, and a summary of experience for implementing the new ICRP recommendations and revised BSSs for emergency exposure situations and for resultant existing exposure situations. (authors)

Diagnostic reference levels in medical imaging

International Nuclear Information System (INIS)

Rosenstein, M.

2001-01-01

The paper proposes additional advice to national or local authorities and the clinical community on the application of diagnostic reference levels as a practical tool to manage radiation doses to patients in diagnostic radiology and nuclear medicine. A survey was made of the various approaches that have been taken by authoritative bodies to establish diagnostic reference levels for medical imaging tasks. There are a variety of ways to implement the idea of diagnostic reference levels, depending on the medical imaging task of interest, the national or local state of practice and the national or local preferences for technical implementation. The existing International Commission on Radiological Protection (ICRP) guidance is reviewed, the survey information is summarized, a set of unifying principles is espoused and a statement of additional advice that has been proposed to ICRP Committee 3 is presented. The proposed advice would meet a need for a unifying set of principles to provide a framework for diagnostic reference levels but would allow flexibility in their selection and use. While some illustrative examples are given, the proposed advice does not specify the specific quantities to be used, the numerical values to be set for the quantities or the technical details of how national or local authorities should implement diagnostic reference levels. (author)

IAEA activities related to safety indicators, time frames and reference scenarios

International Nuclear Information System (INIS)

Batandjieva, B.; Hioki, K.; Metcalf, P.

2002-01-01

which was completed last year. The project proposed a methodology for the development of reference biospheres which is in line with suggestions of the International Commission for Radiological Protection 81 (ICRP). Its application is illustrated by way of three example biospheres. At present the Agency is continuing the work on development of reference scenarios for waste disposal in order to assist in the development of international agreement on the main principles and concepts for development and use of reference scenarios for use in the development of safety cases and decision making related to waste repositories. (authors)

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

Science.gov (United States)

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

2016-01-01

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

Evaluation of health and safety impacts of defense high-level waste in geologic repositories

International Nuclear Information System (INIS)

Smith, E.D.; Kocher, D.C.; Witherspoon, J.P.

1985-02-01

Pursuant to the requirement of the Nuclear Waste Policy Act of 1982 that the President evaluate the use of commercial high-level waste repositories for the disposal of defense high-level wastes, a comparative assessment has been performed of the potential health and safety impacts of disposal of defense wastes in commercial or defense-only repositories. Simplified models were used to make quantitative estimates of both long- and short-term health and safety impacts of several options for defense high-level waste disposal. The results indicate that potential health and safety impacts are not likely to vary significantly among the different disposal options for defense wastes. Estimated long-term health and safety impacts from all defense-waste disposal options are somewhat less than those from commercial waste disposal, while short-term health and safety impacts appear to be insensitive to the differences between defense and commercial wastes. In all cases, potential health and safety impacts are small because of the need to meet stringent standards promulgated by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. We conclude that health and safety impacts should not be a significant factor in the choice of a disposal option for defense high-level wastes. 20 references, 14 tables

Reference design and operations for deep borehole disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

2011-01-01

A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall

Reference design and operations for deep borehole disposal of high-level radioactive waste.

Energy Technology Data Exchange (ETDEWEB)

Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

2011-10-01

A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall

On Optimum Safety Levels of Breakwaters

DEFF Research Database (Denmark)

Burcharth, Hans F.; Sørensen, John Dalsgaard

2006-01-01

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

Environmental safety evaluation in test sea disposal of low-level radioactive wastes

International Nuclear Information System (INIS)

1979-01-01

The study results on the environmental safety in the test sea disposal of low-level wastes by Subcommittee on Radioactive Waste Safety Technology in Nuclear Safety Commission are given in connection with the test disposal of radioactive wastes into sea reported by the Nuclear Safety Bureau. The Subcommittee concludes that the effect of the test disposal of radioactive wastes into sea on the environment is extremely small. The contents are as follows. The full text of the report; attached data, (1) prediction of the concentrations of radioactive nuclides in sea, (2) calculation of the concentrations of radioactive nuclides in marine life with biological paths, and (3) estimation of exposure dose in general people; references (1) radiation exposure of the personnel engaged in sea disposal, (2) the effect of a sea disaster during ocean transport. (J.P.N.)

Safety Study of the X-Ray Reference Laboratory for Radiation Protection Levels (IR-14D)

International Nuclear Information System (INIS)

Garcia, G.

1999-01-01

This report is a study about the safety of the X-ray reference laboratory that has been recently constructed in the building 2 of the CIEMAT. After a brief description of the apparatus, we present the method used to calculate the exposure and absorbed dose rates in the most characteristic points of the laboratory. This method takes into account the spectral distribution of the radiation beams as a function of the accelerating voltage. The built-up factors of the absorbent materials have been considered to calculate the transmission of the radiation beams through the filters and shielding. Scattered radiations has been introduced in the calculations by means of a semiempirical method. This model supposes that multiple scattering processes give an isotropic contribution to the reflected beams and the single scattered can be described in terms of the differential cross section of Klein-Nishina. The results of this study have been applied to determine the maximum dose equivalent that the personnel of the laboratory could receive in normal operation conditions. (Author) 5 refs

Comparison of potential health and safety impacts of different disposal options for defense high-level wastes

International Nuclear Information System (INIS)

Kocher, D.C.; Smith, E.D.; Witherspoon, J.P.

1984-01-01

A comparative assessment has been performed of the potential long- and short-term health and safety impacts of different disposal options for defense high-level wastes. Conservative models and assumptions were used. The assessment suggests that considerations of health and safety will not be significant in choosing among disposal options, primarily because of the need to meet stringent standards in all cases. Rather, the ease and cost of assuring compliance of a particular disposal option with health and safety standards may be a more important factor. 11 references

Architecture Level Safety Analyses for Safety-Critical Systems

Directory of Open Access Journals (Sweden)

K. S. Kushal

2017-01-01

Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.

Individual employee's perceptions of " Group-level Safety Climate" (supervisor referenced) versus " Organization-level Safety Climate" (top management referenced): Associations with safety outcomes for lone workers.

Science.gov (United States)

Huang, Yueng-Hsiang; Lee, Jin; McFadden, Anna C; Rineer, Jennifer; Robertson, Michelle M

2017-01-01

Research has shown that safety climate is among the strongest predictors of safety behavior and safety outcomes in a variety of settings. Previous studies have established that safety climate is a multi-faceted construct referencing multiple levels of management within a company, most generally: the organization level (employee perceptions of top management's commitment to and prioritization of safety) and group level (employee perceptions of direct supervisor's commitment to and prioritization of safety). Yet, no research to date has examined the potential interaction between employees' organization-level safety climate (OSC) and group-level safety climate (GSC) perceptions. Furthermore, prior research has mainly focused on traditional work environments in which supervisors and workers interact in the same location throughout the day. Little research has been done to examine safety climate with regard to lone workers. The present study aims to address these gaps by examining the relationships between truck drivers' (as an example of lone workers) perceptions of OSC and GSC, both potential linear and non-linear relationships, and how these predict important safety outcomes. Participants were 8095 truck drivers from eight trucking companies in the United States with an average response rate of 44.8%. Results showed that employees' OSC and GSC perceptions are highly correlated (r= 0.78), but notable gaps between the two were observed for some truck drivers. Uniquely, both OSC and GSC scores were found to have curvilinear relationships with safe driving behavior, and both scores were equally predictive of safe driving behavior. Results also showed the two levels of climate significantly interacted with one another to predict safety behavior such that if either the OSC or GSC scores were low, the other's contribution to safety behavior became stronger. These findings suggest that OSC and GSC may function in a compensatory manner and promote safe driving behavior even

Probabilistic safety criteria at the safety function/system level

International Nuclear Information System (INIS)

1989-09-01

A Technical Committee Meeting was held in Vienna, Austria, from 26-30 January 1987. The objectives of the meeting were: to review the national developments of PSC at the level of safety functions/systems including future trends; to analyse basic principles, assumptions, and objectives; to compare numerical values and the rationale for choosing them; to compile the experience with use of such PSC; to analyse the role of uncertainties in particular regarding procedures for showing compliance. The general objective of establishing PSC at the level of safety functions/systems is to provide a pragmatic tool to evaluate plant safety which is placing emphasis on the prevention principle. Such criteria could thus lead to a better understanding of the importance to safety of the various functions which have to be performed to ensure the safety of the plant, and the engineering means of performing these functions. They would reflect the state-of-the-art in modern PSAs and could contribute to a balance in system design. This report, prepared by the participants of the meeting, reviews the current status and future trends in the field and should assist Member States in developing their national approaches. The draft of this document was also submitted to INSAG to be considered in its work to prepare a document on safety principles for nuclear power plants. Five papers presented at the meeting are also included in this publication. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Technology, safety, and costs of decommissioning a reference pressurized water reactor power station. Appendices

International Nuclear Information System (INIS)

Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.

1978-05-01

Detailed appendices are presented under the following headings: reference PWR facility description, reference PWR site description, estimates of residual radioactivity, alternative methods for financing decommissioning, radiation dose methodology, generic decommissioning activities, intermediate dismantlement activities, safe storage and deferred dismantlement activities, compilation of unit cost factors, and safety assessment details

Reference to the Safety Engineering Undergraduate Courses to Improve the Subjects and Contents of the Certified Safety Engineer Qualification and Examination System of China

OpenAIRE

Haibin Qiu; Shanghong Shi; Tingdi Zhao; Yiwei Qiao; Jiangshi Zhang

2013-01-01

The aim of this paper is to recommend that the subjects and contents of certified safety engineers use safety engineering undergraduate curriculum system for reference. Human resources play an important role in accident prevention and loss control. Education on safety engineering develops quickly in China. Moreover, the State Administration of Work Safety and the National Human Resources and Social Security Ministry have implemented a certified safety engineer qualification and examination sy...

Adaptive Reference Levels in a Level-Crossing Analog-to-Digital Converter

Directory of Open Access Journals (Sweden)

Andrew C. Singer

2008-11-01

Full Text Available Level-crossing analog-to-digital converters (LC ADCs have been considered in the literature and have been shown to efficiently sample certain classes of signals. One important aspect of their implementation is the placement of reference levels in the converter. The levels need to be appropriately located within the input dynamic range, in order to obtain samples efficiently. In this paper, we study optimization of the performance of such an LC ADC by providing several sequential algorithms that adaptively update the ADC reference levels. The accompanying performance analysis and simulation results show that as the signal length grows, the performance of the sequential algorithms asymptotically approaches that of the best choice that could only have been chosen in hindsight within a family of possible schemes.

FLIGHT SAFETY MANAGEMENT PROBLEMS AND EVALUATION OF FLIGHT SAFETY LEVEL OF AN AVIATION ENTERPRISE

Directory of Open Access Journals (Sweden)

B. V. Zubkov

2017-01-01

Full Text Available This article is devoted to studying the problem of safety management system (SMS and evaluating safety level of an aviation enterprise.This article discusses the problems of SMS, presented at the 41st meeting of the Russian Aviation Production Commanders Club in June 2014 in St. Petersburg in connection with the verification of the status of the CA of the Russian Federation by the International Civil Aviation Organization (ICAO in the same year, a set of urgent measures to eliminate the deficiencies identified in the current safety management system by participants of this meeting were proposed.In addition, the problems of evaluating flight safety level based on operation data of an aviation enterprise were analyzed. This analysis made it possible to take into account the problems listed in this article as a tool for a comprehensive study of SMS parameters and allows to analyze the quantitative indicators of the flights safety level.The concepts of Acceptable Safety Level (ASL indicators are interpreted differently depending on the available/applicable methods of their evaluation and how to implement them in SMS. However, the indicators for assessing ASL under operational condition at the aviation enterprise should become universal. Currently, defined safety levels and safety indicators are not yet established functionally and often with distorted underrepresented models describing their contextual contents, as well as ways of integrating them into SMS aviation enterprise.The results obtained can be used for better implementation of SMS and solving problems determining the aviation enterprise technical level of flight safety.

Disentangling the roles of safety climate and safety culture: Multi-level effects on the relationship between supervisor enforcement and safety compliance.

Science.gov (United States)

Petitta, Laura; Probst, Tahira M; Barbaranelli, Claudio; Ghezzi, Valerio

2017-02-01

Despite increasing attention to contextual effects on the relationship between supervisor enforcement and employee safety compliance, no study has yet explored the conjoint influence exerted simultaneously by organizational safety climate and safety culture. The present study seeks to address this literature shortcoming. We first begin by briefly discussing the theoretical distinctions between safety climate and culture and the rationale for examining these together. Next, using survey data collected from 1342 employees in 32 Italian organizations, we found that employee-level supervisor enforcement, organizational-level safety climate, and autocratic, bureaucratic, and technocratic safety culture dimensions all predicted individual-level safety compliance behaviors. However, the cross-level moderating effect of safety climate was bounded by certain safety culture dimensions, such that safety climate moderated the supervisor enforcement-compliance relationship only under the clan-patronage culture dimension. Additionally, the autocratic and bureaucratic culture dimensions attenuated the relationship between supervisor enforcement and compliance. Finally, when testing the effects of technocratic safety culture and cooperative safety culture, neither safety culture nor climate moderated the relationship between supervisor enforcement and safety compliance. The results suggest a complex relationship between organizational safety culture and safety climate, indicating that organizations with particular safety cultures may be more likely to develop more (or less) positive safety climates. Moreover, employee safety compliance is a function of supervisor safety leadership, as well as the safety climate and safety culture dimensions prevalent within the organization. Copyright © 2016 Elsevier Ltd. All rights reserved.

78 FR 21850 - Federal Motor Vehicle Safety Standards; Matters Incorporated by Reference

Science.gov (United States)

2013-04-12

... to the 1985 Annual Book of ASTM Standards, Vol. 05.04, ``Test Methods for Rating Motor, Diesel... for Rating Motor, Diesel, Aviation Fuels, A2. Reference Materials and Blending Accessories, (``ASTM... [Docket No. NHTSA-2011-0185] RIN 2127-AL25 Federal Motor Vehicle Safety Standards; Matters Incorporated by...

An N+3 Technology Level Reference Propulsion System

Science.gov (United States)

Jones, Scott M.; Haller, William J.; Tong, Michael To-Hing

2017-01-01

An N+3 technology level engine, suitable as a propulsion system for an advanced single-aisle transport, was developed as a reference cycle for use in technology assessment and decision-making efforts. This reference engine serves three main purposes: it provides thermodynamic quantities at each major engine station, it provides overall propulsion system performance data for vehicle designers to use in their analyses, and it can be used for comparison against other proposed N+3 technology-level propulsion systems on an equal basis. This reference cycle is meant to represent the expected capability of gas turbine engines in the N+3 timeframe given reasonable extrapolations of technology improvements and the ability to take full advantage of those improvements.

Practical considerations in the development and application of reference levels

International Nuclear Information System (INIS)

Selby, J. M.; Kennedy, W. E. Jr.; Swinth, K. L.; Gilbert, R. O.; Soldat, J. K.

1988-05-01

Radiological exposures to the public during an accident that releases radioactive materials to the environment occur initially through the air pathway, secondly through a water pathway, and finally, through the food chain. From these exposure pathways, reference levels of radionuclides must be applied to limit public radiation doses. Such reference levels are difficult to develop and apply, particularly if the contaminants are transuranic elements such as plutonium. The reference level must be related to a resultant theoretical dose to the public through generic calculations. However, to permit rapid decisions after an accident, the reference levels need to be related to conditions or media in the environment that can be directly measured. They must also be tied to specific countermeasures such as sheltering, evacuation, or interdiction of drinking water or food supplies. This paper discusses the practical considerations in the development and application of reference levels for transuranic in the environment and the present capability to rapidly detect and quantify their concentrations following accidental contamination events. 16 refs., 3 tabs

Temperature and level measurements realized for Nuclear Safety Level Improvement of Slovak NPPs

International Nuclear Information System (INIS)

Badiar, S.; Slanina, M.; Stanc, S.; Golan, P.; Krupa, J.

2001-01-01

Process of continual safety improvement in the individual Slovak nuclear power plants has been in progress since the beginning of nineties with the objective to upgrade the safety level of units in operation up to the European standards. In the framework of these activities, safety instrumentation systems with 1E qualification for the control of WWER reactor coolant systems were built and added. Methods for implementation of safety instrumentation systems for monitoring temperature and level in reactor coolant systems in the particular plants in Slovakia are presented showing the objectives and methods of their implementation. (Authors)

X-ray radiographic experimental investigation of the reference level in hydrostatic level measurement systems for boiling water reactors; Roentgen-radiografische Untersuchung des Referenzfuellstandes bei der hydrostatischen Fuellstandsmessung in Siedewasserreaktoren

Energy Technology Data Exchange (ETDEWEB)

Schulz, S.; Hampel, R. [Hochschule Zittau/Goerlitz, Zittau (Germany). Inst. fuer Prozesstechnik, Prozessautomatisierung und Messtechnik (IPM); Boden, S. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Dresden (Germany). Inst. fuer Fluiddynamik

2012-11-01

Hydrostatic level control is of high priority for normal operation safety of BWR-type reactors. For the prevention of undershooting the limiting values precise and secure measuring techniques are indispensable. Actually the filling level is determined form the hydrostatic pressure difference to a reference column. For the secure non-invasive detection of the phase boundary water/steam in inclined tubes the X-ray radiography has been chosen. The experiments were aimed to study possible geometric influences on the water/steam phase boundary. It was shown that the reference filling level is not significantly changed in spite of permanent phase transitions, provided an ideal mechanical construction of the system is given. Future experiments shall be focused on the analysis of interface behavior in case of non-ideal geometries (welds).

Regulatory review of probabilistic safety assessment (PSA) Level 2

International Nuclear Information System (INIS)

2001-07-01

Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from deterministic analysis. Many regulatory authorities consider the current state of the art in PSA to be sufficiently well developed for results to be used centrally in the regulatory decision making process-referred to as risk informed regulation. For these applications to be successful, it will be necessary for the regulatory authority to have a high degree of confidence in the PSA. However, at the 1994 IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997, the IAEA and OECD Nuclear Energy Agency agreed to produce, in cooperation, guidance on Regulatory Review of PSA. This led to the publication of IAEA-TECDOC-1135 on the Regulatory Review of Probabilistic Safety Assessment (PSA) Level 1, which gives advice for the review of Level 1 PSA for initiating events occurring at power plants. This TECDOC extends the coverage to address the regulatory review of Level 2 PSA.These publications are intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable level of quality so that it can be used as the

Probabilistic safety analysis second level of WWER-TOI

International Nuclear Information System (INIS)

Chekin, A.A.; Bajkova, E.V.; Levin, V.N.; Shishina, E.S.

2015-01-01

Probabilistic safety assessment (PSA) of Level-1 and Level-2 gives a comprehensive qualitative and quantitative evaluation of the safety of the project. The operation of the unit at rated power is considered. As sources of radioactivity in the development of the second-level PSA, nuclear fuel in the core of the reactor is considered. As initiating events, internal initiating events (including de-energizing) are considered, which may arise due to failures of NPP systems, equipment or components, or due to erroneous actions of personnel. In general, an assessment of the level of project safety shows that the WWER-TOI project complies with the requirements of the TOR, as well as all the requirements of modern Russian and foreign regulatory documents in the field of security [ru

Occupational Health and Safety. Numeracy. Level 1. Level 2. Level 3. Support Materials for Agricultural Training.

Science.gov (United States)

Batman, Kangan; Tully, Chris

This publication contains the three numeracy units of the three levels of Support Materials for Agricultural Training (SMAT) in the area of occupational health and safety: Level 1 (starting), 2 (continuing), and 3 (completing). The units are designed to help the learner improve his or her numeracy skills needed to deal with occupational safety and…

Establishing diagnostic reference levels in digital radiology

International Nuclear Information System (INIS)

Bana, Remy Wilson

2016-04-01

Medical application of radiation has gained wider study since diagnostic radiology plays a very important role in modern medicine. The need of the service seems to increase since the invention of digital radiology as a new technology that promises greater accuracy while minimizing patient dose. However, it is not exempted in the harmonization of doses delivered to the patient undergoing same radiologic examination in different institutions either regional or nationwide. The objective of this project was to review the establishment of Diagnostic Reference Levels (DRLs) in digital radiology at National level with the aim to reduce patient dose while maintaining appropriate image quality. A general discussion on digital radiology has been presented focusing on the optimization of patient dose as well as dosimetric quantities used for the establishment of DRLs. Recommendations have been provided for Rwanda to initiate steps to establish National Diagnostic Reference Levels for common procedures in digital radiology. (au)

Confidence improvement of disosal safety bydevelopement of a safety case for high-level radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Baik, Min Hoon; Ko, Nak Youl; Jeong, Jong Tae; Kim, Kyung Su [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

Many countries have developed a safety case suitable to their own countries in order to improve the confidence of disposal safety in deep geological disposal of high-level radioactive waste as well as to develop a disposal program and obtain its license. This study introduces and summarizes the meaning, necessity, and development process of the safety case for radioactive waste disposal. The disposal safety is also discussed in various aspects of the safety case. In addition, the status of safety case development in the foreign countries is briefly introduced for Switzerland, Japan, the United States of America, Sweden, and Finland. The strategy for the safety case development that is being developed by KAERI is also briefly introduced. Based on the safety case, we analyze the efforts necessary to improve confidence in disposal safety for high-level radioactive waste. Considering domestic situations, we propose and discuss some implementing methods for the improvement of disposal safety, such as construction of a reliable information database, understanding of processes related to safety, reduction of uncertainties in safety assessment, communication with stakeholders, and ensuring justice and transparency. This study will contribute to the understanding of the safety case for deep geological disposal and to improving confidence in disposal safety through the development of the safety case in Korea for the disposal of high-level radioactive waste.

Determining supply chain safety stock level and location

Directory of Open Access Journals (Sweden)

Bahareh Amirjabbari

2014-01-01

Full Text Available Purpose: The lean methodology and its principles have widely been applied in supply chain management in recent decades. Manufacturers are one of the most important contributors in a supply chain and inventory plays a paramount role for them to become lean. Therefore, there should be appropriate management of inventory and all of its drivers in accordance with a lean strategy. Safety stock is one of the main drivers of inventory; it protects against increasing the stretch in the breaking points of the supply chain, which in turn can result in possible reduction of inventory. In this paper an optimization model and a simulation model are developed and applied in a real case to optimize the safety stock level with the objective of logistics cost minimization.Design/methodology/approach: In order to optimize the safety stock level while minimizing logistics costs, a nonlinear cost minimization safety stock model is developed in this paper and then it is applied in a real world manufacturing case company. A safety stock simulation model based on appropriate metrics in the case company’s supply chain performance is also provided.Findings: These models result in not only the optimum levels but also locations of safety stock within the supply chain.Originality/value: In this research, two models of cost minimization and simulation have been developed and also applied in a real case company to result in not only optimized levels but also optimized locations of safety stock across the whole supply chain. In addition, the appropriate supply chain performance measurement metrics have been introduced in this paper and the simulation model is developed based on those.

Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants. Specific Safety Guide

International Nuclear Information System (INIS)

2010-01-01

The objective of this Safety Guide is to provide recommendations for meeting the IAEA safety requirements in performing or managing a level 2 probabilistic safety assessment (PSA) project for a nuclear power plant; thus it complements the Safety Guide on level 1 PSA. One of the aims of this Safety Guide is to promote a standard framework, standard terms and a standard set of documents for level 2 PSAs to facilitate regulatory and external peer review of their results. It describes all elements of the level 2 PSA that need to be carried out if the starting point is a fully comprehensive level 1 PSA. Contents: 1. Introduction; 2. PSA project management and organization; 3. Identification of design aspects important to severe accidents and acquisition of information; 4. Interface with level 1 PSA: Grouping of sequences; 5. Accident progression and containment analysis; 6. Source terms for severe accidents; 7. Documentation of the analysis: Presentation and interpretation of results; 8. Use and applications of the PSA; Annex I: Example of a typical schedule for a level 2 PSA; Annex II: Computer codes for simulation of severe accidents; Annex III: Sample outline of documentation for a level 2 PSA study.

Determination of carbon-14 in environmental level, solid reference materials

Energy Technology Data Exchange (ETDEWEB)

Blowers, Paul, E-mail: paul.blowers@cefas.co.uk [Cefas Lowestoft Laboratory, Pakefield Road, Lowestoft, Suffolk, NR33 0HT (United Kingdom); Caborn, Jane, E-mail: jane.a.caborn@nnl.co.uk [NNL, Springfields, Salwick, Preston, Lancashire, PR4 0XJ (United Kingdom); Dell, Tony [Veterinary Laboratories Agency, New Haw, Addlestone, Surrey, KT15 3NB (United Kingdom); Gingell, Terry [DSTL, Radiation Protection Services, Crescent Road, Alverstoke, Gosport, Hants, PO12 2DL (United Kingdom); Harms, Arvic [National Physical Laboratory, Hampton Road, Teddington, Middlesex, TW11 0LW (United Kingdom); Long, Stephanie [Radiological Protection Institute of Ireland, 3 Clonskeagh Square, Clonskeagh Road, Dublin 14, Ireland (United Kingdom); Sleep, Darren [Centre for Ecology and Hydrology, Lancaster Environment Centre, Library Avenue, Bailrigg, Lancaster, LA1 4AP (United Kingdom); Stewart, Charlie [UKAEA (Waste Management Group), Chemical Support Services, D1310/14, Dounreay, Thurso, Caithness, KW14 7TZ (United Kingdom); Walker, Jill [Radiocarbon Dating, The Old Stables, East Lockinge, Wantage, Oxon OX12 8QY (United Kingdom); Warwick, Phil E. [GAU-Radioanalytical, National Oceanography Centre Southampton, European Way, Southampton, SO14 3ZH (United Kingdom)

2011-10-15

An intercomparison exercise to determine the {sup 14}C activity concentrations in a range of solid, environmental level materials was conducted between laboratories in the UK. IAEA reference materials, C2, C6 and C7, and an in-house laboratory QA material were dispatched in 2006 to ten laboratories comprising of members of the Analyst Informal Working Group (AIWG) and one other invited party. The laboratories performed the determinations using a number of techniques, and using the results each one was evaluated in terms of levels of precision, sensitivity and limits of detection. The results of the study show that all techniques are capable of successfully analysing {sup 14}C in environmental level materials, however, a shortage of certified environmental reference materials exists. The suitability of the IAEA reference materials and other material for use as reference materials was also assessed.

Determination of carbon-14 in environmental level, solid reference materials

International Nuclear Information System (INIS)

Blowers, Paul; Caborn, Jane; Dell, Tony; Gingell, Terry; Harms, Arvic; Long, Stephanie; Sleep, Darren; Stewart, Charlie; Walker, Jill; Warwick, Phil E.

2011-01-01

An intercomparison exercise to determine the 14 C activity concentrations in a range of solid, environmental level materials was conducted between laboratories in the UK. IAEA reference materials, C2, C6 and C7, and an in-house laboratory QA material were dispatched in 2006 to ten laboratories comprising of members of the Analyst Informal Working Group (AIWG) and one other invited party. The laboratories performed the determinations using a number of techniques, and using the results each one was evaluated in terms of levels of precision, sensitivity and limits of detection. The results of the study show that all techniques are capable of successfully analysing 14 C in environmental level materials, however, a shortage of certified environmental reference materials exists. The suitability of the IAEA reference materials and other material for use as reference materials was also assessed.

The relationship between patient safety climate and occupational safety climate in healthcare - A multi-level investigation.

Science.gov (United States)

Pousette, Anders; Larsman, Pernilla; Eklöf, Mats; Törner, Marianne

2017-06-01

Patient safety climate/culture is attracting increasing research interest, but there is little research on its relation with organizational climates regarding other target domains. The aim of this study was to investigate the relationship between patient safety climate and occupational safety climate in healthcare. The climates were assessed using two questionnaires: Hospital Survey on Patient Safety Culture and Nordic Occupational Safety Climate Questionnaire. The final sample consisted of 1154 nurses, 886 assistant nurses, and 324 physicians, organized in 150 work units, within hospitals (117units), primary healthcare (5units) and elderly care (28units) in western Sweden, which represented 56% of the original sample contacted. Within each type of safety climate, two global dimensions were confirmed in a higher order factor analysis; one with an external focus relative the own unit, and one with an internal focus. Two methods were used to estimate the covariation between the global climate dimensions, in order to minimize the influence of bias from common method variance. First multilevel analysis was used for partitioning variances and covariances in a within unit part (individual level) and a between unit part (unit level). Second, a split sample technique was used to calculate unit level correlations based on aggregated observations from different respondents. Both methods showed associations similar in strength between the patient safety climate and the occupational safety climate domains. The results indicated that patient safety climate and occupational safety climate are strongly positively related at the unit level, and that the same organizational processes may be important for the development of both types of organizational climate. Safety improvement interventions should not be separated in different organizational processes, but be planned so that both patient safety and staff safety are considered concomitantly. Copyright © 2017 National Safety

Setting the standard: The IAEA safety standards set the global reference

International Nuclear Information System (INIS)

Williams, L.

2003-01-01

For the IAEA, setting and promoting standards for nuclear radiation, waste, and transport safety have been priorities from the start, rooted in the Agency's 1957 Statute. Today, a corpus of international standards are in place that national regulators and industries in many countries are applying, and more are being encouraged and assisted to follow them. Considerable work is done to keep safety standards updated and authoritative. They cover five main areas: the safety of nuclear facilities; radiation protection and safety of radiation sources; safe management of radioactive waste; safe transport of radioactive material; and thematic safety areas, such as emergency preparedness or legal infrastructures. Overall, the safety standards reflect an international consensus on what constitutes a high level of safety for protecting people and the environment. All IAEA Member States can nominate experts for the Agency standards committees and provide comments on draft standards. Through this ongoing cycle of review and feedback, the standards are refined, updated, and extended where needed

Reference hearing threshold levels for short duration signals

DEFF Research Database (Denmark)

Poulsen, Torben; Legarth, Søren Vase

2008-01-01

for the determination of reference hearing threshold levels. The results are given as peak-to-peak equivalent threshold sound pressure levels (peETSPL). The results are in good agreement with other sparse results from literature and are part of the basis for the ISO 389-6 standard from 2007....

Technology, safety, and costs of decommissioning a reference large irradiator and reference sealed sources

Energy Technology Data Exchange (ETDEWEB)

Haffner, D.R.; Villelgas, A.J. [Pacific Northwest Lab., Richland, WA (United States)

1996-01-01

This report contains the results of a study sponsored by the US Nuclear Regulatory Commission (NRC) to examine the decommissioning of large radioactive irradiators and their respective facilities, and a broad spectrum of sealed radioactive sources and their respective devices. Conceptual decommissioning activities are identified, and the technology, safety, and costs (in early 1993 dollars) associated with decommissioning the reference large irradiator and sealed source facilities are evaluated. The study provides bases and background data for possible future NRC rulemaking regarding decommissioning, for evaluation of the reasonableness of planned decommissioning actions, and for determining if adequate funds are reserved by the licensees for decommissioning of their large irradiator or sealed source facilities. Another purpose of this study is to provide background and information to assist licensees in planning and carrying out the decommissioning of their sealed radioactive sources and respective facilities.

Technology, safety, and costs of decommissioning a reference large irradiator and reference sealed sources

International Nuclear Information System (INIS)

Haffner, D.R.; Villelgas, A.J.

1996-01-01

This report contains the results of a study sponsored by the US Nuclear Regulatory Commission (NRC) to examine the decommissioning of large radioactive irradiators and their respective facilities, and a broad spectrum of sealed radioactive sources and their respective devices. Conceptual decommissioning activities are identified, and the technology, safety, and costs (in early 1993 dollars) associated with decommissioning the reference large irradiator and sealed source facilities are evaluated. The study provides bases and background data for possible future NRC rulemaking regarding decommissioning, for evaluation of the reasonableness of planned decommissioning actions, and for determining if adequate funds are reserved by the licensees for decommissioning of their large irradiator or sealed source facilities. Another purpose of this study is to provide background and information to assist licensees in planning and carrying out the decommissioning of their sealed radioactive sources and respective facilities

Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

International Nuclear Information System (INIS)

1999-11-01

A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as 36 Cl and 93 Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon

CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application

International Nuclear Information System (INIS)

Mankamo, T.; Bjoere, S.; Olsson, Lena

1992-12-01

Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems

Analysis of data relative to the update of diagnosis reference levels in radiology and nuclear medicine. Situation 2004-2006

International Nuclear Information System (INIS)

2008-01-01

This report presents the results of the analysis of patients dosimetry data the radiology and nuclear medicine institutions have to transmit yearly to I.R.S.N. in application of the 12. february decree disposal relative to the diagnosis reference levels in radiology and nuclear medicine. The analysed dosimetry data concern the evaluations realised between the date of decree publication, the 16. march 2004 and 31. december 2006. The so considered results have to allow to the Nuclear Safety Authority to define the evolution needs of the regulation. Particularly, the analysis of delivered doses in radiology and the activities given in nuclear medicine lead to propositions on the possible update of reference values of some examination types. (N.C.)

SCART guidelines. Reference report for IAEA Safety Culture Assessment Review Team (SCART)

International Nuclear Information System (INIS)

2008-01-01

The IAEA Director General stressed the role of safety culture in his concluding remarks at the Meeting of the Contracting Parties to the Convention on Nuclear Safety in 2002: 'As we have learned in other areas, it is not enough simply to have a structure; it is not enough to say that we have the necessary laws and the appropriate regulatory bodies. All these are important, but equally important is that we have in place a safety culture that gives effect to the structure that we have developed. To me, effectiveness and transparency are keys. So, it is an issue which I am pleased to see, you are giving the attention it deserves and we will continue to work with you in clarifying, developing and applying safety culture through our programmes and through our technical cooperation activities.' The concept of safety culture was initially developed by the International Nuclear Safety Advisory Group (INSAG) after the Chernobyl accident in 1986. Since then the IAEA's perspective of safety culture has expanded with time as its recognition of the complexities of the concept developed. Safety culture is considered to be specific organizational culture in all types of organizations with activities that give rise to radiation risks. The aim is to make safety culture strong and sustainable, so that safety becomes a primary focus for all activities in such organizations, even for those, which might not look safety-related at first. SCART (Safety Culture Assessment Review Team) is a safety review service, which reflects the expressed interest of Members States for methods and tools for safety culture assessment. It is a replacement for the earlier service ASCOT (Assessment of Safety Culture in Organizations Team). The IAEA Safety Fundamentals, Requirements and Guides (Safety Standards) are the basis for the SCART Safety Review Service. The reports of INSAG, identifying important current nuclear safety issues, serve also as references during a SCART mission. SCART missions are based

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Reference biospheres for the long term safety assessment of radioactive waste disposal facilities

International Nuclear Information System (INIS)

Crossland, I.G.; Torres, C.

2002-01-01

Regulatory guidance on the safety assessment of radioactive waste disposals usually requires the consequences of any radionuclide releases to be considered in terms of their potential impact on human health. This requires consideration of the prevailing biosphere and the habits of the potentially exposed humans within it. However, it could take many thousands of years for migrating radionuclides to reach the surface environment. In these circumstances, an assessment model that was based on the present-day biosphere could be inappropriate while future biospheres would be unpredictable. These and other considerations suggest that a standardised, or reference biosphere, approach may be useful. Theme 1 of the IAEA BIOMASS project was established to develop the concept of reference biospheres into a practical system that can be applied to the assessment of the long term safety of geological disposal facilities for radioactive waste. The technical phase of the project lasted for four years until November 2000 and brought together disparate interests from many countries including waste disposal agencies, regulators and technical experts. Building on the experience from earlier BIOMOVS projects, a methodology was constructed for the logical and defensible construction of mathematical biosphere models that can be used in the total system performance assessment of radioactive waste disposal. The methodology was then further developed through the creation of a series of BIOMASS Example Reference Biospheres ('Examples'). These are stylised biosphere models that, in addition to illustrating the methodology, are intended to be useful assessment tools in their own right. (author)

АSSESSMENT AND FORECASTING OF FLIGHT SAFETY LEVEL OF AIRLINE

Directory of Open Access Journals (Sweden)

E. S. Prozorov

2015-01-01

Full Text Available The article presents methods based on probability theory and mathematical statistics for solving a number of basic problems: formation and evaluation of the current flight safety level; forecasting the level of flight safety; ranking the objects (planes, pilots in terms of flight safety; evaluation of the presence (or absence of control actions arising in the context of the organization of corporate safety management system. At the same time as the main source of information are considered forward-looking events received from flight data.

Nuclear safety policy working group recommendations on nuclear propulsion safety for the space exploration initiative

Science.gov (United States)

Marshall, Albert C.; Lee, James H.; Mcculloch, William H.; Sawyer, J. Charles, Jr.; Bari, Robert A.; Cullingford, Hatice S.; Hardy, Alva C.; Niederauer, George F.; Remp, Kerry; Rice, John W.

1993-01-01

An interagency Nuclear Safety Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program. These recommendations, which are contained in this report, should facilitate the implementation of mission planning and conceptual design studies. The NSPWG has recommended a top-level policy to provide the guiding principles for the development and implementation of the SEI nuclear propulsion safety program. In addition, the NSPWG has reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. These recommendations should be useful for the development of the program's top-level requirements for safety functions (referred to as Safety Functional Requirements). The safety requirements and guidelines address the following topics: reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations.

Common basis of establishing safety standards and other safety decision-making levels for different sources of health risk

International Nuclear Information System (INIS)

Demin, V.F.

2002-01-01

Current approaches in establishing safety standards and other decision-making levels for different sources of health risk are critically analysed. To have a common basis for this decision-making a specific risk index R is recommended. In the common sense R is quantitatively defined as LLE caused by the annual exposure to the risk source considered: R = annual exposure, damage (LLE) from the exposure unit. This common definition is also rewritten in specific forms for a set of different risk sources (ionising radiation, chemical pollutants, etc): for different risk sources the exposure can be measured with different quantities (the probability of death, the exposure dose, etc.). R is relative LLE: LLE in years referred to 1 year under the risk. The dimension of this value is [year/year]. In the statistical sense R is conditionally the share of the year, which is lost due to exposure to a risk source during this year. In this sense R can be called as the relative damage. Really lifetime years are lost after the exposure. R can be in some conditional sense considered as a dimensionless quantity. General safety standards R n for the public and occupational workers have been suggested in terms of this index: R n = 0.0007 and 0.01 accordingly. Secondary safety standards are derived for a number of risk sources (ionising radiation, environmental chemical pollutants, etc). Values of R n are chosen in such a way that to have the secondary radiation BSS being equivalent to the current one's. Other general and derived levels for safety decision-making are also proposed including the de-minimus levels. Their possible dependence on the national or regional health-demographic data (HDD) is considered. Such issues as the ways of the integration and averaging of risk indices considered through the national or regional HDD for different risk sources and the use of non-threshold linear exposure - response relationships for ionising radiation and chemical pollutants are analysed

Use of standard reliability levels in design and safety assessment of in-pile loops

International Nuclear Information System (INIS)

Bogani, G.; Verre, A.; Balestreri, S.; Colombo, A.G.; Luisi, T.

1975-01-01

This paper describes a logic-probabilistic analysis technique for a critical design review and safety assessment of in-pile loops. The examples in this paper refer to the analysis performed for the experimental loops already constructed or under construction in the ESSOR reactor of the Joint Research Centre of Ispra, as irradiation facilities for fuel element research and development tests. The proposed technique is based on the classification into categories of components and protective device malfunctions. Such subdivision into categories was agreed upon by the Italian Safety Authority and Euratom JRC, and adopted for the safety assessment of the ESSOR reactor in-pile loops. For each category, the method makes a link with a corresponding malfunction probability range (probability level). This probability level is defined taking into account design, construction, inspection and maintenance criteria as well as periodic controls; therefore the quality level and consequently the reliability level are thus also defined. The analysis is developed in the following stages: (1) definition of the analysis object (top event) and drawing of the relative fault-tree; (2) loop design analysis and preliminary optimization based on logic criteria; (3) classification into categories of the fault-tree primary events; (4) final loop design analysis and optimization based on defined component quality requirements. Stages 2 and 4 are quite different since stage 2 mainly consists of a redundance optimization, while stage 4 acts on the component quality level in such a way that each minimum cut-set leading to the top has an acceptable probability level. During analysis development, use is made of computer codes which, among other things enable the verification of fault-tree logic makeup, the listing of the minimum cut-sets with and without event categorization, and the evaluation of each cut-set order. (author)

Dynamic probability evaluation of safety levels of earth-rockfill dams using Bayesian approach

Directory of Open Access Journals (Sweden)

Zi-wu Fan

2009-06-01

Full Text Available In order to accurately predict and control the aging process of dams, new information should be collected continuously to renew the quantitative evaluation of dam safety levels. Owing to the complex structural characteristics of dams, it is quite difficult to predict the time-varying factors affecting their safety levels. It is not feasible to employ dynamic reliability indices to evaluate the actual safety levels of dams. Based on the relevant regulations for dam safety classification in China, a dynamic probability description of dam safety levels was developed. Using the Bayesian approach and effective information mining, as well as real-time information, this study achieved more rational evaluation and prediction of dam safety levels. With the Bayesian expression of discrete stochastic variables, the a priori probabilities of the dam safety levels determined by experts were combined with the likelihood probability of the real-time check information, and the probability information for the evaluation of dam safety levels was renewed. The probability index was then applied to dam rehabilitation decision-making. This method helps reduce the difficulty and uncertainty of the evaluation of dam safety levels and complies with the current safe decision-making regulations for dams in China. It also enhances the application of current risk analysis methods for dam safety levels.

French diagnostic reference levels in diagnostic radiology, computed tomography and nuclear medicine: 2004-2008 Review

International Nuclear Information System (INIS)

Roch, P.; Aubert, B.

2013-01-01

After 5 y of collecting data on diagnostic reference levels (DRLs), the Nuclear Safety and Radiation Protection French Inst. (IRSN) presents the analyses of this data. The analyses of the collected data for radiology, computed tomography (CT) and nuclear medicine allow IRSN to estimate the level of regulatory application by health professionals and the representativeness of current DRL in terms of relevant examinations, dosimetric quantities, numerical values and patient morphologies. Since 2004, the involvement of professionals has highly increased, especially in nuclear medicine, followed by CT and then by radiology. Analyses show some discordance between regulatory examinations and clinical practice. Some of the dosimetric quantities used for the DRL setting are insufficient or not relevant enough, and some numerical values should also be reviewed. On the basis of these findings, IRSN formulates recommendations to update regulatory DRL with current and relevant examination lists, dosimetric quantities and numerical values. (authors)

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Appendices. Volume 2

Energy Technology Data Exchange (ETDEWEB)

Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

1980-06-01

Appendices are presented concerning the evaluations of decommissioning financing alternatives; reference site description; reference BWR facility description; radiation dose rate and concrete surface contamination data; radionuclide inventories; public radiation dose models and calculated maximum annual doses; decommissioning methods; generic decommissioning information; immediate dismantlement details; passive safe storage, continuing care, and deferred dismantlement details; entombment details; demolition and site restoration details; cost estimating bases; public radiological safety assessment details; and details of alternate study bases.

Referent 3D tumor model at cellular level in radionuclide therapy

International Nuclear Information System (INIS)

Spaic, R.; Ilic, R.D.; Petrovic, B.J.

2002-01-01

Aim Conventional internal dosimetry has a lot of limitations because of tumor dose nonuniformity. The best approach for absorbed dose at cellular level for different tumors in radionuclide therapy calculation is Monte Carlo method. The purpose of this study is to introduce referent tumor 3D model at cellular level for Monte Carlo simulation study in radionuclide therapy. Material and Methods The moment when tumor is detectable and when same therapy can start is time period in which referent 3D tumor model at cellular level was defined. In accordance with tumor growth rate at that moment he was a sphere with same radius (10 000 μm). In that tumor there are cells or cluster of cells, which are randomly distributed spheres. Distribution of cells/cluster of cells can be calculated from histology data but it was assumed that this distribution is normal with the same mean value and standard deviation (100±50 mm). Second parameter, which was selected to define referent tumor, is volume density of cells (30%). In this referent tumor there are no necroses. Stroma is defined as space between spheres with same concentration of materials as in spheres. Results: Referent tumor defined on this way have about 2,2 10 5 cells or cluster of cells random distributed. Using this referent 3D tumor model and for same concentration of radionuclides (1:100) and energy of beta emitters (1000 keV) which are homogeneously distributed in labeled cells absorbed dose for all cells was calculated. Simulations are done using FOTELP Monte Carlo code, which is modified for this purposes. Results of absorbed dose in cells are given in numerical values (1D distribution) and as the images (2D or 3D distributions). Conclusion Geometrical module for Monte Carlo simulation study can be standardized by introducing referent 3D tumor model at cellular level. This referent 3D tumor model gives most realistic presentation of different tumors at the moment of their detectability. Referent 3D tumor model at

77 FR 33777 - General Aviation Safety Forum: Climbing to the Next Level

Science.gov (United States)

2012-06-07

... NATIONAL TRANSPORTATION SAFETY BOARD General Aviation Safety Forum: Climbing to the Next Level The National Transportation Safety Board (NTSB) will convene a 2- day forum focused on safety issues related to... the Next Level,'' will be chaired by NTSB Chairman Deborah A. P. Hersman and all five Board Members...

Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 1. Main report

Energy Technology Data Exchange (ETDEWEB)

Jenkins, C. E.; Murphy, E. S.; Schneider, K J

1979-01-01

Detailed technology, safety and cost information are presented for the conceptual decommissioning of a reference small mixed oxide fuel fabrication plant. Alternate methods of decommissioning are described including immediate dismantlement, safe storage for a period of time followed by dismantlement and entombment. Safety analyses, both occupational and public, and cost evaluations were conducted for each mode.

The establishment and use of dose reference levels in general paediatric radiology

International Nuclear Information System (INIS)

Marsden, P.J.; Hardwick, J.; Mencik, C.; McLaren, C.; Young, C.; Mashford, P.

2001-01-01

Diagnostic reference levels for general paediatric radiology have been established in terms of delivered exposure parameters rather than skin dose or dose-area product. With supporting measurements from equipment quality assurance and assumptions of standard patient sizes it was possible to derive reference levels in terms of entrance surface dose. This allowed comparison to be made with other published data. The reference levels for common examinations are presented for different age bands. There is a notable variation with patient age for some examinations which is not apparent in other published data. (author)

OSART Guidelines. 2015 Edition. Reference Report for IAEA Operational Safety Review Teams (OSARTs)

International Nuclear Information System (INIS)

2016-01-01

The IAEA works to provide a global nuclear safety and security framework for the protection of people and the environment from the effects of ionizing radiation, the minimization of the likelihood of accidents that could endanger life and property, and effective mitigation of the effects of any such events, should they occur. The strategic approach to achieving such a framework involves continual improvement in four areas: national and international safety infrastructures; the establishment and global acceptance of IAEA safety standards; an integrated approach to the provision for the application of the safety standards; and a global network of knowledge and experience. The IAEA Operational Safety Review Team (OSART) programme provides advice and assistance to Member States to enhance the safety of nuclear power plants during commissioning and operation. The OSART programme, initiated in 1982, is available to all Member States with nuclear power plants under commissioning or in operation. Conservative design, careful manufacture and sound construction are all prerequisites for the safe operation of nuclear power plants. However, the safety of the plant also depends ultimately on: sound management, policies, procedures, processes and practices; the capability and reliability of commissioning and operating personnel; comprehensive instructions; sound accident management and emergency preparedness; and adequate resources. Finally, a positive attitude and conscientiousness on the part of all staff in discharging their responsibilities is important to safety. The OSART programme is based on the safety standards applicable to nuclear power plants. IAEA safety standards reflect the consensus of Member States on nuclear safety matters. The reports of the International Nuclear Safety Group identify important current nuclear safety issues and also serve as references during an OSART review. The publication OSART Guidelines provides overall guidance on the conduct of OSART

Estimation of average hazardous-event-frequency for allocation of safety-integrity levels

International Nuclear Information System (INIS)

Misumi, Y.; Sato, Y.

1999-01-01

One of the fundamental concepts of the draft international standard, IEC 61508, is target failure measures to be allocated to Electric/Electronic/Programmable Electronic Safety-Related Systems, i.e. Safety Integrity Levels. The Safety Integrity Levels consist of four discrete probabilistic levels for specifying the safety integrity requirements or the safety functions to be allocated to Electric/Electronic/Programmable Electronic Safety-Related Systems. In order to select the Safety Integrity Levels the draft standard classifies Electric/Electronic/Programmable Electronic Safety-Related Systems into two modes of operation using demand frequencies only. It is not clear which modes of operation should be applied to Electric/Electronic/Programmable Electronic Safety-Related Systems taking into account the demand-state probability and the spurious demand frequency. It is essential for the allocation of Safety Integrity Levels that generic algorithms be derived by involving possible parameters, which make it possible to model the actuality of real systems. The present paper addresses this issue. First of all, the overall system including Electric/Electronic/programmable Electronic Safety-Related Systems is described using a simplified fault-tree. Then, the relationships among demands, demand-states and proof-tests are studied. Overall systems are classified into two groups: a non-demand-state-at-proof-test system which includes both repairable and non-repairable demand states and a constant-demand-frequency system. The new ideas such as a demand-state, spurious demand-state, mean time between detections, rates of d-failure and h-failure, and an h/d ratio are introduced in order to make the Safety Integrity Levels and modes of operation generic and comprehensive. Finally, the overall system is simplified and modeled by fault-trees using Priority-AND gates. At the same time the assumptions for modeling are described. Generic algorithms to estimate hazardous

On Safety Management. A Frame of Reference for Studies of Safety Management with Examples From Non-Nuclear Contexts of Relevance for Nuclear Safety

International Nuclear Information System (INIS)

Svensson, Ola; Salo, Ilkka; Allwin, Pernilla

2004-11-01

A good knowledge about safety management from risk technologies outside the area of nuclear power may contribute to both broaden the perspectives on safety management in general, and point at new opportunities for improving safety measures within the nuclear industry. First, a theoretical framework for the study of safety management in general is presented, followed by three case studies on safety management from different non-nuclear areas with potential relevance for nuclear safety. The chapters are written as separate reports and can be read independently of each other. The nuclear industry has a long experience about the management of risky activities, involving all the stages from planing to implementation, both on a more generalized level and in the specific branches of activities (management, administration, operation, maintenance, etc.). Here, safety management is a key concept related to these areas of activities. Outside the field of nuclear power there exist a number of different non-nuclear risk technologies, each one with their own specific needs and experiences about safety management. The differences between the areas consist partly of the different experiences caused by the different technologies. Besides using own experiences in safety practices within the own areas of activities, it may be profitable to take advantage in knowledge and experiences from one area and put it in practice in another area. In order to facilitate knowledge transfer from one technological area to another it may be possible to adapt a common theoretical model, for descriptions and explanations, to the different technologies. Such a model should admit that common denominators for safety management across the areas might be identified and described with common concepts. Systems theory gives the opportunity to not only create models that are descriptive for events within the limits of a given technology, but also to generate knowledge that can be transferred to other

Low and intermediate level waste in SFR-1. Reference waste inventory

International Nuclear Information System (INIS)

Riggare, P.; Johansson, Claes

2001-06-01

The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR-1 at the time of closure. This report is a part of the SAFE project (Safety Assessment of Final Repository for Radioactive Operational Waste), i.e. the renewed safety assessment of SFR-1. The accounted waste inventory has been used as input to the release calculation that has been performed in the SAFE project. The waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 40 years and that closure of the SFR repository will happen in 2030. In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemo toxic material has been identified in the waste. The inventory is based on so called waste types and the waste types reference waste package. The reference waste package combined with a prognosis of the number of waste packages to the year 2030 gives the final waste inventory for SFR-1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors. In the SAFE project's prerequisites it was said that one realistic and one conservative (pessimistic) inventory should be produced. The conservative one should then be used for the release calculations. In this report one realistic and one conservative radionuclide inventory is presented. The conservative one adds up to 10 16 Bq. Regarding materials there is only one inventory given since it is not certain what is a conservative assumption

Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-11-01

A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as {sup 36}Cl and {sup 93}Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon.

IAEA safety fundamentals: the safety of nuclear installations and the defence in depth concept

International Nuclear Information System (INIS)

Aro, I.

2005-01-01

This presentation is a replica of the similar presentation provided by the IAEA Basic Professional Training Course on Nuclear Safety. The presentation utilizes the IAEA Safety Series document No. 110, Safety Fundamentals: the Safety of Nuclear Installations. The objective of the presentation is to provide the basic rationale for actions in provision of nuclear safety. The presentation also provides basis to understand national nuclear safety requirements. There are three Safety Fundamentals documents in the IAEA Safety Series: one for nuclear safety, one for radiation safety and one for waste safety. The IAEA is currently revising its Safety Fundamentals by combining them into one general Safety Fundamentals document. The IAEA Safety Fundamentals are not binding requirements to the Member States. But, a very similar text has been provided in the Convention on Nuclear Safety which is legally binding for the Member State after ratification by the Parliament. This presentation concentrates on nuclear safety. The Safety Fundamentals documents are the 'policy documents' of the IAEA Safety Standards Series. They state the basic objectives, concepts and principles involved in ensuring protection and safety in the development and application of atomic energy for peaceful purposes. They will state - without providing technical details and without going into the application of principles - the rationale for actions necessary in meeting Safety Requirements. Chapter 7 of this presentation describes the basic features of defence in depth concept which is referred to in the Safety Fundamentals document. The defence in depth concept is a key issue in reaching high level of safety specifically at the design stage but as the reader can see the extended concept also refers to the operational stage. The appendix has been taken directly from the IAEA Basic Professional Training Course on Nuclear Safety and applied to the Finnish conditions. The text originates from the references

EUMENES, a computer software for managing the radiation safety program information at an institutional level

International Nuclear Information System (INIS)

Hernandez Saiz, Alejandro; Cornejo Diaz, Nestor; Valdes Ramos, Maryzury; Martinez Gonzalez, Alina; Gonzalez Rodriguez, Niurka; Vergara Gil, Alex

2008-01-01

The correct application of national and international regulations in the field of Radiological Safety requires the implementation of Radiation Safety Programs appropriate to the developed practice. These Programs demand the preparation and keeping of an important number of records and data, the compliance with working schedules, systematic quality controls, audits, delivery of information to the Regulatory Authority, the execution of radiological assessments, etc. Therefore, it is unquestionable the necessity and importance of having a computer tool to support the management of the information related to the Radiation Safety Program in any institution. The present work describes a computer program that allows the efficient management of these data. Its design was based on the IAEA International Basic Safety Standards recommendations and on the requirements of the Cuban national standards, with the objective of being flexible enough to be applied in most of the institutions using ionizing radiations. The most important records of Radiation Safety Programs were incorporated and reports can be generated by the users. An additional tools-module allows the user to access to a radionuclide data library, and to carry out different calculations of interest in radiological protection. The program has been developed in Borland Delphi and manages Microsoft Access databases. It is a user friendly code that aims to support the optimization of Radiation Safety Programs. The program contributes to save resources and time, as the generated information is electronically kept and transmitted. The code has different security access levels according to the user responsibility at the institution and also provides for the analysis of the introduced data, in a quick and efficient way, as well as to notice deadlines, the exceeding of reference levels and situations that require attention. (author)

Referring to IAEA system to improve Chinese standards system on nuclear and radiation safety

International Nuclear Information System (INIS)

Shang Zhaorong; Wang Wenhai

2010-01-01

Referring to the standards system of IAEA, to build and improve the Chinese standards system of nuclear and radiation safety is a long term infrastructure work and an assurance to keep sustainable development of nuclear industry and nuclear technology application in China. The paper analyses the current main problem, and gives some suggestions on developing and improving the system. (authors)

Partial Safety Factors and Target Reliability Level in Danish Structural Codes

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Hansen, J. O.; Nielsen, T. A.

2001-01-01

The partial safety factors in the newly revised Danish structural codes have been derived using a reliability-based calibration. The calibrated partial safety factors result in the same average reliability level as in the previous codes, but a much more uniform reliability level has been obtained....... The paper describes the code format, the stochastic models and the resulting optimised partial safety factors....

Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

International Nuclear Information System (INIS)

Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

1984-09-01

The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits

Establishment of local Diagnostic Reference Levels (DRLs) for ...

African Journals Online (AJOL)

Diagnostic reference levels (DRLs) is an essential optimization tool in radiography and radiological sciences. The objective of the study is to establish DRL for radiography examinations in north eastern Nigeria. A Prospective cross- sectional study conducted in two university teaching hospitals in north eastern Nigeria.

Regulatory review of probabilistic safety assessment (PSA) level 1

International Nuclear Information System (INIS)

2000-02-01

Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from the deterministic analysis. Many regulatory authorities consider that the current state of the art in PSA (especially Level 1 PSA) is sufficiently well developed that it can be used centrally in the regulatory decision making process - referred to as 'risk informed regulation'. For these applications to be successful, it will be necessary for regulatory authorities to have a high degree of confidence in PSA. However, at the IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process in 1994 and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' Meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997 the IAEA and OECD Nuclear Energy Agency agreed to produce in co-operation a technical document on the regulatory review of PSA. This publication is intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable standard so that it can be used as the basis for taking risk informed decisions within a regulatory decision making process. The document gives guidance on how to set about reviewing a PSA and on the technical issues that need to be addressed. This publication gives guidance for the review of Level 1 PSA for

Establishing a reference range for triiodothyronine levels in preterm infants.

Science.gov (United States)

Oh, Ki Won; Koo, Mi Sung; Park, Hye Won; Chung, Mi Lim; Kim, Min-ho; Lim, Gina

2014-10-01

Thyroid dysfunction affects clinical complications in preterm infants and older children. However, thyroid hormone replacement in preterm infants has no proven benefits, possibly owing to the lack of an appropriate reference range for thyroid hormone levels. We aimed to establish a reference range for triiodothyronine (T3) levels at 1-month postnatal age (PNA) in preterm infants. This retrospective study included preterm infants born at a tertiary referral neonatal center at gestational age (GA)<35 weeks with no apparent thyroid dysfunction, for 6 consecutive years, with follow-up from PNA 2 weeks to 16 weeks. Using thyroid function tests (TFT), the relationships between T3 levels and thyrotropin (TSH) and free thyroxine (fT4) levels, birth weight, GA, postmenstrual age (PMA), and PNA were examined. The conversion trend for fT4 to T3 was analyzed using the T3/fT4 ratio. Overall, 464 TFTs from 266 infants were analyzed, after excluding 65 infants with thyroid dysfunction. T3 levels increased with fT4 levels, birth weight, GA, PMA, and PNA but not with TSH levels. The T3/fT4 ratio also increased with GA, PNA, and PMA. The average T3 level at 1 month PNA was 72.56 ± 27.83 ng/dL, with significant stratifications by GA. Relatively low T3 and fT4 levels in preterm infants were considered normal, with T3 levels and conversion trends increasing with GA, PMA, and PNA. Further studies are required to confirm the role of the present reference range in thyroid hormone replacement therapy. Copyright © 2014 Elsevier Ltd. All rights reserved.

Occupational Health and Safety. Level 1. Level 2. Level 3. Support Materials for Agricultural Training.

Science.gov (United States)

Batman, Kangan; Gadd, Nick; Lucas, Michele

This publication contains the three communication skills units of the three levels of Support Materials for Agricultural Training (SMAT) in the area of occupational health and safety: Level 1 (starting), 2 (continuing), and 3 (completing). The units are designed to help the learner improve his or her written and spoken communication skills needed…

Criteria of reference radionuclides for safety analysis of spent fuel waste disposal

International Nuclear Information System (INIS)

Suryanto

1998-01-01

Study on the criteria for reference radionuclides selection for assessment on spent fuel disposal have done. The reference radionuclides in this study means radionuclides are predicted to contribute of the most radiological effect for man if spent fuel waste are discharged on deep geology formation. The research was done by investigate critically of parameters were used on evaluation a kind of radionuclide. Especially, this research study of parameter which relevant disposal case and or spent fuel waste on deep geology formation . The research assumed that spent fuel discharged on deep geology by depth 500-1000 meters from surface of the land. The migration scenario Radionuclides from waste form to man was assumed particularly for normal release in which Radionuclides discharge from waste form in a series thorough container, buffer, geological, rock, to fracture(fault) and move together with ground water go to biosphere and than go into human body. On this scenario, the parameter such as radionuclides inventory, half life, heat generation, hazard index based on maximum permissible concentration (MPC) or annual limit on intake (ALI) was developed as criteria of reference radionuclides selection. The research concluded that radionuclides inventory, half live, heat generated, hazard index base on MPC or ALI can be used as criteria for selection of reference Radionuclide. The research obtained that the main radionuclides are predicted give the most radiological effect to human are as Cs-137, Sr-90, I-129, Am-243, Cm-244, Pu-238, Pu-239, Pu-240. The radionuclides reasonable to be used as reference radionuclides in safety analysis at spent fuel disposal. (author)

Low-level lasers alter mRNA levels from traditional reference genes used in breast cancer cells

Science.gov (United States)

Teixeira, A. F.; Canuto, K. S.; Rodrigues, J. A.; Fonseca, A. S.; Mencalha, A. L.

2017-07-01

Cancer is among the leading causes of mortality worldwide, increasing the importance of treatment development. Low-level lasers are used in several diseases, but some concerns remains on cancers. Reverse transcriptase quantitative polymerase chain reaction (RT-qPCR) is a technique used to understand cellular behavior through quantification of mRNA levels. Output data from target genes are commonly relative to a reference that cannot vary according to treatment. This study evaluated reference genes levels from MDA-MB-231 cells exposed to red or infrared lasers at different fluences. Cultures were exposed to red and infrared lasers, incubated (4 h, 37 °C), total RNA was extracted and cDNA synthesis was performed to evaluate mRNA levels from ACTB, GUSB and TRFC genes by RT-qPCR. Specific amplification was verified by melting curves and agarose gel electrophoresis. RefFinder enabled data analysis by geNorm, NormFinder and BestKeeper. Specific amplifications were obtained and, although mRNA levels from ACTB, GUSB or TRFC genes presented no significant variation through traditional statistical analysis, Excel-based tools revealed that the use of these reference genes are dependent of laser characteristics. Our data showed that exposure to low-level red and infrared lasers at different fluences alter the mRNA levels from ACTB, GUSB and TRFC in MDA-MB-231 cells.

Study and design of safety assessment model based on H12 reference case using GoldSim

International Nuclear Information System (INIS)

Nakajima, Kunihiko; Koo, Shigeru; Ebina, Takanori; Ebashi, Takeshi; Inagaki, Manabu

2009-07-01

Reference case of safety assessment analysis at the H12 report was calculated using the numerical code MESHNOTE and MATRICS mainly. On the other hand, recently general simulation software witch has a character of object-oriented is globally used and the numerical code GoldSim is typical software. After the H12 report, probability theory analysis and sensitivity analysis using GoldSim have carried out by statistical method for the purpose of following up safety assessment analysis at the H12 report. On this report, details of the method for the model design using GoldSim are summarized, and to confirm calculation reproducibility, verification between the H12 report and GoldSim results were carried out. And the guide book of calculation method using GoldSim is maintained for other investigators at JAEA who want to calculate reference case on the H12 report. In the future, application resources on this report will be able to upgrade probability theory analysis and other conceptual models. (author)

Measuring patient safety culture: an assessment of the clustering of responses at unit level and hospital level

NARCIS (Netherlands)

Smits, M.; Wagner, C.; Spreeuwenberg, P.; Wal, van der G.

2009-01-01

OBJECTIVES: To test the claim that the Hospital Survey on Patient Safety Culture (HSOPS) measures patient safety culture instead of mere individual attitudes and to determine the most appropriate level (individual, unit or hospital level) for interventions aimed at improving the culture of patient

Measuring patient safety culture : an assessment of the clustering of responses at unit level and hospital level

NARCIS (Netherlands)

Smits, M.; Wagner, C.; Spreeuwenberg, P.; Wal, G. van der; Groenewegen, P.P.

2009-01-01

Objectives: To test the claim that the Hospital Survey on Patient Safety Culture (HSOPS) measures patient safety culture instead of mere individual attitudes and to determine the most appropriate level (individual, unit or hospital level) for interventions aimed at improving the culture of patient

Work-related driver safety: A multi-level investigation

OpenAIRE

AMANDA ROSE WARMERDAM

2017-01-01

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

Long-term safety and efficacy of biosimilar infliximab among patients with inflammatory arthritis switched from reference product

Directory of Open Access Journals (Sweden)

Abdalla A

2017-03-01

Full Text Available Abuelmagd Abdalla, Niamh Byrne, Richard Conway, Thomas Walsh, Geraldine Mannion, Michael Hanly, Miriam O’Sullivan, Ann Maria Curran, John J Carey Department of Rheumatology, Galway University Hospitals, Galway, Ireland Purpose: To evaluate the efficacy and safety of the biosimilar infliximab in adult patients with inflammatory arthritis switched from reference product in our center. Patients and methods: In April 2014, patients attending our rheumatology service for infliximab infusions were switched from reference product to the biosimilar infliximab following consent and hospital approval. Results: Around 34 patients with inflammatory arthritis were switched from reference product to biosimilar infliximab in 2014: 50% female, mean age 55 years (standard deviation=12.9, mean disease duration 14.79 years (9.7, median duration on infliximab 57 months, and two-thirds on oral disease-modifying antirheumatic drugs. There was no difference in efficacy or safety in the first 6 months of therapy. By the end of 2015, the mean follow-up on biosimilar infliximab was 15.8 (standard deviation=6.3 months. Our results showed no significant difference in Health Assessment Questionnaire score, patient global assessment of disease activity, number of disease flares, or the medication dose between the originator and the biosimilar infliximab. However, reported pain and C-reactive protein values were significantly higher during the longer follow-up period (p=0.043, 0.001 respectively. There was no significant difference in the number of adverse events or infusion reactions during follow-up periods. Only five (14.7% patients discontinued the biosimilar infliximab. Conclusion: Our patients experienced similar efficacy and safety for managing their arthritis with the biosimilar infliximab as the reference product infliximab, but at a much lower cost. Keywords: biologic therapy, rheumatic diseases, biosimilar exchange, infliximab, adult

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)

Development of High-Level Safety Requirements for a Pyroprocessing Facility

Energy Technology Data Exchange (ETDEWEB)

Seo, Seok Jun; Jo, Woo Jin; You, Gil Sung; Choung, Won Myung; Lee, Ho Hee; Kim, Hyun Min; Jeon, Hong Rae; Ku, Jeong Hoe; Lee, Hyo Jik [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Korea Atomic Energy Research Institute (KAERI) has been developing a pyroproceesing technology to reduce the waste volume and recycle some elements. The pyroprocessing includes several treatment processes which are related with not only radiological and physical but also chemical and electrochemical properties. Thus, it is of importance to establish safety design requirements considering all the aspects of those properties for a reliable pyroprocessing facility. In this study, high-level requirements are presented in terms of not only radiation protection, nuclear criticality, fire protection, and seismic safety but also confinement and chemical safety for the unique characteristics of a pyroprocessing facility. Several high-level safety design requirements such as radiation protection, nuclear criticality, fire protection, seismic, confinement, and chemical processing were presented for a pyroprocessing facility. The requirements must fulfill domestic and international safety technology standards for a nuclear facility. Furthermore, additional requirements should be considered for the unique electrochemical treatments in a pyroprocessing facility.

The Structure and Application of High Level Safety Goals. A Review by the MDEP Sub-committee on Safety Goals

International Nuclear Information System (INIS)

2011-01-01

One of the aims of MDEP is to work towards greater harmonisation of regulatory requirements. To achieve this aim, it is necessary that there is a degree of convergence on the safety goals that are required to be met by designers and operators. The term 'safety goals' is defined to cover all health and safety requirements which must be met: these may be deterministic rules and/or probabilistic targets. They should cover the safety of workers, public and the environment in line with the IAEA's Basic Safety Objective; encompassing safety in normal operation through to severe accidents. All regulators have safety goals, but these are expressed in many different ways and exercises in comparing them frequently are done at a very low level eg specific temperatures in the reactor vessel. The differences in the requirements from different regulators are difficult to resolve as the goals are derived using different principles and assumptions and are for a specific technology. Therefore MDEP set up a sub-committee to investigate a different approach. This approach was to start with the top level goals and to derive a structure and means of deriving lower tier goals that can be seen to be clearly related to the higher level ones. This approach has the potential to greatly assist in the process of harmonisation of regulatory requirements. The paper reviews the high level goals used in MDEP countries and the relevant work of international groups. From these it draws broad conclusions that the form of the framework should be an Hierarchical Structure of Safety Goals, incorporating an extended Defense-in-Depth approach. The basis concept is that the higher level safety goals can then developed, in a coherent and consistent manner, into lower level safety goals and targets that can be applied within the design and operation of reactors, with a clear connection between the different levels. This structured approach is technology-neutral and is sufficiently flexible that it can be

Levels of safety satisfactory for commercialization of the breeder

International Nuclear Information System (INIS)

Ferguson, R.L.

1979-01-01

A brief discussion is presented of the Department of Energy's LMFBR safety program and the safety levels which DOE believes would be satisfactory for the commercialization of the breeder are indicated. Some observations are offered on the Three Mile Island accident and some of its implications are discussed for the LMFBR program

Determination of diagnostic reference levels in general radiography in Latin America

International Nuclear Information System (INIS)

Blanco, S.; Mora, P.; Almonte, N.; Benavente, T.; Benson, N.; Blanco, D.; Cardenas, J.; Gomez, Y. D.; Edding, O.; Escobar, C.; Fonseca, M.; Gamarra, M.; Aguilar, J. G.; Khoury, H. J.; Quintero, A. R.; Zuniga, N. R.; Zaire, E.; Nader, A.

2013-01-01

The International Atomic Energy Agency (IAEA) through the International Action Plan on Radiation Protection of Patients and the International Commission on Radiological Protection have for some time carried out important efforts to assure that in the medical applications of the ionising radiations, the optimisation of radiological protection of patients is fundamental, to such a point that the IAEA includes it directly as a requirement for these practices (in its International Basic Safety Standards for Protection against Ionising Radiation and for the Safety of Radiation Sources (BSS)-GSR Part 1, 2011). For this reason, among the objectives of Regional Project RLA/9/057 and Regional Project RLA/9/067, the intention was to establish the dose references in conventional radiology for Latin America, for the purposes of determining whether these doses comply with the requirements of the BSS and to tend to improve practices, in order to minimise the dose received by the patients. (authors)

Improving the safety of Ukrainian NPP to reach an internationally accepted level

International Nuclear Information System (INIS)

Bozhko, S.; Helske, J.; Janke, R.; Mayoral, C.

2013-01-01

This paper summarizes the safety status and the modernization progress of Ukrainian NPPs towards an internationally accepted level of safety. After a brief discussion of the concept of what is called an 'international accepted level' for new and operating NPPs, the status of Russian type WWER and in particular the Ukrainian NPPs is presented. Then, the performed investigations of the gaps between international accepted level and the original status of Ukrainian NPPs are presented. The safety objectives of the modernization programs, some examples of defence in depth improvements, and an overall view of the modernization programs of Ukrainian NPPs are produced. Then, few important safety improvements implemented at the oldest Ukrainian WWER-1000 South Ukraine-1 are given in more detail. Finally, a conclusion presents the current status on the way to fulfill the national safety targets and to reach an internationally accepted level for all the Ukrainian NPPs. The paper is followed by the slides of the presentation. (authors)

Analysis of data related to the updating of diagnosis reference levels in radiology and nuclear medicine. Assessment 2007-2008

International Nuclear Information System (INIS)

2010-01-01

This report presents the results of the analysis of 'patient' dosimetric data which radiology and nuclear medicine establishments must supply every year to the IRSN (the French Radioprotection and Nuclear Safety Institute) according to a decree related to diagnosis reference levels in radiology and nuclear medicine. The analysed dosimetric data concern assessments performed during 2007 and 2008. For the different concerned practices (radiology, scanography, nuclear medicine), the report proposes a presentation and a discussion of global data, and then a presentation of data either for different types of examination on adults and on children, or for the different parts of the body

A multilevel model of patient safety culture: cross-level relationship between organizational culture and patient safety behavior in Taiwan's hospitals.

Science.gov (United States)

Chen, I-Chi; Ng, Hui-Fuang; Li, Hung-Hui

2012-01-01

As health-care organizations endeavor to improve their quality of care, there is a growing recognition of the importance of establishing a culture of patient safety. The main objective of this study was to investigate the cross-level influences of organizational culture on patient safety behavior in Taiwan's hospitals. The authors measured organizational culture (bureaucratic, supportive and innovative culture), patient safety culture and behavior from 788 hospital workers among 42 hospitals in Taiwan. Multilevel analysis was applied to explore the relationship between organizational culture (group level) and patient safety behavior (individual level). Patient safety culture had positive impact on patient safety behavior in Taiwan's hospitals. The results also indicated that bureaucratic, innovative and supportive organizational cultures all had direct influence on patient safety behavior. However, only supportive culture demonstrated significant moderation effect on the relationship between patient safety culture and patient safety behavior. Furthermore, organizational culture strength was shown correlated negatively with patient safety culture variability. Overall, organizational culture plays an important role in patient safety activities. Safety behaviors of hospital staff are partly influenced by the prevailing cultural norms in their organizations and work groups. For management implications, constructed patient priority from management commitment to leadership is necessary. For academic implications, research on patient safety should consider leadership, group dynamics and organizational learning. These factors are important for understanding the barriers and the possibilities embedded in patient safety. Copyright © 2011 John Wiley & Sons, Ltd.

Radiation safety in nuclear medicine procedures

International Nuclear Information System (INIS)

Cho, Sang Geon; Kim, Ja Hae; Song, Ho Chun

2017-01-01

Since the nuclear disaster at the Fukushima Daiichi Nuclear Power Plant in 2011, radiation safety has become an important issue in nuclear medicine. Many structured guidelines or recommendations of various academic societies or international campaigns demonstrate important issues of radiation safety in nuclear medicine procedures. There are ongoing efforts to fulfill the basic principles of radiation protection in daily nuclear medicine practice. This article reviews important principles of radiation protection in nuclear medicine procedures. Useful references, important issues, future perspectives of the optimization of nuclear medicine procedures, and diagnostic reference level are also discussed

Radiation safety in nuclear medicine procedures

Energy Technology Data Exchange (ETDEWEB)

Cho, Sang Geon; Kim, Ja Hae; Song, Ho Chun [Dept. of Nuclear Medicine, Medical Radiation Safety Research Center, Chonnam National University Hospital, Gwangju (Korea, Republic of)

2017-03-15

Since the nuclear disaster at the Fukushima Daiichi Nuclear Power Plant in 2011, radiation safety has become an important issue in nuclear medicine. Many structured guidelines or recommendations of various academic societies or international campaigns demonstrate important issues of radiation safety in nuclear medicine procedures. There are ongoing efforts to fulfill the basic principles of radiation protection in daily nuclear medicine practice. This article reviews important principles of radiation protection in nuclear medicine procedures. Useful references, important issues, future perspectives of the optimization of nuclear medicine procedures, and diagnostic reference level are also discussed.

Comparing climate and cost impacts of reference levels for reducing emissions from deforestation

Energy Technology Data Exchange (ETDEWEB)

Busch, Jonah [Center for Applied Biodiversity Science, Conservation International, 2011 Crystal Drive, Suite 500, Arlington, VA (United States); Strassburg, Bernardo [Center for Social and Economic Research on the Global Environment, University of East Anglia, Norwich NR4 7TJ (United Kingdom); Cattaneo, Andrea [Woods Hole Research Center, 149 Woods Hole Road, Falmouth, MA 02540-1644 (United States); Lubowski, Ruben [Environmental Defense Fund, 1875 Connecticut Avenue NW, Washington, DC (United States); Bruner, Aaron; Rice, Richard; Boltz, Frederick [Conservation International, 2011 Crystal Drive, Suite 500, Arlington, VA (United States); Creed, Anna; Ashton, Ralph, E-mail: jbusch@conservation.or [Terrestrial Carbon Group, 900 17th Street NW, Suite 700, Washington, DC (United States)

2009-10-15

The climate benefit and economic cost of an international mechanism for reducing emissions from deforestation and degradation (REDD) will depend on the design of reference levels for crediting emission reductions. We compare the impacts of six proposed reference level designs on emission reduction levels and on cost per emission reduction using a stylized partial equilibrium model (the open source impacts of REDD incentives spreadsheet; OSIRIS). The model explicitly incorporates national incentives to participate in an international REDD mechanism as well as international leakage of deforestation emissions. Our results show that a REDD mechanism can provide cost-efficient climate change mitigation benefits under a broad range of reference level designs. We find that the most effective reference level designs balance incentives to reduce historically high deforestation emissions with incentives to maintain historically low deforestation emissions. Estimates of emission reductions under REDD depend critically on the degree to which demand for tropical frontier agriculture generates leakage. This underscores the potential importance to REDD of complementary strategies to supply agricultural needs outside of the forest frontier.

Comparing climate and cost impacts of reference levels for reducing emissions from deforestation

International Nuclear Information System (INIS)

Busch, Jonah; Strassburg, Bernardo; Cattaneo, Andrea; Lubowski, Ruben; Bruner, Aaron; Rice, Richard; Boltz, Frederick; Creed, Anna; Ashton, Ralph

2009-01-01

The climate benefit and economic cost of an international mechanism for reducing emissions from deforestation and degradation (REDD) will depend on the design of reference levels for crediting emission reductions. We compare the impacts of six proposed reference level designs on emission reduction levels and on cost per emission reduction using a stylized partial equilibrium model (the open source impacts of REDD incentives spreadsheet; OSIRIS). The model explicitly incorporates national incentives to participate in an international REDD mechanism as well as international leakage of deforestation emissions. Our results show that a REDD mechanism can provide cost-efficient climate change mitigation benefits under a broad range of reference level designs. We find that the most effective reference level designs balance incentives to reduce historically high deforestation emissions with incentives to maintain historically low deforestation emissions. Estimates of emission reductions under REDD depend critically on the degree to which demand for tropical frontier agriculture generates leakage. This underscores the potential importance to REDD of complementary strategies to supply agricultural needs outside of the forest frontier.

THE CONDITION AND THE DYNAMICS OF CHANGES OF REGIONAL ENERGETIC SAFETY LEVEL

Directory of Open Access Journals (Sweden)

A.L. Myzin

2006-12-01

Full Text Available On the basis of indicative analysis method use, the dynamic processes of changes of energetic safety condition of federal districts and subjects of Russian Federation for last 5 years are investigated. The results of diagnosing safety levels for separate indicators, their blocks and the results of situation evaluation as a whole are discussed. The comparison of regions’ energetic safety condition is given, the causes of crisis situations appearance are discovered, and on this basis the suggestions for regions’ safety levels increasing are formulated.

Analysis on safety production in coal mines Henan Province

Institute of Scientific and Technical Information of China (English)

KONG Liu-an; ZHANG Wen-yong

2006-01-01

Based on the rigorous situation of safety production in coal mines, the paper analyzed the statistical data of recent accidents indexes in Henan's coal mines. Using investigation and comparison analysis methods, a specified analysis on mining conditions, technical facility level, safety input and vocational quality of workers in Henan's coal mines was conducted. The result indicates that there have been existing such main safety production problems as weak safety management, low-level facilities, inadequate safety input and poor vocational quality and so on. Finally it proposes such reference solutions as to establish and perfect coal mining supervision and management system, to increase safety investment into techniques and facilities and to strengthen workers' safety education and introduction of more high-level professional talents.

Radon reference levels and priority areas considering optimisation and avertable lung cancers

International Nuclear Information System (INIS)

Bochicchio, F.; Venoso, G.; Antignani, S.; Carpentieri, C.

2017-01-01

Protection from radon exposure in workplaces and dwellings, as included in the latest relevant international regulations and recommendations, is based on the new concept of 'reference level' whose meaning is significantly different from that of previous 'action level' concept. In fact, whereas remedial actions had to be considered only for radon concentrations above the action level, actions to optimise radon exposure are requested with priority above reference level but optimisation should be applied also for radon concentrations below reference level. Similar considerations can be applied to the usually called 'Rn-prone' areas, which are here proposed to be regulated as 'priority' areas. The main implication of these new challenging concepts is a substantial increase of avertable lung cancer deaths, as it will be shown using Italian data. Some practical examples of possible policy actions fitting an approach based on these new concepts will also be given, which could be useful for the implementation of the Council Directive 2013/59/Euratom. (authors)

Technology, safety and costs of decommissioning reference light water reactors following postulated accidents

International Nuclear Information System (INIS)

Konzek, G.J.; Smith, R.I.

1990-12-01

The estimated costs for post-accident cleanup at the reference BWR (developed previously in NUREG/CR-2601, Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents) are updated to January 1989 dollars in this report. A simple formula for escalating post-accident cleanup costs is also presented. Accident cleanup following the most severe accident described in NUREG/CR-2601 (i.e., the Scenario 3 accident) is estimated to cost from $1.22 to 1.44 billion, in 1989 dollars, for assumed escalation rates of 4% or 8% in the years following 1989. The time to accomplish cleanup remained unchanged from the 8.3 years originally estimated. No reanalysis of current information on the technical aspects of TMI-2 cleanup has been performed. Only the cost of inflation has been evaluated since the original PNL analysis was completed. 32 refs., 12 tabs

Reference levels for assays in opened installations of industrial radiography

International Nuclear Information System (INIS)

Leocadio, Joao C.; Tauhata, Luiz; Crispim, Verginia R.

2001-01-01

This work had as objectives to analyze the facilities open of industrial x-ray to obtain the distribution of doses in the operators and present proposed for the reference levels. The results of the monitoring revealed an improvement of the radiation protection conditions in the facilities and that the risk of potential exposure was reduced. The advantage of the proposed reference levels is that the supervisors would enlarge the frequency of audits in the facilities opened to accomplish the investigations and interventions.The facilities open with 'bunkers' presented distributions with 95% of the doses below 0,2 mSv and the distributions of the facilities with cordoned area they had 75% of the doses below 0,4 mSv. (author)

Safety Management. An Introduction to a Frame of Reference Exemplified with Case Studies from Non-Nuclear Contexts

Energy Technology Data Exchange (ETDEWEB)

Svensson, Ola (Risk Analysis, Social and Decision Research Unit, Dept. of Psychology, Stockholm Univ., Stockholm (Sweden)); Salo, Ilkka (Dept. of Psychology, Lund Univ., Lund (Sweden))

2004-11-15

A systems perspective on safety management is introduced followed by two briefly presented case studies of safety management in non-nuclear contexts. The first study concerns a car manufacturer and the second study a road traffic tunnel system. The risks of a car accident in the first case study are evident. The great exposure generates many incidents and accidents. In the second study, the rather low traffic intensity through the tunnel produces few incidents and accidents and only a few fatal accidents over the years. Yet, the risk of the individual traveler is much greater in the tunnel than on the average road. The case studies are presented in a systems perspective with emphasis on information feedback about the risks of the systems. The first case study illustrates high quality safety management, while the second case study shows many weaknesses of the safety management in the tunnel system. Some differences in safety management between the case studies are noted. The last part of the study presents an organizational perspective on safety management and offers alternative theoretical perspectives on the concept of safety management. The report shows that further studies are needed both (1) to develop a frame of reference for describing safety management across industries and activities and (2) to collect data illustrating of good and poor safety management. Then, the results can be used to strengthen and/or improving safety management in the nuclear power industry and its regulators

Safety Management. An Introduction to a Frame of Reference Exemplified with Case Studies from Non-Nuclear Contexts

International Nuclear Information System (INIS)

Svensson, Ola; Salo, Ilkka

2004-11-01

A systems perspective on safety management is introduced followed by two briefly presented case studies of safety management in non-nuclear contexts. The first study concerns a car manufacturer and the second study a road traffic tunnel system. The risks of a car accident in the first case study are evident. The great exposure generates many incidents and accidents. In the second study, the rather low traffic intensity through the tunnel produces few incidents and accidents and only a few fatal accidents over the years. Yet, the risk of the individual traveler is much greater in the tunnel than on the average road. The case studies are presented in a systems perspective with emphasis on information feedback about the risks of the systems. The first case study illustrates high quality safety management, while the second case study shows many weaknesses of the safety management in the tunnel system. Some differences in safety management between the case studies are noted. The last part of the study presents an organizational perspective on safety management and offers alternative theoretical perspectives on the concept of safety management. The report shows that further studies are needed both (1) to develop a frame of reference for describing safety management across industries and activities and (2) to collect data illustrating of good and poor safety management. Then, the results can be used to strengthen and/or improving safety management in the nuclear power industry and its regulators

The nuclear law: safety. 2006-2010

International Nuclear Information System (INIS)

Bringuier, P.

2010-01-01

The author discusses the legal evolutions related to nuclear safety between 2006 and 2010. He identifies three main topics of unequal importance. Firstly, he comments the implementation of an international reference framework which has been completed at the European level and which aims at the harmonization of safety and security rules. Secondly, he comments the creation of the French Nuclear Safety Authority (ASN, Autorite de Surete Nucleaire). Thirdly, he comments the recast of the standard framework in order to update the French law with respect to the international reference framework. This leaded to a new distribution of power and authority, to more complete and constraining procedures, and to the definition of procedures for each step of an installation life cycle

TU-EF-BRD-01: Topics in Quality and Safety Research and Level of Evidence

Energy Technology Data Exchange (ETDEWEB)

Pawlicki, T. [UCSD Medical Center (United States)

2015-06-15

Research related to quality and safety has been a staple of medical physics academic activities for a long time. From very early on, medical physicists have developed new radiation measurement equipment and analysis techniques, created ever increasingly accurate dose calculation models, and have vastly improved imaging, planning, and delivery techniques. These and other areas of interest have improved the quality and safety of radiotherapy for our patients. With the advent of TG-100, quality and safety is an area that will garner even more research interest in the future. As medical physicists pursue quality and safety research in greater numbers, it is worthwhile to consider what actually constitutes research on quality and safety. For example, should the development of algorithms for real-time EPID-based in-vivo dosimetry be defined as “quality and safety” research? How about the clinical implementation of such as system? Surely the application of failure modes and effects analysis to a clinical process would be considered quality and safety research, but is this type of research that should be included in the medical physics peer-reviewed literature? The answers to such questions are of critical importance to set researchers in a direction that will provide the greatest benefit to our field and the patients we serve. The purpose of this symposium is to consider what constitutes research in the arena of quality and safety and differentiate it from other research directions. The key distinction here is developing the tool itself (e.g. algorithms for EPID dosimetry) vs. studying the impact of the tool with some quantitative metric. Only the latter would I call quality and safety research. Issues of ‘basic’ versus ‘applied’ quality and safety research will be covered as well as how the research results should be structured to provide increasing levels of support that a quality and safety intervention is effective and sustainable. Examples from existing

Playground Safety

Science.gov (United States)

... Prevention Fall Prevention Playground Safety Poisoning Prevention Road Traffic Safety Sports Safety Get Email Updates To receive ... at the Consumer Product Safety Commission’s Playground Safety website . References U.S. Consumer Product Safety Commission. Injuries and ...

Biosphere modelling for the safety assessment of high-level radioactive waste disposal in the Japanese H12 assessment

International Nuclear Information System (INIS)

Kato, Tomoko; Suzuki, Yuji; Ishiguro, Katsuhiko; Naito, Morimasa; Ishiguro, Katsuhiko; Ikeda, Takao; Little, Richard H.; Smith, Graham M.

2002-01-01

JNC has an on-going programme of research and development relating to the safety assessment of the deep geological disposal system of high-level radioactive waste (HLW). In the safety assessment of a HLW disposal system, it is often necessary to estimate future radiological impacts on human beings (e.g. radiation dose). In order to estimate dose, consideration needs to be given to the surface environment (biosphere) into which future releases of radionuclides might occur and to the associated future human behaviour. However, for a deep repository, such releases might not occur for many thousands of years after disposal. Over such timescales, it is not possible to predict with any certainty how the biosphere and human behaviour will evolve. To avoid endless speculation aimed at reducing such uncertainty, the reference biosphere le concept has been developed for use in the safety assessment of HLW disposal. The Reference Biospheres Methodology was originally developed by the BIOMOVS II Reference Biospheres Working Group and subsequently enhanced within Theme 1 of the BIOMASS programme. As the aim of the H12 assessment with a hypothetical HLW disposal system was to demonstrate the technical feasibility and reliability of the Japanese disposal concept for a range of geological and surface environments, some assessment specific reference biospheres were developed for the biosphere modelling in the H12 assessment using an approach consistent with the BIOMOVS II/BIOMASS approach. They have been used to derive factors to convert the radionuclide flux from a geosphere to a biosphere into a dose. The influx to dose conversion factor also have been derived for a range of different geosphere-biosphere interfaces (well, river and marine) and potential exposure groups (farming, freshwater-fishing and marine-fishing). This paper summarises the approach used for the derivation of the influx to dose conversion factor also for the range of geosphere-biosphere interfaces and

Safety analysis of urban arterials at the meso level.

Science.gov (United States)

Li, Jia; Wang, Xuesong

2017-11-01

Urban arterials form the main structure of street networks. They typically have multiple lanes, high traffic volume, and high crash frequency. Classical crash prediction models investigate the relationship between arterial characteristics and traffic safety by treating road segments and intersections as isolated units. This micro-level analysis does not work when examining urban arterial crashes because signal spacing is typically short for urban arterials, and there are interactions between intersections and road segments that classical models do not accommodate. Signal spacing also has safety effects on both intersections and road segments that classical models cannot fully account for because they allocate crashes separately to intersections and road segments. In addition, classical models do not consider the impact on arterial safety of the immediately surrounding street network pattern. This study proposes a new modeling methodology that will offer an integrated treatment of intersections and road segments by combining signalized intersections and their adjacent road segments into a single unit based on road geometric design characteristics and operational conditions. These are called meso-level units because they offer an analytical approach between micro and macro. The safety effects of signal spacing and street network pattern were estimated for this study based on 118 meso-level units obtained from 21 urban arterials in Shanghai, and were examined using CAR (conditional auto regressive) models that corrected for spatial correlation among the units within individual arterials. Results showed shorter arterial signal spacing was associated with higher total and PDO (property damage only) crashes, while arterials with a greater number of parallel roads were associated with lower total, PDO, and injury crashes. The findings from this study can be used in the traffic safety planning, design, and management of urban arterials. Copyright © 2017 Elsevier Ltd. All

FLIGHT SAFETY MANAGEMENT PROBLEMS AND EVALUATION OF FLIGHT SAFETY LEVEL OF AN AVIATION ENTERPRISE

OpenAIRE

B. V. Zubkov; H. E. Fourar

2017-01-01

This article is devoted to studying the problem of safety management system (SMS) and evaluating safety level of an aviation enterprise.This article discusses the problems of SMS, presented at the 41st meeting of the Russian Aviation Production Commanders Club in June 2014 in St. Petersburg in connection with the verification of the status of the CA of the Russian Federation by the International Civil Aviation Organization (ICAO) in the same year, a set of urgent measures to eliminate the def...

Panel 1: Safety design criteria

International Nuclear Information System (INIS)

Yllera, Javier

2013-01-01

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

Analysis of Traffic Safety Factors at Level Rail-Road Crossings

Directory of Open Access Journals (Sweden)

Tomislav Mlinarić

2012-10-01

Full Text Available The paper analyses the main factors of traffic safety andreliabilityat level crossings. The number and causes of accidentsare stated, that result from ignorance, insufficient training ofthe traffic participants, their ilnsponsibility and insufficient orincomplete legislation, as well as from insufficiently professionaland scientifically not serious enough approach to solvingthis cardinal problem in road and railway traffic. Based on theanalysis the causes are determined and solutions proposed, aswell as more efficient methods to improve safety and reduce thenumber of traffic accidents at level crossings.

Predisposal management of high level radioactive waste. Safety guide

International Nuclear Information System (INIS)

2005-01-01

, treatment and conditioning) of HLW. (B) The storage of liquid and solidified HLW. (C) The storage of conditioned spent fuel. Unless specific reference is made to one or more categories, the recommendations in this Safety Guide apply generally to all categories of HLW. The possibility of criticality is more significant for spent fuel than it is for other categories of HLW, and should be given appropriate consideration in all activities in which spent fuel is involved. The possibility of criticality involving liquid HLW should, however, also always be considered

Technical reliability of geological disposal for high-level radioactive wastes in Japan. The second progress report. Part 3. Safety assessment for geological disposal systems

International Nuclear Information System (INIS)

1999-11-01

Based on the Advisory Committee Report on Nuclear Fuel Cycle Backend Policy submitted to the Japanese Government in 1997, JNC documents the progress of research and development program in the form of the second progress report (the first one published in 1992). It summarizes an evaluation of the technical reliability and safety of the geological disposal concept for high-level radioactive wastes (HLW) in Japan. The present document, the part 3 of the progress report, concerns safety assessment for geological disposal systems definitely introduced in part 1 and 2 of this series and consists of 9 chapters. Chapter I concerns the methodology for safety assessment while Chapter II deals with diversity and uncertainty about the scenario, the adequate model and the required data of the systems above. Chapter III summarizes the components of the geological disposal system. Chapter IV refers to the relationship between radioactive wastes and human life through groundwater, i.e. nuclide migration. In Chapter V is made a reference case which characterizes the geological environmental data using artificial barrier specifications. (Ohno. S.)

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Application of diagnostic reference levels in medical practice

Energy Technology Data Exchange (ETDEWEB)

Bourguignon, Michel [Faculty of Medicine of Paris, Deputy Director General, Nuclear Safety Authority (ASN), Paris (France)

2006-07-01

Diagnosis reference levels (D.R.L.) are defined in the Council Directive 97/43 EURATOM as 'Dose levels in medical radio diagnosis practices or in the case of radiopharmaceuticals, levels of activity, for typical examinations for groups of standards-sized patients or standards phantoms for broadly defined types of equipment. These levels are expected not to be exceeded for standard procedures when good and normal practice regarding diagnostic and technical performance is applied'. Thus D.R.L. apply only to diagnostic procedures and does not apply to radiotherapy. Radiation protection of patients is based on the application of 2 major radiation protection principles, justification and optimization. The justification principle must be respected first because the best way to protect the patient is not to carry a useless test. Radiation protection of the patient is a continuous process and local dose indicator values in the good range should not prevent the radiologist or nuclear medicine physician to continue to optimize their practice. (N.C.)

Application of diagnostic reference levels in medical practice

International Nuclear Information System (INIS)

Bourguignon, Michel

2006-01-01

Diagnosis reference levels (D.R.L.) are defined in the Council Directive 97/43 EURATOM as 'Dose levels in medical radio diagnosis practices or in the case of radiopharmaceuticals, levels of activity, for typical examinations for groups of standards-sized patients or standards phantoms for broadly defined types of equipment. These levels are expected not to be exceeded for standard procedures when good and normal practice regarding diagnostic and technical performance is applied'. Thus D.R.L. apply only to diagnostic procedures and does not apply to radiotherapy. Radiation protection of patients is based on the application of 2 major radiation protection principles, justification and optimization. The justification principle must be respected first because the best way to protect the patient is not to carry a useless test. Radiation protection of the patient is a continuous process and local dose indicator values in the good range should not prevent the radiologist or nuclear medicine physician to continue to optimize their practice. (N.C.)

Identifying organizational cultures that promote patient safety.

Science.gov (United States)

Singer, Sara J; Falwell, Alyson; Gaba, David M; Meterko, Mark; Rosen, Amy; Hartmann, Christine W; Baker, Laurence

2009-01-01

Safety climate refers to shared perceptions of what an organization is like with regard to safety, whereas safety culture refers to employees' fundamental ideology and orientation and explains why safety is pursued in the manner exhibited within a particular organization. Although research has sought to identify opportunities for improving safety outcomes by studying patterns of variation in safety climate, few empirical studies have examined the impact of organizational characteristics such as culture on hospital safety climate. This study explored how aspects of general organizational culture relate to hospital patient safety climate. In a stratified sample of 92 U.S. hospitals, we sampled 100% of senior managers and physicians and 10% of other hospital workers. The Patient Safety Climate in Healthcare Organizations and the Zammuto and Krakower organizational culture surveys measured safety climate and group, entrepreneurial, hierarchical, and production orientation of hospitals' culture, respectively. We administered safety climate surveys to 18,361 personnel and organizational culture surveys to a 5,894 random subsample between March 2004 and May 2005. Secondary data came from the 2004 American Hospital Association Annual Hospital Survey and Dun & Bradstreet. Hierarchical linear regressions assessed relationships between organizational culture and safety climate measures. Aspects of general organizational culture were strongly related to safety climate. A higher level of group culture correlated with a higher level of safety climate, but more hierarchical culture was associated with lower safety climate. Aspects of organizational culture accounted for more than threefold improvement in measures of model fit compared with models with controls alone. A mix of culture types, emphasizing group culture, seemed optimal for safety climate. Safety climate and organizational culture are positively related. Results support strategies that promote group orientation and

Hearing Tests on Mobile Devices: Evaluation of the Reference Sound Level by Means of Biological Calibration.

Science.gov (United States)

Masalski, Marcin; Kipiński, Lech; Grysiński, Tomasz; Kręcicki, Tomasz

2016-05-30

Hearing tests carried out in home setting by means of mobile devices require previous calibration of the reference sound level. Mobile devices with bundled headphones create a possibility of applying the predefined level for a particular model as an alternative to calibrating each device separately. The objective of this study was to determine the reference sound level for sets composed of a mobile device and bundled headphones. Reference sound levels for Android-based mobile devices were determined using an open access mobile phone app by means of biological calibration, that is, in relation to the normal-hearing threshold. The examinations were conducted in 2 groups: an uncontrolled and a controlled one. In the uncontrolled group, the fully automated self-measurements were carried out in home conditions by 18- to 35-year-old subjects, without prior hearing problems, recruited online. Calibration was conducted as a preliminary step in preparation for further examination. In the controlled group, audiologist-assisted examinations were performed in a sound booth, on normal-hearing subjects verified through pure-tone audiometry, recruited offline from among the workers and patients of the clinic. In both the groups, the reference sound levels were determined on a subject's mobile device using the Bekesy audiometry. The reference sound levels were compared between the groups. Intramodel and intermodel analyses were carried out as well. In the uncontrolled group, 8988 calibrations were conducted on 8620 different devices representing 2040 models. In the controlled group, 158 calibrations (test and retest) were conducted on 79 devices representing 50 models. Result analysis was performed for 10 most frequently used models in both the groups. The difference in reference sound levels between uncontrolled and controlled groups was 1.50 dB (SD 4.42). The mean SD of the reference sound level determined for devices within the same model was 4.03 dB (95% CI 3

OSH technical reference manual

Energy Technology Data Exchange (ETDEWEB)

1993-11-01

In an evaluation of the Department of Energy (DOE) Occupational Safety and Health programs for government-owned contractor-operated (GOCO) activities, the Department of Labor`s Occupational Safety and Health Administration (OSHA) recommended a technical information exchange program. The intent was to share written safety and health programs, plans, training manuals, and materials within the entire DOE community. The OSH Technical Reference (OTR) helps support the secretary`s response to the OSHA finding by providing a one-stop resource and referral for technical information that relates to safe operations and practice. It also serves as a technical information exchange tool to reference DOE-wide materials pertinent to specific safety topics and, with some modification, as a training aid. The OTR bridges the gap between general safety documents and very specific requirements documents. It is tailored to the DOE community and incorporates DOE field experience.

Forget the Desk Job: Current Roles and Responsibilities in Entry-Level Reference Job Advertisements

Science.gov (United States)

Detmering, Robert; Sproles, Claudene

2012-01-01

This study examines the evolving roles and responsibilities of entry-level academic reference positions, as stated in recent job advertisements posted on the American Library Association's JobLIST Web site and other sources. Findings from a content analysis of these advertisements indicate that current entry-level reference positions in academic…

Nuclear power: levels of safety

International Nuclear Information System (INIS)

Lidsky, L.M.

1988-01-01

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the nuclear establishment itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired

Reference dose levels for dental panoramic radiography in Gwangju (South Korea)

International Nuclear Information System (INIS)

Lee, J. S.; Kim, Y. H.; Yoon, S. J.; Kang, B. C.

2010-01-01

This study assessed the reference dose levels for dental panoramic radiography in Gwangju city (South Korea) based on the dose width product (DWP) and compared them with those already established elsewhere. A total of 44 panoramic dental radiographic sets (36 digital and 8 analogue panoramic sets) in 41 dental clinics in Gwangju city were chosen. The third quartile DWP was determined from 429 surface dose measurements of the adult surface dose in panoramic dental radiography. The third quartile DWP for panoramic radiography was 60.1 mGy mm. The proposed DWP reference levels of 60.1 mGy mm were less than or equal to those previously reported in other countries, such as Italy and UK, and acceptable for panoramic radiography in Gwangju (KR). (authors)

Deepening om Safety culture Auto evaluation

International Nuclear Information System (INIS)

Lopez Churruca, I.; Buedo Jimenez, J. L.

2009-01-01

The concept of safety culture used in nuclear forums refers to the series of actions aimed at guaranteeing that safety issues in nuclear power plants are adequately addressed. The activities to which this concept refers have been gradually extended over the years so that, today, they encompass the whole organizations structure. In other words, the safety culture implies that all positions in the organizational structure perform their tasks with a level of attention such that all their senses are focused on them. And this performance is what leads us to excellence in plant operation. In addition, the implementation of a self-assessment system in the Cofrentes Nuclear Power Plant of its activities and processes has resulted in the identification of expectations on the basis of which we can identify strengths and weaknesses, enabling us to leverage the former and correct the latter. (Author)

Safety of geologic disposal of high level radioactive waste

International Nuclear Information System (INIS)

Zaitsu, Tomohisa; Ishiguro, Katsuhiko; Masuda, Sumio

1992-01-01

This article introduces current concepts of geologic disposal of high level radioactive waste and its safety. High level radioactive waste is physically stabilized by solidifying it in a glass form. Characteristics of deep geologic layer are presented from the viewpoint of geologic disposal. Reconstruction of multi-barrier system receives much attention to secure the safety of geologic disposal. It is important to research performance assessment of multi-barrier system for preventing dissolution or transfer of radionuclides into the ground water. Physical and chemical modeling for the performance assessment is outlined in the following terms: (1) chemical property of deep ground water, (2) geochemical modeling of artificial barrier spatial water, (3) hydrology of deep ground water, (4) hydrology of the inside of artificial barrier, and (5) modeling of radionuclide transfer from artificial barrier. (N.K.)

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

International Nuclear Information System (INIS)

1998-01-01

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

Optimum Safety Levels and Design Rules for the Icelandic Type Berm Breakwater

DEFF Research Database (Denmark)

Sigurdarson, Sigurdur; van der Meer, Jentsje W.; Burcharth, Hans F.

2007-01-01

Guidance on selection of breakwater types and related design safety levels for breakwaters are almost non-existent, which is the reason that PIANC has initiated working group 47 on this subject. This paper presents ongoing work particulary on the Icelandic type berm breakwater within the PIANC...... working group. It will concentrate on design guidance and on the optimum safety levels for this type of structure....

The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

International Nuclear Information System (INIS)

1992-01-01

The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

Predisposal Management of Low and Intermediate Level Radioactive Waste. Safety Guide

International Nuclear Information System (INIS)

2009-01-01

The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established for the predisposal management of low and intermediate level waste. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. General safety considerations; 5. Safety features for the predisposal management of LILW; 6. Record keeping and reporting; 7. Safety assessment; 8. Quality assurance; Annex I: Nature and sources of LILW from nuclear facilities; Annex II: Development of specifications for waste packages; Annex III: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Postulated initiating events for consideration in a safety assessment (internal phenomena).

Method of evaluation of diagnostics reference levels in computerized tomography

International Nuclear Information System (INIS)

Vega, Walter Flores

1999-04-01

Computerized tomography is a complex technique with several selectable exposition parameters delivering high doses to the patient. In this work it was developed a simple methodology to evaluate diagnostic reference levels in computerized tomography, using the concept of Multiple Scan Average Dose (MSAD), recently adopted by the Health Ministry. For evaluation of the MSAD, a dose distribution was obtained through a measured dose profile on the axial axis of a water phantom with thermoluminescence dosemeters, TLD-100, for different exam technique. The MSAD was evaluated hrough two distinct methods. First, it was evaluated by the integration of the dose profile of a single slice and, second, obtained by the integration on central slice of the profile of several slices. The latter is in of accordance with the ionization chamber method, suggesting to be the most practical method of dose evaluation to be applied in the diagnostic reference level assessment routine for CT, using TLDs. (author)

Determination of toxic trace elements in body fluid reference samples

International Nuclear Information System (INIS)

Gills, T.E.; McClendon, L.T.; Maienthal, E.J.; Becker, D.A.; Durst, R.A.; LaFleur, P.D.

1974-01-01

The measurement of elemental concentration in body fluids has been widely used to give indication of exposures to certain toxic materials and/or a measure of body burden. To understand fully the toxicological effect of these trace elements on our physiological system, meaningful analytical data are required along with accurate standards or reference samples. The National Bureau of Standards has prepared for the National Institute for Occupational Safety and Health (NIOSH) a number of reference samples containing selected toxic trace elements in body fluids. The reference samples produced include mercury in urine at three concentration levels, five elements (Se, Cu, As, Ni and Cr) in freeze-dried urine at two levels, fluorine in freeze-dried urine at two levels and lead in blood at two concentration levels. These reference samples have been found to be extremely useful for the evaluation of field and laboratory analytical methods for the analysis of toxic trace elements. In particular the use of at least two calibration points (i.e., ''normal'' and ''elevated'' levels) for a given matrix provides a more positive calibration for most analytical techniques over the range of interest for occupational toxicological levels of exposure. (U.S.)

Patient dose with quality image under diagnostic reference levels

International Nuclear Information System (INIS)

Akula, Suresh Kumar; Singh, Gurvinder; Chougule, Arun

2016-01-01

Need to set Diagnostic Reference Level (DRL) for locations for all diagnostic procedures in local as compared to National. The review of DRL's should compare local with national or referenced averages and a note made of any significant variances to these averages and the justification for it. To survey and asses radiation doses to patient and reduce the redundancy in patient imaging to maintain DRLs

The establishment of local diagnostic reference levels for paediatric interventional cardiology

International Nuclear Information System (INIS)

McFadden, S.; Hughes, C.; D'Helft, C.I.; McGee, A.; Rainford, L.; Brennan, P.C.; McCrum-Gardner, E.; Winder, R.J.

2013-01-01

Background: There is a paucity of information worldwide on radiation exposure in paediatric interventional cardiology. At present Nationally established Diagnostic Reference Levels exist for adult interventional cardiology procedures in the UK but little data is available for paediatrics. In addition, interventional cardiology has been identified as one the highest contributors to medical exposure to ionising radiation and children are more radiosensitive than adults. Objective: This study sought to determine current radiation dose levels in paediatric interventional cardiology (IC) with a view to establishing local diagnostic reference levels (LDRL). Methods: Radiation dose and examination details were recorded for 354 paediatric patients examined by IC in a specialised paediatric centre in Europe. Radiation doses were recorded using a Dose Area Product meter along with examination details. Procedures were categorised as either diagnostic (A) or therapeutic (B). Data was further sub-divided into five age ranges; (1) newborn <1 year (2) 1 <5 years (3) 5 <10 years (4) 10 <15 years (5) 15 years and over. Proposed LDRL were calculated from the mean dose area product readings. Results: The mean patient age was 2.6 years (range 0.0 days–16 years) and weight was 14.9 kg (range 2.4–112 kg). LDRL for the five age groupings were calculated as 190, 421, 582, 1289 and 1776 cGycm² respectively. Conclusion: Local dose reference levels have been proposed for paediatric IC and can be used as a benchmark for other hospitals to compare against their own radiation doses

Predisposal management of low and intermediate level radioactive waste. Safety guide

International Nuclear Information System (INIS)

2003-01-01

considered in this publication begins with the refining and conversion of uranium concentrates. Recommendations on the management of radioactive waste from the mining and milling of uranium and thorium ores are provided. Some parts of the nuclear fuel cycle generate both high level waste and LLW. The management of high level waste itself generates LLW. The predisposal management of this LLW is included in the scope of this Safety Guide. Recommendations on the predisposal management of high level waste are provided. The recommendations in this Safety Guide primarily concern complex management activities for LLW. The regulatory body should decide which parts of this Safety Guide are relevant and appropriate for particular circumstances, and the extent to which the recommendations and guidance apply. This Safety Guide provides only introductory material on the transport and storage of LLW. Requirements and recommendations are provided in Refs. There may be non-radiological hazards associated with the predisposal management of LLW. Some guidance is given on the safety measures to be taken against non-radiological hazards if they have potential consequences for radiation safety. However, detailed recommendations are beyond the scope of this Safety Guide. The user should seek guidance from the regulatory body in the areas of health and safety and environmental protection

Diagnostic Reference Levels in the 1990 and 1996 Recommendations of the ICRP (invited paper)

International Nuclear Information System (INIS)

Drexler, G.

1998-01-01

A review of and some comments on the paragraphs in the ICRP Publications 60 and 73 are presented, which are relevant to diagnostic reference doses. The content of the statements is traced back by approximately 50 years when ICRP's preoccupation with the future health and well-being of the population is reflected in guidance for characterisation of 'normal operational conditions'. The early ICRP levels of reference doses are compared with the values currently discussed to demonstrate the importance of diagnostic reference doses in the process of optimisation and to show the importance of a continuous review and update of these levels. (author)

Statistical re-evaluation of the ASME KIC and KIR fracture toughness reference curves

International Nuclear Information System (INIS)

Wallin, K.

1999-01-01

Historically the ASME reference curves have been treated as representing absolute deterministic lower bound curves of fracture toughness. In reality, this is not the case. They represent only deterministic lower bound curves to a specific set of data, which represent a certain probability range. A recently developed statistical lower bound estimation method called the 'master curve', has been proposed as a candidate for a new lower bound reference curve concept. From a regulatory point of view, the master curve is somewhat problematic in that it does not claim to be an absolute deterministic lower bound, but corresponds to a specific theoretical failure probability that can be chosen freely based on application. In order to be able to substitute the old ASME reference curves with lower bound curves based on the master curve concept, the inherent statistical nature (and confidence level) of the ASME reference curves must be revealed. In order to estimate the true inherent level of safety, represented by the reference curves, the original database was re-evaluated with statistical methods and compared to an analysis based on the master curve concept. The analysis reveals that the 5% lower bound master curve has the same inherent degree of safety as originally intended for the K IC -reference curve. Similarly, the 1% lower bound master curve corresponds to the K IR -reference curve. (orig.)

Statistical re-evaluation of the ASME KIC and KIR fracture toughness reference curves

International Nuclear Information System (INIS)

Wallin, K.; Rintamaa, R.

1998-01-01

Historically the ASME reference curves have been treated as representing absolute deterministic lower bound curves of fracture toughness. In reality, this is not the case. They represent only deterministic lower bound curves to a specific set of data, which represent a certain probability range. A recently developed statistical lower bound estimation method called the 'Master curve', has been proposed as a candidate for a new lower bound reference curve concept. From a regulatory point of view, the Master curve is somewhat problematic in that it does not claim to be an absolute deterministic lower bound, but corresponds to a specific theoretical failure probability that can be chosen freely based on application. In order to be able to substitute the old ASME reference curves with lower bound curves based on the master curve concept, the inherent statistical nature (and confidence level) of the ASME reference curves must be revealed. In order to estimate the true inherent level of safety, represented by the reference curves, the original data base was re-evaluated with statistical methods and compared to an analysis based on the master curve concept. The analysis reveals that the 5% lower bound Master curve has the same inherent degree of safety as originally intended for the K IC -reference curve. Similarly, the 1% lower bound Master curve corresponds to the K IR -reference curve. (orig.)

The Influence of the Terrestrial Reference Frame on Studies of Sea Level Change

Science.gov (United States)

Nerem, R. S.; Bar-Sever, Y. E.; Haines, B. J.; Desai, S.; Heflin, M. B.

2015-12-01

The terrestrial reference frame (TRF) provides the foundation for the accurate monitoring of sea level using both ground-based (tide gauges) and space-based (satellite altimetry) techniques. For the latter, tide gauges are also used to monitor drifts in the satellite instruments over time. The accuracy of the terrestrial reference frame (TRF) is thus a critical component for both types of sea level measurements. The TRF is central to the formation of geocentric sea-surface height (SSH) measurements from satellite altimeter data. The computed satellite orbits are linked to a particular TRF via the assumed locations of the ground-based tracking systems. The manner in which TRF errors are expressed in the orbit solution (and thus SSH) is not straightforward, and depends on the models of the forces underlying the satellite's motion. We discuss this relationship, and provide examples of the systematic TRF-induced errors in the altimeter derived sea-level record. The TRF is also crucial to the interpretation of tide-gauge measurements, as it enables the separation of vertical land motion from volumetric changes in the water level. TRF errors affect tide gauge measurements through GNSS estimates of the vertical land motion at each tide gauge. This talk will discuss the current accuracy of the TRF and how errors in the TRF impact both satellite altimeter and tide gauge sea level measurements. We will also discuss simulations of how the proposed Geodetic Reference Antenna in SPace (GRASP) satellite mission could reduce these errors and revolutionize how reference frames are computed in general.

Thoughts of the nuclear safety culture and 'star-level' management

International Nuclear Information System (INIS)

Wang Sen

2004-01-01

From the point of view that enterprise management has come into the stage of cultural management, this article divides the contents of nuclear safety culture into target management, safety management, quality management, site management, cost management, authority management, teamwork, information communication and continuous improvement. Each aspect win be classified by five 'star- level's according to the appearance, and the present situation of the company should be assessed with those star-level indices so as to find out the disadvantages. Improvement will follow with the promotion of company management level. (author)

A New Dual-purpose Quality Control Dosimetry Protocol for Diagnostic Reference-level Determination in Computed Tomography.

Science.gov (United States)

Sohrabi, Mehdi; Parsi, Masoumeh; Sina, Sedigheh

2018-05-17

A diagnostic reference level is an advisory dose level set by a regulatory authority in a country as an efficient criterion for protection of patients from unwanted medical exposure. In computed tomography, the direct dose measurement and data collection methods are commonly applied for determination of diagnostic reference levels. Recently, a new quality-control-based dose survey method was proposed by the authors to simplify the diagnostic reference-level determination using a retrospective quality control database usually available at a regulatory authority in a country. In line with such a development, a prospective dual-purpose quality control dosimetry protocol is proposed for determination of diagnostic reference levels in a country, which can be simply applied by quality control service providers. This new proposed method was applied to five computed tomography scanners in Shiraz, Iran, and diagnostic reference levels for head, abdomen/pelvis, sinus, chest, and lumbar spine examinations were determined. The results were compared to those obtained by the data collection and quality-control-based dose survey methods, carried out in parallel in this study, and were found to agree well within approximately 6%. This is highly acceptable for quality-control-based methods according to International Atomic Energy Agency tolerance levels (±20%).

Methodology of safety evaluation about land disposal of low level radioactive wastes

International Nuclear Information System (INIS)

Suzuki, Atsuyuki

1986-01-01

Accompanying the progress of the construction project of low level radioactive waste storage facilities in Aomori Prefecture, the full scale land disposal of low level radioactive wastes shows its symptom also in Japan. In this report, the scientific methodology to explain the safety about the land disposal of low level radioactive wastes is discussed. The land disposal of general wastes by shallow burying has already had sufficient results. In the case of low level radioactive wastes, also the land disposal by shallow burying is considered. Low level radioactive wastes can be regarded as one form of industrial wastes, as there are many common parts in the scientific and theoretical base of the safety. Attention is paid most to the contamination of ground water. Low level radioactive wastes are solid wastes, accordingly the degree of contamination should be less. The space in which ground water existes, the phenomena of ground water movement, the phenomena of ground water dispersion and Fick's law, the adsorption effect of strata, and the evaluation of source term are explained. These are the method to analyze the degree of contamination from safety evaluation viewpoint. (Kako, I.)

Basic concept on safety regulation for land disposal of low level radioactive solid wastes

International Nuclear Information System (INIS)

1985-01-01

As to the land disposal of low level radioactive solid wastes, to which the countermeasures have become the urgent problem at present, it is considered to be a realistic method to finally store the solid wastes concentratedly outside the sites of nuclear power stations and others, and effort has been exerted by those concerned to realize it. Besides, as for extremely low level radioactive solid wastes, the measures of disposing them corresponding to the radioactivity level are necessary, and the concrete method has been examined. The Committee on Safety Regulation for Radioactive Wastes has discussed the safety regulation for those since April, 1984, and the basic concept on the safety regulation was worked up. It is expected that the safety of the land disposal of low level radioactive solid wastes can be ensured when the safety regulation is carried out in conformity with this basic concept. The present status of the countermeasures to the land disposal of low level radioactive solid wastes is shown. As the concrete method, the disposal in shallow strate has been generally adopted. At present, the plan for the final storage in Aomori Prefecture is considered, and it will be started with the first stage of four-stage control. (Kako, I.)

MP-Division health and safety reference handbook

International Nuclear Information System (INIS)

Putnam, T.M.

1987-09-01

This report presents the objectives, organization, policies, and essential rules and procedures that have been adopted by MP Division and that form the basis of the Health and Safety Program of the Clinton P. Anderson Meson Physics Facility (LAMPF). The facility includes the beam-delivery systems for the Los Alamos Neutron Scattering Center and the Weapons Neutron Research Facility (LANSCE/WNR). The program is designed not only to assure the health and safety of all personnel, including users, in their work at LAMPF, and of MP-Division staff in their work on the LANSCE/WNR beam lines, but also to protect the facility (buildings and equipment) and the environment. 33 refs., 18 figs., 2 tabs

Patient dose measurements in fluoroscopic examinations, aiming to the establishment of reference levels in Brazil

International Nuclear Information System (INIS)

Canevaro, L.; Drexler, G.

2001-01-01

This work was performed to investigate the actual exposure levels of the patients submitted to fluoroscopic procedures in diagnostic radiology. The data will be useful for a baseline in the establishment of local reference levels for fluoroscopic procedures, as recommended by the European Commission and IAEA. At present time there are no internationally accepted definitions for references levels for fluoroscopic complex procedures. Dose-area product (DAP) meters were employed in a pilot survey expressing the radiation exposures in terms of this quantity. This class of instrumentation has not yet been employed in Brazil. Parameters recorded were radiographic technique, fluoroscopy time, number of images, fluoroscopic and radiographic field sizes and DAPs. For fluoroscopy practice, a reference parameters set is recommended, instead of one diagnostic reference level. High patient exposures were found, calling for joined actions of health authorities, physicians, medical physicists, technicians and manufacturers. Monitoring of patient exposure, optimizing the radiation protection and establishing quantitative assessments of the exposition to the population in Brazil in this kind of procedure is important. (author)

Operating safety requirements for the intermediate level liquid waste system

International Nuclear Information System (INIS)

1980-07-01

The operation of the Intermediate Level Liquid Waste (ILW) System, which is described in the Final Safety Analysis, consists of two types of operations, namely: (1) the operation of a tank farm which involves the storage and transportation through pipelines of various radioactive liquids; and (2) concentration of the radioactive liquids by evaporation including rejection of the decontaminated condensate to the Waste Treatment Plant and retention of the concentrate. The following safety requirements in regard to these operations are presented: safety limits and limiting control settings; limiting conditions for operation; and surveillance requirements. Staffing requirements, reporting requirements, and steps to be taken in the event of an abnormal occurrence are also described

Behavioral integrity for safety, priority of safety, psychological safety, and patient safety : a team-level study

NARCIS (Netherlands)

Leroy, H.; Dierynck, B.; Anseel, F.; Simons, T.; Halbesleben, J.R.; McCaughey, D.; Savage, G.T.; Sels, L.

2012-01-01

This article clarifies how leader behavioral integrity for safety helps solve follower's double bind between adhering to safety protocols and speaking up about mistakes against protocols. Path modeling of survey data in 54 nursing teams showed that head nurse behavioral integrity for safety

Investigation of reference levels and radiation dose associated with abdominal EVAR (endovascular aneurysm repair) procedures across several European Centres

Energy Technology Data Exchange (ETDEWEB)

Tuthill, E.; Rainford, L. [University College Dublin, Diagnostic Imaging, School of Medicine, Dublin (Ireland); O' Hora, L.; O' Donohoe, M. [Mater Misericordiae University Hospital, Dublin (Ireland); Panci, S. [San Giovanni di Dio Hospital, Florence (Italy); Gilligan, P.; Fox, E. [Mater Private Hospital, Dublin (Ireland); Campion, D. [Mauriziano-Umberto Hospital, Turin (Italy); Trenti, R. [Policlinico S. Orsola-Malpighi, Bologna (Italy); Catania, D. [AITRI, Association of Italian Interventional Radiographers, Milan (Italy)

2017-11-15

Endovascular aneurysm repair (EVAR) is considered the treatment of choice for abdominal aortic aneurysms with suitable anatomy. In order to improve radiation safety, European Directive (2013/59) requires member states to implement diagnostic reference levels (DRLs) in radio-diagnostic and interventional procedures. This study aimed to determine local DRLs for EVAR across five European centres and identify an interim European DRL, which currently remains unestablished. Retrospective data was collected for 180 standard EVARs performed between January 2014 and July 2015 from five specialist centres in Ireland (n=2) and Italy (n=3). Data capture included: air kerma-area product (P{sub KA}), total air kerma at the reference point (K{sub a,r}), fluoroscopic time (FT), number of acquisitions, frame rate of acquisition, type of acquisition, patient height, weight, and gender. The mean values for each site A, B, C, D, and E were: P{sub KA}s of 4343 ± 994 μGym{sup 2}, 18,200 ± 2141 μGym{sup 2}, 11,423 ± 1390 μGym{sup 2}, 7796 ± 704 μGym{sup 2}, 31,897 ± 5798 μGym{sup 2}; FTs of 816 ± 92 s, 950 ± 150 s, 708 ± 70 s, 972 ± 61 s, 827 ± 118 s; and number of acquisitions of 6.72 ± 0.56, 10.38 ± 1.54, 4.74 ± 0.19, 5.64 ± 0.36, 7.28 ± 0.65, respectively. The overall pooled 75th percentile P{sub KA} was 15,849 μGym{sup 2}. Local reference levels were identified. The pooled data has been used to establish an interim European DRL for EVAR procedures. (orig.)

Synthesis of the IRSN report on severe accidents and level 2 probabilistic safety studies within the frame of the safety re-examination associated with the third decennial inspection of 1300 MW reactors

International Nuclear Information System (INIS)

2013-01-01

The objective of this report is to analyze studies related to severe accidents and performed within the framework of the third decennial safety re-examination of the French 1300 We nuclear reactors. It also reports the main conclusions of a detailed analysis of level-2 probabilistic safety studies performed according to another procedure. The report first addresses the 'severe accident' system of reference. It presents the general approach and the safety objectives, discusses the management of a site with a unit in severe accident (this encompasses the management of neighbouring units, the conditions of intervention in terms of habitability of the control room and of manoeuvrability of the venting-filtration system), discusses the expected equipment performance (concerned equipment, safety requirements for equipment needed in case of severe accident, loadings). A second part addresses and comments the results of level 2 probabilistic studies. The report then addresses the water management in the vessel sink with two main objectives (to keep corium in the vessel while promoting its cooling, to cool corium fallen in the vessel sink). The next part addresses modifications planned by EDF in terms of instrumentation associated with a severe accident situation, of improvement of confinement and reduction of risks of important and early releases, of enclosure depressurization in case of unavailability of the enclosure sprinkling system, and of strategy of opening the venting-filtration device in case of total loss of electricity supplies

Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

International Nuclear Information System (INIS)

Almkvist, Lisa; Gordon, Ann

2007-11-01

The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements of 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

Comparing performance level estimation of safety functions in three distributed structures

International Nuclear Information System (INIS)

Hietikko, Marita; Malm, Timo; Saha, Heikki

2015-01-01

The capability of a machine control system to perform a safety function is expressed using performance levels (PL). This paper presents the results of a study where PL estimation was carried out for a safety function implemented using three different distributed control system structures. Challenges relating to the process of estimating PLs for safety related distributed machine control functions are highlighted. One of these examines the use of different cabling schemes in the implementation of a safety function and its effect on the PL evaluation. The safety function used as a generic example in PL calculations relates to a mobile work machine. It is a safety stop function where different technologies (electrical, hydraulic and pneumatic) can be utilized. It was detected that by replacing analogue cables with digital communication the system structure becomes simpler with less number of failing components, which can better the PL of the safety function. - Highlights: • Integration in distributed systems enables systems with less components. • It offers high reliability and diagnostic properties. • Analogue signals create uncertainty in signal reliability and difficult diagnostics

Safety Training: places available in October 2014

CERN Multimedia

2014-01-01

There are places available in the forthcoming Safety courses. For updates and registrations, please refer to the Safety Training Catalogue. Safety Training, HSE Unit safety-training@cern.ch Title of the course EN Title of the course FR Date Hours Language Chemical Safety ATEX Habilitation - Level 2 Habilitation ATEX - Niveau 2 16-Oct-14 to 17-Oct-14 9:00 - 17:30 French Cryogenic Safety Cryogenic Safety - Fundamentals Sécurité Cryogénie - Fondamentaux 23-Oct-14 10:00 - 12:00 English Cryogenic Safety - Helium Transfer Sécurité Cryogénie - Transfert d'hélium 30-Oct-14 9:30 - 12:00 English Electrical Safety Habilitation Electrique - Electrician Low Voltage - Initial Habilitation électrique - Électricien basse tension - Initial 02-Oct-14 to 06-Oct-14 9:00 - 17:30 English 20-Oct-14 to 22-Oct-14 9:00 -...

Study on the development of safety regulations for geological disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Wei Fangxin

2012-01-01

The development of regulations under Regulations on Safety Management of Radioactive Waste has become necessary as the issuance of it. The regulations related to geological disposal of high-level radioactive waste can promote the progress of research and development on geological disposal of high-level radioactive waste in China. This paper has present suggestions on development of regulations on geological disposal of high-level radioactive waste by analyzing development of safety regulations on geological disposal of high-level radioactive waste in foreign countries and problems occurred in China and discussed important issues related to the development of safety regulations on geological disposal of high-level radioactive waste. (author)

Screening-Level Safety Assessment of Personal Care Product Constituents Using Publicly Available Data

Directory of Open Access Journals (Sweden)

Ernest S. Fung

2018-06-01

Full Text Available Organizations recommend evaluating individual ingredients when assessing the safety of personal care or cosmetic products. The goal of this study was to present a screening-level safety assessment methodology to evaluate the safety of a product by identifying individual ingredients, determining their frequency of use in on-market products, and examining published safe-level-of-use information for each ingredient. As a case study, we evaluated WEN by Chaz Dean (WCD cleansing conditioners since there have been claims of adverse health effects associated with product use. We evaluated 30 ingredients in three on-market WCD cleansing conditioners. We then analyzed the National Library of Medicine’s Household Products Database and the Environmental Working Group’s (EWG Skin Deep Cosmetic Database, two of the largest publicly available databases, for other on-market personal care and cosmetic products that contained these ingredients. Safe-level-of-use information for each ingredient was obtained by reviewing peer-reviewed literature, the Food and Drug Administration’s (FDA generally recognized as safe (GRAS database, available Cosmetic Ingredient Review (CIR publications, and available product safety publications. The results of this analysis showed that more than 20,000 personal care and cosmetic products contained one or more of the evaluated ingredients used in WCD cleaning conditioners. Published safety information was available for 21 of the 30 evaluated ingredients: seven identified ingredients were designated as GRAS by the FDA and 16 ingredients had safe-level-of-use information available from the CIR. This study presents a screening-level safety assessment methodology that can serve as an initial screening tool to evaluate the safety of an ingredient intended for use in personal care and cosmetic products before a product is launched onto the market. This study provides evidence that the evaluated WCD cleansing conditioner ingredients

SU-E-P-10: Establishment of Local Diagnostic Reference Levels of Routine Exam in Computed Tomography

International Nuclear Information System (INIS)

Yeh, M; Wang, Y; Weng, H

2015-01-01

Introduction National diagnostic reference levels (NDRLs) can be used as a reference dose of radiological examination can provide radiation dose as the basis of patient dose optimization. Local diagnostic reference levels (LDRLs) by periodically view and check doses, more efficiency to improve the way of examination. Therefore, the important first step is establishing a diagnostic reference level. Computed Tomography in Taiwan had been built up the radiation dose limit value,in addition, many studies report shows that CT scan contributed most of the radiation dose in different medical. Therefore, this study was mainly to let everyone understand DRL’s international status. For computed tomography in our hospital to establish diagnostic reference levels. Methods and Materials: There are two clinical CT scanners (a Toshiba Aquilion and a Siemens Sensation) were performed in this study. For CT examinations the basic recommended dosimetric quantity is the Computed Tomography Dose Index (CTDI). Each exam each different body part, we collect 10 patients at least. Carried out the routine examinations, and all exposure parameters have been collected and the corresponding CTDIv and DLP values have been determined. Results: The majority of patients (75%) were between 60–70 Kg of body weight. There are 25 examinations in this study. Table 1 shows the LDRL of each CT routine examination. Conclusions: Therefore, this study would like to let everyone know DRL’s international status, but also establishment of computed tomography of the local reference levels for our hospital, and providing radiation reference, as a basis for optimizing patient dose

SU-E-P-10: Establishment of Local Diagnostic Reference Levels of Routine Exam in Computed Tomography

Energy Technology Data Exchange (ETDEWEB)

Yeh, M; Wang, Y; Weng, H [Chiayi Chang Gung Memorial Hospital of The C.G.M.F, Puzi City, Chiayi County, Taiwan (China)

2015-06-15

Introduction National diagnostic reference levels (NDRLs) can be used as a reference dose of radiological examination can provide radiation dose as the basis of patient dose optimization. Local diagnostic reference levels (LDRLs) by periodically view and check doses, more efficiency to improve the way of examination. Therefore, the important first step is establishing a diagnostic reference level. Computed Tomography in Taiwan had been built up the radiation dose limit value,in addition, many studies report shows that CT scan contributed most of the radiation dose in different medical. Therefore, this study was mainly to let everyone understand DRL’s international status. For computed tomography in our hospital to establish diagnostic reference levels. Methods and Materials: There are two clinical CT scanners (a Toshiba Aquilion and a Siemens Sensation) were performed in this study. For CT examinations the basic recommended dosimetric quantity is the Computed Tomography Dose Index (CTDI). Each exam each different body part, we collect 10 patients at least. Carried out the routine examinations, and all exposure parameters have been collected and the corresponding CTDIv and DLP values have been determined. Results: The majority of patients (75%) were between 60–70 Kg of body weight. There are 25 examinations in this study. Table 1 shows the LDRL of each CT routine examination. Conclusions: Therefore, this study would like to let everyone know DRL’s international status, but also establishment of computed tomography of the local reference levels for our hospital, and providing radiation reference, as a basis for optimizing patient dose.

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

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.

Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks

Energy Technology Data Exchange (ETDEWEB)

ROGERS, C.A.

2000-02-17

This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

Criticality Safety Evaluation of Hanford Site High-Level Waste Storage Tanks

International Nuclear Information System (INIS)

ROGERS, C.A.

2000-01-01

This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions

The condition and the dynamics of changes of regional energetic safety level

OpenAIRE

Anatoliy Myzin; Aleksey Kalina; Andrey Kozitsyn; Pavel Pykhov

2006-01-01

On the basis of indicative analysis method use, the dynamic processes of changes of energetic safety condition of federal districts and subjects of Russian Federation for last 5 years are investigated. The results of diagnosing safety levels for separate indicators, their blocks and the results of situation evaluation as a whole are discussed. The comparison of regions’ energetic safety condition is given, the causes of crisis situations appearance are discovered, and on this basis the sugg...

Macro-level safety analysis of pedestrian crashes in Shanghai, China.

Science.gov (United States)

Wang, Xuesong; Yang, Junguang; Lee, Chris; Ji, Zhuoran; You, Shikai

2016-11-01

Pedestrian safety has become one of the most important issues in the field of traffic safety. This study aims at investigating the association between pedestrian crash frequency and various predictor variables including roadway, socio-economic, and land-use features. The relationships were modeled using the data from 263 Traffic Analysis Zones (TAZs) within the urban area of Shanghai - the largest city in China. Since spatial correlation exists among the zonal-level data, Bayesian Conditional Autoregressive (CAR) models with seven different spatial weight features (i.e. (a) 0-1 first order, adjacency-based, (b) common boundary-length-based, (c) geometric centroid-distance-based, (d) crash-weighted centroid-distance-based, (e) land use type, adjacency-based, (f) land use intensity, adjacency-based, and (g) geometric centroid-distance-order) were developed to characterize the spatial correlations among TAZs. Model results indicated that the geometric centroid-distance-order spatial weight feature, which was introduced in macro-level safety analysis for the first time, outperformed all the other spatial weight features. Population was used as the surrogate for pedestrian exposure, and had a positive effect on pedestrian crashes. Other significant factors included length of major arterials, length of minor arterials, road density, average intersection spacing, percentage of 3-legged intersections, and area of TAZ. Pedestrian crashes were higher in TAZs with medium land use intensity than in TAZs with low and high land use intensity. Thus, higher priority should be given to TAZs with medium land use intensity to improve pedestrian safety. Overall, these findings can help transportation planners and managers understand the characteristics of pedestrian crashes and improve pedestrian safety. Copyright © 2016 Elsevier Ltd. All rights reserved.

Food safety

Science.gov (United States)

... safety URL of this page: //medlineplus.gov/ency/article/002434.htm Food safety To use the sharing features on this page, please enable JavaScript. Food safety refers to the conditions and practices that preserve the quality of food. These practices prevent contamination and foodborne ...

The need for national diagnostic reference levels: entrance surface dose measurement in intraoral radiography

International Nuclear Information System (INIS)

Mortazavi, S. M. J.; Shareghi, A.; Kavousi, A.; Ghiassi-Nejad, M.; Jafari-Zadeh, M.; Nazeri, F.; Mozdarani, H.

2004-01-01

Background: Intraoral radiographies are the most frequent X-ray examinations in humans. According to International Commission on Radiation Protection recommendations, the selection of a diagnostic reference level should be specific to a country or region. Critical organs such as thyroid gland are exposed to X-rays in intraoral radiography and these exposures should be kept as low as reasonably achievable. To assist the development of DRLs for intraoral radiography, a National Radiation Protection Department-sponsored pilot study was carried out. Materials and methods: thermoluminescent dosimetry is widely acknowledged to be the recommended method for measuring entrance surface doses. In this study, entrance surface doses was measured using LiF thermoluminescent dosimeters on the skin (either mandibular or maxillary arcs) of 40 patients. Three thermoluminescent dosimetry chips were placed on the skin of each patient. The doses were averaged for each radiography and mean entrance surface doses of all patients calculated. Results: the mean ±SD entrance surface dose at the center of the beam on the patient's skin in intraoral radiography was 1.173 ±0.606 mGy (ranged from 0.01 o 0.40 m Gy). The mean entrance surface doses for male and female patients were 1.380± 0.823, and 1.004± 0.258 respectively. No statistically significant difference was found between these means. Despite its necessity , in national level , there is no published data on the diagnostic reference levels for intraoral radiography. However, the results obtained in this study are lower than those reported by investigators in other countries. Conclusion: in IR Iran , due to lack of large scale studies, no diagnostic reference levels have been set for X-ray diagnostic procedures. Due to lack of national diagnostic reference levels, it is not possible to clarify whether in intraoral radiographies any dose reduction techniques are needed. We intend to perform similar nationwide studies to set the

Diagnostic reference levels in intraoral radiology: From the laboratory to clinical practice

International Nuclear Information System (INIS)

Alcaraz, M.; Velasco, E.; Martinez-Beneyto, Y.; Velasco, F.; Parra, C.; Canteras, M.

2010-01-01

To determine the diagnostic reference levels (DRLs) for obtaining a diagnostic image in the normal conditions of clinical practice and to explain the differences between the levels found and the DRLs obtained in other experimental conditions, suggesting that there has been a reduction in the European Union (EU) recommended levels. A total of 2296 official reports on dental surgeries from 16 Spanish autonomous regions compiled during 2008 were studied. A mean DRL of 3.3 mGy was determined: 2.6 mGy for installations using direct digital systems, 3.4 mGy for those using indirect systems, 4.4 mGy for those using Ultra-speed film and 3.7 mGy for those using Insight. The DRLs found in this survey are below the EU recommended values but far above previously described values, possibly because all the different systems were considered and because values refer to those of the normal work conditions of clinical practice. (authors)

Defining safety culture and the nexus between safety goals and safety culture. 4. Enhancing Safety Culture Through the Establishment of Safety Goals

International Nuclear Information System (INIS)

Tateiwa, Kenji; Miyata, Koichi; Yahagi, Kimitoshi

2001-01-01

Safety culture is the perception of each individual and organization of a nuclear power plant that safety is the first priority, and at Tokyo Electric Power Company (TEPCO), we have been practicing it in everyday activities. On the other hand, with the demand for competitiveness of nuclear power becoming even more intense these days, we need to pursue efficient management while maintaining the safety level at the same time. Below, we discuss how to achieve compatibility between safety culture and efficient management as well as enhance safety culture. Discussion at Tepco: safety culture-nurturing activities such as the following are being implemented: 1. informing the employees of the 'Declaration of Safety Promotion' by handing out brochures and posting it on the intranet home page; 2. publishing safety culture reports covering stories on safety culture of other industry sectors, recent movements on safety culture, etc.; 3. conducting periodic questionnaires to employees to grasp how deeply safety culture is being established; 4. carrying out educational programs to learn from past cases inside and outside the nuclear industry; 5. committing to common ownership of information with the public. The current status of safety culture in Japan sometimes seems to be biased to the quest of ultimate safety; rephrasing it, there have been few discussions regarding the sufficiency of the quantitative safety level in conjunction with the safety culture. Safety culture is one of the most crucial foundations guaranteeing the plant's safety, and for example, the plant safety level evaluated by probabilistic safety assessment (PSA) could be said to be valid only on the ground that a sound and sufficient safety culture exists. Although there is no doubt that the safety culture is a fundamental and important attitude of an individual and organization that keeps safety the first priority, the safety culture in itself should not be considered an obstruction to efforts to implement

On Safety Management. A Frame of Reference for Studies of Safety Management with Examples From Non-Nuclear Contexts of Relevance for Nuclear Safety

Energy Technology Data Exchange (ETDEWEB)

Svensson, Ola; Salo, Ilkka; Allwin, Pernilla (Risk Analysis, Social and Decision Research Unit, Dept. of Psychology, Stockholm Univ., Stockholm (Sweden))

2004-11-15

A good knowledge about safety management from risk technologies outside the area of nuclear power may contribute to both broaden the perspectives on safety management in general, and point at new opportunities for improving safety measures within the nuclear industry. First, a theoretical framework for the study of safety management in general is presented, followed by three case studies on safety management from different non-nuclear areas with potential relevance for nuclear safety. The chapters are written as separate reports and can be read independently of each other. The nuclear industry has a long experience about the management of risky activities, involving all the stages from planing to implementation, both on a more generalized level and in the specific branches of activities (management, administration, operation, maintenance, etc.). Here, safety management is a key concept related to these areas of activities. Outside the field of nuclear power there exist a number of different non-nuclear risk technologies, each one with their own specific needs and experiences about safety management. The differences between the areas consist partly of the different experiences caused by the different technologies. Besides using own experiences in safety practices within the own areas of activities, it may be profitable to take advantage in knowledge and experiences from one area and put it in practice in another area. In order to facilitate knowledge transfer from one technological area to another it may be possible to adapt a common theoretical model, for descriptions and explanations, to the different technologies. Such a model should admit that common denominators for safety management across the areas might be identified and described with common concepts. Systems theory gives the opportunity to not only create models that are descriptive for events within the limits of a given technology, but also to generate knowledge that can be transferred to other

Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

Energy Technology Data Exchange (ETDEWEB)

Almkvist, Lisa (Vattenfall Power Consultant AB, Stockholm (SE)); Gordon, Anna (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE))

2007-11-15

The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60Co and 137Cs in waste packages and on measurements of 239Pu and 240Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

CERN's new safety policy

CERN Multimedia

2014-01-01

The documents below, published on 29 September 2014 on the HSE website, together replace the document SAPOCO 42 as well as Safety Codes A1, A5, A9, A10, which are no longer in force. As from the publication date of these documents any reference made to the document SAPOCO 42 or to Safety Codes A1, A5, A9 and A10 in contractual documents or CERN rules and regulations shall be deemed to constitute a reference to the corresponding provisions of the documents listed below.   "The CERN Safety Policy" "Safety Regulation SR-SO - Responsibilities and organisational structure in matters of Safety at CERN" "General Safety Instruction GSI-SO-1 - Departmental Safety Officer (DSO)" "General Safety Instruction GSI-SO-2 - Territorial Safety Officer (TSO)" "General Safety Instruction GSI-SO-3 - Safety Linkperson (SLP)" "General Safety Instruction GSI-SO-4 - Large Experiment Group Leader In Matters of Safety (LEXGLI...

Multimegawatt Space Reactor Safety

International Nuclear Information System (INIS)

Stanley, M.L.

1989-01-01

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

49 CFR 193.2013 - Incorporation by reference.

Science.gov (United States)

2010-10-01

... MATERIALS SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) PIPELINE SAFETY LIQUEFIED NATURAL... incorporated materials are available for inspection in the Pipeline and Hazardous Materials Safety... referenced material 49 CFR Reference A. American Gas Association (AGA): (1) “Purging Principles and Practices...

Additional safety assessments, follow-up of stress tests of the French nuclear power stations. Action plan of the nuclear safety authority

International Nuclear Information System (INIS)

2012-01-01

This document presents the French national action plan defined by the Nuclear Safety Authority (ASN) in compliance with the recommendations made by the ENSREG (European nuclear safety regulators group). It refers to decisions taken at the national level after the Fukushima accident, recommendations after European stress tests, and recommendations after the extraordinary meeting of contracting parties at the Convention on Nuclear Safety of August 2012. For different topics, this document recalls the recommendations made by the peer review, indicates the ASN's opinion and progress or ASN's requirements for different power stations. The first part addresses the recommendations made by peers and based on the European review. Topics concern natural hazards (effects, detection, inspections and controls related to earthquakes, margin assessment with respect to flooding and natural hazards), the loss of safety systems (cooling systems, electricity supplies, actions, instruments, training, and so on), the management of a severe accident (reference levels, measures, guides, exercises, training, management of contaminated water, radiation protection, premises). The second part deals with topics addressed within the frame of the Convention: national organisations, organisations in an emergency and post-accidental situation, international cooperation. A third part addresses the follow-up of additional measures issued by the ASN

Derivation methods for clearance levels and safety assessments for very low-level radioactive waste disposal

International Nuclear Information System (INIS)

Okoshi, Minoru

2001-01-01

The clearance level was evaluated by the dose of concrete and metal when they would be recycled and reused from shallow land burial of radioactive facilities. The state of waste after clearance is not specified, so that we studied large scale of exposure pathways. The parameter values used for safety assessment were determined as the average values under the consideration of natural and social environment in Japan. Propriety of these values was confirmed by a probability analysis. On the safety assessment of very low-level waste disposal facility, the disposer pathway and parameters were determined under the consideration of special site conditions (natural and social environment) and properties of waste. However, the same exposure pathway of them used the same model for external (exposure by sky shine' s ray) and internal exposure. The calculation results of estimated pathway showed 1.2x10 -5 mSv/y the largest dose for the external exposure pathway by sky shine's ray. (S.Y.)

Experience in the development and practical use of working control levels for radiation safety

International Nuclear Information System (INIS)

Epishin, A.V.

1981-01-01

The experience of development and practical use of working control levels (WCL) of radiation safety in the Gorky region, is discussed. WCL are introduced by ''Radiation Safety Guides'' (RSG-76) and have great practical importance. Regional control levels of radiation safety are determined for certain types of operations implying radioactive hazard and differentiated according to the types of sources applied and types of operation. Dose rates, radioactive contamination of operating surfaces, skin, air and waste water are subject to normalization. Limits of individual radiation doses specified according to operation categories are included. 10 tables of regional WCL indices are developed [ru

High level issues in reliability quantification of safety-critical software

International Nuclear Information System (INIS)

Kim, Man Cheol

2012-01-01

For the purpose of developing a consensus method for the reliability assessment of safety-critical digital instrumentation and control systems in nuclear power plants, several high level issues in reliability assessment of the safety-critical software based on Bayesian belief network modeling and statistical testing are discussed. Related to the Bayesian belief network modeling, the relation between the assessment approach and the sources of evidence, the relation between qualitative evidence and quantitative evidence, how to consider qualitative evidence, and the cause-consequence relation are discussed. Related to the statistical testing, the need of the consideration of context-specific software failure probabilities and the inability to perform a huge number of tests in the real world are discussed. The discussions in this paper are expected to provide a common basis for future discussions on the reliability assessment of safety-critical software. (author)

Application of fundamental aquatic chemistry to the safety case and the role of thermodynamic reference data bases

International Nuclear Information System (INIS)

Altmaier, Marcus; Gaona, Xavier; Fellhauer, David; Geckeis, Horst

2015-01-01

solution formation, must use qualitatively and quantitatively correct radionuclide speciation schemes. The high relevance of correct, complete and consistent thermodynamic data and supplying databases which allow a robust prediction of solution chemistry has been recognized by the international nuclear waste disposal community since decades. Over the last 20 years, the Thermodynamic Database Project of OECD-NEA (http://www.oecd-nea.org/dbtdb) has significantly contributed to the present, largely positive, situation. The NEA-TDB project publishes a series of critically reviewed and evaluated compilations of consistent thermodynamic data, widely accepted as reference values for key elements at low or intermediate ionic strengths conditions. With regard to modeling systems at extremely high ionic strength, activities are currently initiated to (i) prepare a state-of-art-report on Pitzer modeling within NEA-TDB and (ii) set up a working group within the NEA Salt Club to work towards a Joint International Pitzer Database. In Germany, the THEREDA project is developing a German Thermodynamic Reference Database, aiming at providing a comprehensive and internally consistent thermodynamic reference database for the geochemical modeling of all near-field and far-field processes relevant in Germany. In this presentation, the important contributions from fundamental aquatic chemistry in support of the Nuclear Waste Disposal Safety Case are highlighted. Thermodynamic reference databases and their relevance for the Safety Case are analyzed. Based upon a critical assessment of the status quo, positive directions for future research activities and international cooperation are discussed and prioritized.

Application of fundamental aquatic chemistry to the safety case and the role of thermodynamic reference data bases

Energy Technology Data Exchange (ETDEWEB)

Altmaier, Marcus; Gaona, Xavier; Fellhauer, David; Geckeis, Horst [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear Waste Disposal

2015-07-01

solid solution formation, must use qualitatively and quantitatively correct radionuclide speciation schemes. The high relevance of correct, complete and consistent thermodynamic data and supplying databases which allow a robust prediction of solution chemistry has been recognized by the international nuclear waste disposal community since decades. Over the last 20 years, the Thermodynamic Database Project of OECD-NEA (http://www.oecd-nea.org/dbtdb) has significantly contributed to the present, largely positive, situation. The NEA-TDB project publishes a series of critically reviewed and evaluated compilations of consistent thermodynamic data, widely accepted as reference values for key elements at low or intermediate ionic strengths conditions. With regard to modeling systems at extremely high ionic strength, activities are currently initiated to (i) prepare a state-of-art-report on Pitzer modeling within NEA-TDB and (ii) set up a working group within the NEA Salt Club to work towards a Joint International Pitzer Database. In Germany, the THEREDA project is developing a German Thermodynamic Reference Database, aiming at providing a comprehensive and internally consistent thermodynamic reference database for the geochemical modeling of all near-field and far-field processes relevant in Germany. In this presentation, the important contributions from fundamental aquatic chemistry in support of the Nuclear Waste Disposal Safety Case are highlighted. Thermodynamic reference databases and their relevance for the Safety Case are analyzed. Based upon a critical assessment of the status quo, positive directions for future research activities and international cooperation are discussed and prioritized.

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

International Nuclear Information System (INIS)

Lipar, M.

1997-01-01

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

Using level-I PRA for enhanced safety of the advanced neutron source reactor

International Nuclear Information System (INIS)

Ramsey, C.T.; Linn, M.A.

1995-01-01

The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

Swiss National Reference Levels in Fluoroscopy

International Nuclear Information System (INIS)

Aroua, A.; Baechler, S.; Verdun, F.R.; Rickli, H.; Trueb, Ph.R.; Vock, P.

2006-01-01

A nationwide survey was launched in Switzerland in order to investigate the use of fluoroscopy and to establish national reference levels (R.L.) for dose-intensive procedures particularly in interventional radiology. The 2-year investigation covered 5 radiology and 9 cardiology departments in public hospitals and private clinics, and focused on twelve types of examinations: six diagnostic and six interventional. The performance of the fluoroscopy units used in these health-care centres (image quality and dose) was assessed extensively and 1000 examinations were registered. Information on the fluoroscopy time (T), the number of frames (N), the dose-area product (D.A.P.), the difficulty of the case, the age, gender, height and weight of the patient, as well as the experience of the practitioner was provided. The whole set of data was used in relative values (to the mean values for each type of examination) to establish the distributions of T, N and the D.A.P.. From these distributions a set of R.L. values was deduced for the types of examinations investigated using the 3.-quartile method. The R.L. values found are compared to the data published in the literature. (authors)

Sizewell 'B' PWR reference design

International Nuclear Information System (INIS)

1982-04-01

The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

Technology, Safety and Costs of Decommissioning a Reference Uranium Hexafluoride Conversion Plant

Energy Technology Data Exchange (ETDEWEB)

Elder, H. K.

1981-10-01

Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0

Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

2014-01-01

As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia

Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mazleha Maskin; Phongsakorn, P.T.; Tonny, A.L.; Fedrick, C.M.B.; Faizal Mohamed; Mohamad Fauzi Saad; Ahmad Razali Ismail; Mohamad Puad Haji Abu

2013-01-01

Full-text: As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia. (author)

Advice concerning the advantages of a reference incinerator for low-level and intermediate-level radioactive waste processing

International Nuclear Information System (INIS)

Luyten, G.B.

1985-05-01

In this report, an inventory is presented of new incinerators and flue gas filters used in low and intermediate-level radioactive waste combustion. It is argued that a 'reference equipment' for the combustion of solid and liquid low- and intermediate-level wastes best meets existing Dutch radiation protection standards. A cost-benefit analysis of such an equipment is given including annual costs of investment, capital and exploration. A separate combustion process of organic liquids and carrions is considered finally. (G.J.P.)

Trending of low level events and near misses to enhance safety performance in nuclear power plants

International Nuclear Information System (INIS)

2005-11-01

The IAEA Safety Fundamentals publication, Safety of Nuclear Installations, Safety Series No. 110, states the need for operating organizations to establish a programme for the collection and analysis of operating experience in nuclear power plants. Such a programme ensures that operating experience is analysed, events important to safety are reviewed in depth, and lessons learned are disseminated to the staff of the organization and to relevant national and international organizations. As a result of the effort to enhance safety in operating organizations, incidents are progressively decreasing in number and significance. This means that in accordance with international reporting requirements the amount of collected data becomes less sufficient to draw meaningful statistical conclusions. This is where the collection and trend analysis of low level events and near misses can prove to be very useful. These trends can show which of the safety barriers are weak or failing more frequently. Evaluation and trending of low level events and near misses will help to prevent major incidents because latent weaknesses have been identified and corrective actions taken to prevent recurrence. This leads to improved safety and production. Low level events and near misses, which may reach several thousand per reactor operating year, need to be treated by the organizations as learning opportunities. A system for capturing these low level events and near misses truly needs to be an organization-wide system in which all levels of the organization, including contractors, participate. It is desirable that the overall operational experience feedback (OEF) process should integrate the lessons learned and the associated data from significant events with those of lower level events and near misses. To be able to effectively implement a process dealing with low level events and near misses, it is necessary that the organization have a well established OEF process for significant events

Nuclear power and safety

International Nuclear Information System (INIS)

Chidambaram, R.

1992-01-01

Some aspects of safety of nuclear power with special reference to Indian nuclear power programme are discussed. India must develop technology to protect herself from the adverse economic impact arising out of the restrictive regime which is being created through globalization of safety and environmental issues. Though the studies done and experience gained so far have shown that the PHWR system adopted by India has a number of superior safety features, research work is needed in the field of operation and maintenance of reactors and also in the field of reactor life extension through delaying of ageing effects. Public relations work must be pursued to convince the public at large of the safety of nuclear power programme. The new reactor designs in the second stage of evolution are based on either further improvement of existing well-proven designs or adoptions of more innovative ideas based on physical principles to ensure a higher level of safety. The development of Indian nuclear power programme is characterised by a balanced approach in the matter of assuring safety. Safety enforcement is not just looked upon as a pure administrative matter, but experts with independent minds are also involved in safety related matters. (M.G.B.)

Safety and protection aspects of the management of high-level radioactive wastes

International Nuclear Information System (INIS)

Candes, P.; Pradel, J.

1977-01-01

Appropriate consideration is given in France to safety and protection problems to be solved from production up to the final disposal of high-level radioactive wastes. The first stage of this work consisted in emphasizing the various technical options. Different strategies appear to be possible, taking into account the technical, political and psychological difficulties. This results in evaluating the safety problems to be solved in the framework of those strategies. In this field the main safety and protection principles do not differ from those applying to other nuclear facilities. Nevertheless, the factor of time is in most cases quite different (thousands or millions of years). The question is then raised of evaluating the importance to be given to very remote consequences, both at philosophical and at scientific levels. As a first result of these considerations, the application of the barrier concept is recommended. This concept is familiar to safety specialists. Different barriers, for which particular problems are listed and evaluated, are defined. Another result with regard to radiation protection principles is to consider that if safety provisions should lead to as efficient a containment of radioactive products as possible, it would not be realistic to consider such a containment as absolute, in particular for disposal lasting anything up to thousands of years. It is therefore assumed that a limited radioactivity transfer should be taken into account, and its consequences for the environment and man calculated. This is especially true in the study of an appropriate site for final storage, and the study should necessarily include a detailed investigation of the retention characteristics of soil layers, and the implementation of appropriate models giving a sufficiently accurate evaluation of the consequences of transfers, including those related to the effect of various elements after their arrival into the biosphere. The authors review the different

Criticality Safety Evaluation for Small Sample Preparation and Non-Destructive Assay (NDA) Operations in Wing 7 Basement of the CMR Facility

Energy Technology Data Exchange (ETDEWEB)

Kunkle, Paige Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zhang, Ning [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-02

Nuclear Criticality Safety (NCS) has reviewed the fissionable material small sample preparation and NDA operations in Wing 7 Basement of the CMR Facility. This is a Level-1 evaluation conducted in accordance with NCS-AP-004 [Reference 1], formerly NCS-GUIDE-01, and the guidance set forth on use of the Standard Criticality Safety Requirements (SCSRs) [Reference 2]. As stated in Reference 2, the criticality safety evaluation consists of both the SCSR CSED and the SCSR Application CSED. The SCSR CSED is a Level-3 CSED [Reference 3]. This Level-1 CSED is the SCSR Application CSED. This SCSR Application (Level-1) evaluation does not derive controls, it simply applies controls derived from the SCSR CSED (Level-3) for the application of operations conducted here. The controls derived in the SCSR CSED (Level-3) were evaluated via the process described in Section 6.6.5 of SD-130 (also reproduced in Section 4.3.5 of NCS-AP-004 [Reference 1]) and were determined to not meet the requirements for consideration of elevation into the safety basis documentation for CMR. According to the guidance set forth on use of the SCSRs [Reference 2], the SCSR CSED (Level-3) is also applicable to the CMR Facility because the process and the normal and credible abnormal conditions in question are bounded by those that are described in the SCSR CSED. The controls derived in the SCSR CSED include allowances for solid materials and solution operations. Based on the operations conducted at this location, there are less-than-accountable (LTA) amounts of ²³³U. Based on the evaluation documented herein, the normal and credible abnormal conditions that might arise during the execution of this process will remain subcritical with the following recommended controls.

Supporting Multiprocessors in the Icecap Safety-Critical Java Run-Time Environment

DEFF Research Database (Denmark)

Zhao, Shuai; Wellings, Andy; Korsholm, Stephan Erbs

The current version of the Safety Critical Java (SCJ) specification defines three compliance levels. Level 0 targets single processor programs while Level 1 and 2 can support multiprocessor platforms. Level 1 programs must be fully partitioned but Level 2 programs can also be more globally...... scheduled. As of yet, there is no official Reference Implementation for SCJ. However, the icecap project has produced a Safety-Critical Java Run-time Environment based on the Hardware-near Virtual Machine (HVM). This supports SCJ at all compliance levels and provides an implementation of the safety......-critical Java (javax.safetycritical) package. This is still work-in-progress and lacks certain key features. Among these is the ability to support multiprocessor platforms. In this paper, we explore two possible options to adding multiprocessor support to this environment: the “green thread” and the “native...

Dose evaluation and establishment of reference levels in activity for nuclear medicine

International Nuclear Information System (INIS)

Ribeiro, Julio Cesar de Souza

2017-01-01

The International Commission on Radiation Protection (ICRP) has emphasized the importance of accurately determining the mean dose levels, or administered activity, received by the patients for each medical procedure that uses ionizing radiation. However, the number of bibliographic references addressing the need to know and optimize these levels is insufficient, or rather limited, which may lead to non-standardizes techniques, a lack of exposures control, and also the increase of associated radiological risks of these procedures. In this context, a software in Visual Basic® of Microsoft© language was developed whose function is to elaborate a method of obtaining the Reference Levels in Activity (RLA) for nuclear medicine patients by determining the third quartile of the examinations carried out. The program also allows obtaining absorbed dose values in critical organs based on patient specificities as age, sex and Body Mass Index (BMI) in order to evaluate the risk involved in each procedure. The main nuclear medicine diagnostic procedures were evaluated through the database of two public hospitals and a private clinic, obtaining the NRAs of each facility, where the software was validated by comparison with the traditionally accepted calculation methods. Due to the results obtained in each installation, in addition to NRA determination, gaps in treatment capacities and unjustified dose variations for the same procedure were identified, indicating the need for optimization. Thus, the developed program is able to provide the estimated values of effective and absorbed doses involved in each procedure, for each patient, providing reference values for nuclear medicine field, not available in the national scenario so far. (author)

Patient safety in maternal healthcare at secondary and tertiary level facilities in Delhi, India

Directory of Open Access Journals (Sweden)

Chandrakant Lahariya

2015-01-01

Full Text Available Background: There is insufficient information on causes of unsafe care at facility levels in India. This study was conducted to understand the challenges in government hospitals in ensuring patient safety and to propose solutions to improve patient care. Materials and Methods: Desk review, in-depth interviews, and focused group discussions were conducted between January and March 2014. Healthcare providers and nodal persons for patient safety in Gynecology and Obstetrics Departments of government health facilities from Delhi state of India were included. Data were analyzed using qualitative research methods and presented adopting the "health system approach." Results: The patient safety was a major concern among healthcare providers. The key challenges identified were scarcity of resources, overcrowding at health facilities, poor communications, patient handovers, delay in referrals, and the limited continuity of care. Systematic attention on the training of care providers involved in service delivery, prescription audits, peer reviews, facility level capacity building plan, additional financial resources, leadership by institutional heads and policy makers were suggested as possible solutions. Conclusions: There is increasing awareness and understanding about challenges in patient safety. The available local information could be used for selection, designing, and implementation of measures to improve patient safety at facility levels. A systematic and sustained approach with attention on all functions of health systems could be beneficial. Patient safety could be used as an entry point to improve the quality of health care services in India.

Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

1997-01-01

In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

Main factors determining the KNP units 5 and 6 safety level according to the PSA level 1 result

International Nuclear Information System (INIS)

Manchev, B.; Marinova, B.; Nenkova, B.

2004-01-01

The Probabilistic Safety Analysis (PSA) is a powerful tool for ascertainment of the safety level reached at nuclear power plants operation. The results of PSA determine very clearly the functions, systems, equipment or operator actions that have to be improved in order to increase the plant safety level as a whole. The present report presents the main results of the last upgraded revision of PSA level 1 of units 5 and 6 of KNPP. The objective of the report is to lay emphasis on the factors determining the result obtained, i.e. to demonstrate the scopes whose improvement leads to an increase of the safety level reached at the units power operation. In the frame of the study presented the following categories of initiating events are included: Internal initiating events; Initiating events result of internal fires; Initiating events result of seismic action; Floods. Only the reactor core is considered as a source of radioactive contamination. Only initiating events related to the reactor work on power are analyzed. Unit 5 of KNPP is accepted as a basic unit for the study. All modifications and design changes implemented up to year 2000 are taken into account. The results of PSA level 1 for units 5 and 6 of KNPP covering the risk of internal initiators are presented. The assessment of the core damage due to internal initiators is based on the analysis of 18 groups of initiating events. 932 consequences and two groups of initial events are identified, leading to core damage. As a result of the quantitative calculation, over 15000 minimal cuts for the core damage are obtained. The first 80 cuts bear over 75% of the frequency obtained, and the first 700 cuts bear over 90%. Distribution of the core damage frequency by different groups of initiators is presented in tables and diagrams. A comparison of the result obtained for the reactor core damage of KNPP units 5 and 6 with assessment obtained for similar power plants is presented. The data for different NPPs are taken

A Methodology for Safety Culture Impact Assessment

Energy Technology Data Exchange (ETDEWEB)

Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of)

2014-05-15

The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively.

A Methodology for Safety Culture Impact Assessment

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2014-01-01

The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively

Patient radiation doses and reference levels in pediatric interventional radiology

Energy Technology Data Exchange (ETDEWEB)

Habib Geryes, Bouchra; Lachaux, Julie; Boddaert, Nathalie; Brunelle, Francis [Hopital Universitaire Necker Enfants Malades, Department of Paediatric Radiology, Paris (France); Bak, Adeline; Ozanne, Augustin; Saliou, Guillaume [Hopital Bicetre, Hopitaux Universitaires Paris-Sud, Department of Neuroradiology, Le Kremlin Bicetre (France); Naggara, Olivier [Hopital Universitaire Necker Enfants Malades, Department of Paediatric Radiology, Paris (France); Centre Hospitalier Sainte-Anne, Universite Paris Descartes Sorbonne Paris Cite, Centre de Psychiatrie et Neurosciences, INSERM S894, DHU Neurovasculaire, Paris (France); Centre Hospitalier Sainte-Anne, Department of Neuroradiology, Universite Paris Descartes, Sorbonne Paris Cite, INSERM UMR894, Paris (France)

2017-09-15

To describe, in a multicentric paediatric population, reference levels (RLs) for three interventional radiological procedures. From January 2012 to March 2015, children scheduled for an interventional radiological procedure in two French tertiary centres were retrospectively included and divided into four groups according to age: children younger than 2 years (A1), aged 2-7 years (A5), 8-12 years (A10) and 13-18 years (A15). Three procedures were identified: cerebral digital subtraction angiography (DSA), brain arteriovenous malformation (bAVM) embolization, and head and neck superficial vascular malformation (SVM) percutaneous sclerotherapy. Demographic and dosimetric data, including dose area product (DAP), were collected. 550 procedures were included. For DSA (162 procedures), the proposed RL values in DAP were 4, 18, 12 and 32 Gy.cm{sup 2} in groups A1, A5, A10 and A15, respectively. For bAVM embolization (258 procedures), values were 33, 70, 105 and 88 Gy.cm{sup 2} in groups A1, A5, A10 and A15, respectively. For SVM sclerotherapy (130 procedures), values were 350, 790, 490 and 248 mGy.cm{sup 2} in groups A1, A5, A10 and A15, respectively. Consecutive data were available to permit a proposal of reference levels for three major paediatric interventional radiology procedures. (orig.)

Statistical re-evaluation of the ASME K{sub IC} and K{sub IR} fracture toughness reference curves

Energy Technology Data Exchange (ETDEWEB)

Wallin, K.; Rintamaa, R. [Valtion Teknillinen Tutkimuskeskus, Espoo (Finland)

1998-11-01

Historically the ASME reference curves have been treated as representing absolute deterministic lower bound curves of fracture toughness. In reality, this is not the case. They represent only deterministic lower bound curves to a specific set of data, which represent a certain probability range. A recently developed statistical lower bound estimation method called the `Master curve`, has been proposed as a candidate for a new lower bound reference curve concept. From a regulatory point of view, the Master curve is somewhat problematic in that it does not claim to be an absolute deterministic lower bound, but corresponds to a specific theoretical failure probability that can be chosen freely based on application. In order to be able to substitute the old ASME reference curves with lower bound curves based on the master curve concept, the inherent statistical nature (and confidence level) of the ASME reference curves must be revealed. In order to estimate the true inherent level of safety, represented by the reference curves, the original data base was re-evaluated with statistical methods and compared to an analysis based on the master curve concept. The analysis reveals that the 5% lower bound Master curve has the same inherent degree of safety as originally intended for the K{sub IC}-reference curve. Similarly, the 1% lower bound Master curve corresponds to the K{sub IR}-reference curve. (orig.)

Safety culture in industrial radiography facility

International Nuclear Information System (INIS)

Vincent-Furo, Evelyn

2015-02-01

This project reviewed published IAEA materials and other documents on safety culture with specific references to industrial radiography. Safety culture requires all duties important to safety to be carried out correctly, with alertness, due thought and full knowledge, sound judgment and a proper sense of accountability. The development and maintenance of safety culture in an operating organization has to cover management systems, policies, responsibilities, procedures and organizational arrangements. The essence is to control radiation hazard, optimize radiation protection to prevent or reduce exposures and mitigate the consequences of accidents and incidents. To achieve a high degree of safety culture appropriate national and international infrastructure should exist to ensure effective training of workers and management system that supports commitment to safety culture at all level of the organization; management, managers and workforce. The result of the review revealed that all accidents in industrial radiography facilities were due to poor safety culture practices including inadequate regulatory control oversight. Some recommendations are provided and if implemented could improve safety culture leading to good safety performance which will significantly reduce accidents and their consequences in industrial radiography. These examples call for a review of safety culture in Industrial radiography. (au)

Nuclear Safety. 1997

International Nuclear Information System (INIS)

1998-01-01

A quick review of the nuclear safety at EDF may be summarized as follows: - the nuclear safety at EDF maintains at a rather good standard; - none of the incidents that took place has had any direct impact upon safety; - the availability remained good; - initiation of the floor 4 reactor generation (N4 unit - 1450 MW) ensued without major difficulties (the Civaux 1 NPP has been coupled to the power network at 24 december 1997); - the analysis of the incidents interesting from the safety point of view presents many similarities with earlier ones. Significant progress has been recorded in promoting actively and directly a safe operation by making visible, evident and concrete the exertion of the nuclear operation responsibility and its control by the hierarchy. The report develops the following chapters and subjects: 1. An overview on 1997; 1.1. The technical issues of the nuclear sector; 1.2. General performances in safety; 1.3. The main incidents; 1.4. Wastes and radiation protection; 2. Nuclear safety management; 2.1. Dynamics and results; 2.2. Ameliorations to be consolidated; 3. Other important issues in safety; 3.1. Probabilistic safety studies; 3.2. Approach for safety re-evaluation; 3.3. The network safety; 3.4. Crisis management; 3.5. The Lifetime program; 3.6. PWR; 3.7. Documentation; 3.8. Competence; 4. Safety management in the future; 4.1. An open future; 4.2. The fast neutron NPP at Creys-Malville; 4.3. Stabilization of the PWR reference frame; 4.4. Implementing the EURATOM directive regarding the radiation protection standards; 4.5. Development of biomedical research and epidemiological studies; 4.6. New regulations concerning the liquid and gaseous effluents; 5. Visions of an open future; 5.1. Alternative views upon safety ay EDF; 5.2. Safety authority; 5.3. International considerations; 5.4. What happens abroad; 5.5. References from non-nuclear domain. Four appendices are added referring to policy of safety management, policy of human factors in NPPs

Metrics design for safety assessment

NARCIS (Netherlands)

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

2016-01-01

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

Cost update: Technology, safety, and costs of decommissioning a reference uranium fuel fabrication plant

International Nuclear Information System (INIS)

Miles, T.L.; Liu, Y.

1994-06-01

The cost estimates originally developed in NUREG/CR-1266 for commissioning a reference low-enrichment uranium fuel fabrication plant are updated from 1978 to early 1993 dollars. During this time, the costs for labor and materials increased approximately at the rate of inflation, the cost of energy increased more slowly than the rate of inflation, and the cost of low-level radioactive waste disposal increased much more rapidly than the rate of inflation. The results of the analysis indicate that the estimated costs for the immediate dismantlement and decontamination for unrestricted facility release (DECON) of the reference plant have increased from the mid-1978 value of $3.57 million to $8.08 million in 1993 with in-compact low-level radioactive waste disposal at the US Ecoloay facility near Richland, Washington. The cost estimate rises to $19.62 million with out-of-compact radioactive waste disposal at the Chem-Nuclear facility near Barnwell, South Carolina. A methodology and a formula are presented for estimating the cost of decommissioning the reference uranium fuel fabrication plant at some future time, based on these early 1993 cost estimates. The formula contains essentially the same elements as the formula given in 10 CFR 50.75 for escalating the decommissioning costs for nuclear power reactors to some future time

The Radiation Safety Culture: Image Gently

International Nuclear Information System (INIS)

Applegate, E.K.

2015-01-01

Barriers to Implementing Safety include Silos of Knowledge, Time, training and Resources. Creating a Safety Culture in Healthcare include Decreased authority gradients, Checklists and audits (QA), Use of structured language (SBAR), Situation, Background, Assessment, Recommendation Team briefings and debriefings (immediate learning, team building tools), Lifelong learning (PQI). Use of Collective Learning Opportunities - QA and PQI that include Web sites: IG, WFPI, IAEA, ISR and Data Registries: ACR . The Key Principles of Radiation Protection: When do we learn them? For Occupational Workers:Time, Distance and Shielding while those of For Patients: Justification, Optimization and Dose Limits (dose reference levels)

Local diagnostic reference levels at the Portuguese Institute of oncology francisco gentil of coimbra

International Nuclear Information System (INIS)

Bras, S.; Sousa, M.C. de; Lopes, M.C.

2006-01-01

In the scope of medical radiological exposures and according to European recommendations and national legal requirements [14], the maximum responsible for an installation must assure the establishment of local dose levels for each type of radiological examination and also assure that they are available for the doctor who prescribes the examination. In the absence of national reference dose levels, the so called Local Diagnostic Reference Levels (L.D.R.L.) should be in agreement with the European Diagnostic Reference Levels published for the different types of medical exposures. The aim of this work was to establish a protocol of measurement for each type of more frequent examination, namely in conventional radiology, in CT and in mammography performed in our hospital. For each kind of examination the recommended dose descriptor was adopted and directly measured or derived from basic charge measurements. The patient sample corresponded in each case to at least a minimum of 10 standard-sized patients, as recommended, in order to obtain averages that constitute for each type of radiological examination the L.D.R.L. that could be compared with the European D.R.L.. The results obtained are in the large majority of the situations below the corresponding D.R.L.. Nevertheless we have identified some situations that deserve more attention.In conventional radiology all L.D.R.L. are below the reference levels. However we have detected some skull post anterior (skull P.A.) exposures where entrance skin doses exceeded the standard value. This may be due to equipment age problems but, as always, improvement of staff education will contribute to better practices and we hope that this work can contribute to this objective. In CT the L.D.R.L. that corresponds to single slice scan meet the standards whereas complete examinations described by dose length product (D.L.P.) values show a larger variation and also some situations where the reference level is exceeded. It is possible that

Local diagnostic reference levels at the Portuguese Institute of oncology francisco gentil of coimbra

Energy Technology Data Exchange (ETDEWEB)

Bras, S.; Sousa, M.C. de [Algarve Univ., LIP-Algarve, FCT (Portugal); Lopes, M.C. [IPOFG-CROC, S.A., Medical Physics Dept., Coimbra (Portugal)

2006-07-01

In the scope of medical radiological exposures and according to European recommendations and national legal requirements [14], the maximum responsible for an installation must assure the establishment of local dose levels for each type of radiological examination and also assure that they are available for the doctor who prescribes the examination. In the absence of national reference dose levels, the so called Local Diagnostic Reference Levels (L.D.R.L.) should be in agreement with the European Diagnostic Reference Levels published for the different types of medical exposures. The aim of this work was to establish a protocol of measurement for each type of more frequent examination, namely in conventional radiology, in CT and in mammography performed in our hospital. For each kind of examination the recommended dose descriptor was adopted and directly measured or derived from basic charge measurements. The patient sample corresponded in each case to at least a minimum of 10 standard-sized patients, as recommended, in order to obtain averages that constitute for each type of radiological examination the L.D.R.L. that could be compared with the European D.R.L.. The results obtained are in the large majority of the situations below the corresponding D.R.L.. Nevertheless we have identified some situations that deserve more attention.In conventional radiology all L.D.R.L. are below the reference levels. However we have detected some skull post anterior (skull P.A.) exposures where entrance skin doses exceeded the standard value. This may be due to equipment age problems but, as always, improvement of staff education will contribute to better practices and we hope that this work can contribute to this objective. In CT the L.D.R.L. that corresponds to single slice scan meet the standards whereas complete examinations described by dose length product (D.L.P.) values show a larger variation and also some situations where the reference level is exceeded. It is possible that

Preliminary Diagnostic Reference Levels of Adult CT at Aristide Ledantec National Hospital

International Nuclear Information System (INIS)

Diagne, M.; Gning, F.; Dieng, M. M.; Gueye, L.

2015-01-01

The number of Computed Tomography (CT) procedures performed in Senegal has widely increased as the CT is a powerful tool for the accurate and effective diagnosis. CT is a diagnostic imaging modality giving higher patient dose in comparison with other radiological procedures. The establishing of diagnostic reference levels (DRLs) is a way to optimize the radiation arising from CT procedures to as low as reasonable (ALARA) and to ensure good practice. Objective: The purpose of this study is to establish Local Diagnostic Reference Levels (LDRLs) at the University Hospital of Aristide LeDantec for CT examinations and to compare these values with the international Diagnostic Reference Levels (DRLs) to benchmark the local practice. Materials/Methods: This was a cross-sectional survey carried out in HALD between August 2014 and January 2015. Demographic data and acquisition parameters of 700 CT scan examinations performed on adult patients were collected from request forms and CT scan consoles. The values of CTDIw, CTDIvol and DLP were calculated using ImPACT (Imaging Performance and Assessment of Computed Tomography) software for Siemens Definition AS scanner of HALD. This was done by correlating the measurements from the National Radiological Protection Board (NRPB–R250) scanners with the effective dose calculated, using the CT–EXPO software. Data was analyzed using mean, range, 3rd quartile, as well as mean. Frequency tables and histogrammes were used to summarise the data. Results: The 3rd quartile doses in this study for head, chest, abdomen and pelvis were 89 mGy, 12 mGy, 16.5 mGy, and 15 mGy, respectively. These values were in good agreement with the values reported from the literature. (author)

Optimal safety levels via social indicators

International Nuclear Information System (INIS)

Lind, N.C.; Nathwani, J.S.

1992-01-01

In the management of natural or technological hazards in a society, the objective should be to serve the public interest in a rational manner. Decisions with regard to risk levels for the public - if they are to be defensible and self-consistent - require an integrated system of values that covers the entire range of hazards under public regulation. The process for setting risk levels (or safety goals) should ideally involve a thorough consideration of cost and benefit of all kinds, supported by explicit quantified comparison on a widely acceptable scale. The purpose of the paper is to show how quantitative criteria within the context of an appropriate framework can be used to guide risk management decisions. Social indicators are time series, statistics that reflect some aspect of the quality of life in a society or group of individuals. Development, validation, and use of social indicators is an important current research activity, as exemplified by journals such as Social Indicators Research. The basic objective is to provide quantitative measures for assessing the rationales and effectiveness of public decision-making. The concept is applicable to the nuclear industry

Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Timothy Solack; Carol Mason

2012-03-01

A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

Annual plan of research on safety techniques against low level radioactive wastes, FY1994-FY1999

International Nuclear Information System (INIS)

1994-01-01

The safety research on the disposal of low level radioactive waste has been promoted based on the annual plan decided by the committee on radiative waste safety regulation of the Nuclear Safety Commission. Hereafter, the disposal of low level radioactive waste in ocean is never selected. As to the subjects of the safety research which should be carried out for five years from 1994, the necessity, the contents of research, the organs that carry out the research and so on were deliberated, and the results are made into the annual plan, therefore, it is reported. The way of thinking on the safety research, the contents for which efforts should be exerted as the safety research, and the matters to which attention should be paid are shown. As for the annual plan of safety research, the necessity and the outline of the safety research on the disposal in strata, the concrete subjects and their contents, and the necessity and the outline of the safety research on the reuse, the concrete subjects and their contents are reported. The radioactive waste is those produced by the operation of nuclear reactor facilities, those containing TRU nuclides and RI waste. (K.I.)

Transformational and passive leadership as cross-level moderators of the relationships between safety knowledge, safety motivation, and safety participation.

Science.gov (United States)

Jiang, Lixin; Probst, Tahira M

2016-06-01

While safety knowledge and safety motivation are well-established predictors of safety participation, less is known about the impact of leadership styles on these relationships. The purpose of the current study was to examine whether the positive relationships between safety knowledge and motivation and safety participation are contingent on transformational and passive forms of safety leadership. Using multilevel modeling with a sample of 171 employees nested in 40 workgroups, we found that transformational safety leadership strengthened the safety knowledge-participation relationship, whereas passive leadership weakened the safety motivation-participation relationship. Under low transformational leadership, safety motivation was not related to safety participation; under high passive leadership, safety knowledge was not related to safety participation. These results are discussed in light of organizational efforts to increase safety-related citizenship behaviors. Copyright © 2016 Elsevier Ltd and National Safety Council. All rights reserved.

Exhaled nitric oxide levels in asthma: Personal best versus reference values.

Science.gov (United States)

Smith, Andrew D; Cowan, Jan O; Taylor, D Robin

2009-10-01

Factors affecting the fraction of nitric oxide in exhaled air (FE(NO)) are multiple. Interpreting values when assessing airways disease may be problematic. Clinically optimum levels have not been defined. We aimed to establish the relationship between predicted values for FE(NO) obtained from equations by Olin et al, Travers et al, and Dressel et al, and normalized levels after oral prednisone. We also compared postprednisone FE(NO) levels with those obtained during optimized treatment with inhaled fluticasone. Data were obtained before and after a trial of oral prednisone (30mg/d for 14 days), and also from a previously published study in which patients had their dose of inhaled corticosteroid adjusted using either FE(NO) or symptoms/lung function to optimize treatment. Seventy-three patients completed the study. The geometric mean FE(NO) after prednisone (17.7 parts per billion [ppb]; 95% CI, 15.5-20.2) was significantly lower than mean FE(NO) at the optimized fluticasone dose (20.2 ppb; 95% CI, 17.1-23.8; P=.04) and at loss of control (27.6 ppb; 95% CI, 22.8-33.4; P values of Olin et al (16.8 ppb, 95% CI, 16.0-17.5; P=.44), but were significantly lower than values of Travers et al (predicted, 21.5 ppb; 95% CI, 20.9-22.2; P=.005) and Dressel et al (predicted, 27.8 ppb; 95% CI, 26.7-28.9; P values from the reference equation by Olin et al. However, at optimized doses of inhaled corticosteroid, although FE(NO) levels were higher than predicted, asthma was well controlled. Targeting FE(NO) on reference values is not justified.

Areva - 2016 Reference document

International Nuclear Information System (INIS)

2017-01-01

Areva supplies high added-value products and services to support the operation of the global nuclear fleet. The company is present throughout the entire nuclear cycle, from uranium mining to used fuel recycling, including nuclear reactor design and operating services. Areva is recognized by utilities around the world for its expertise, its skills in cutting-edge technologies and its dedication to the highest level of safety. Areva's 36,000 employees are helping build tomorrow's energy model: supplying ever safer, cleaner and more economical energy to the greatest number of people. This Reference Document contains information on Areva's objectives, prospects and development strategies. It contains estimates of the markets, market shares and competitive position of Areva

Management of health and safety in the organization of worktime at the local level.

Science.gov (United States)

Jeppesen, H J; Bøggild, H

1998-01-01

This study examined the consideration of health and safety issues in the local process of organizing worktime within the framework of regulations. The study encompassed all 7 hospitals in one region of Denmark. Twenty-three semi-structured interviews were carried out with 2 representatives from the different parties involved (management, cooperation committees, health and safety committees from each hospital, and 2 local unions). Furthermore, a questionnaire was sent to all 114 wards with day and night duty. The response rate was 84%. Data were collected on alterations in worktime schedules, responsibilities, reasons for the present design of schedules, and use of inspection reports. The organization of worktime takes place in single wards without external interference and without guidelines other than the minimum standards set in regulations. At the ward level, management and employees were united in a mutual desire for flexibility, despite the fact that regulations were not always followed. No interaction was found in the management of health and safety factors between the parties concerned at different levels. The demands for flexibility in combination with the absence of guidelines and the missing dynamics between the parties involved imply that the handling of health and safety issues in the organization of worktime may be accidental and unsystematic. In order to consider the health and safety of night and shift workers within the framework of regulations, a clarification of responsibilities, operational levels, and cooperation is required between the parties concerned.

Technology, safety, and costs of decommissioning a reference nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Schneider, K.J.; Jenkins, C.E.; Rhoads, R.E.

1977-09-01

Safety and cost information were developed for the conceptual decommissioning of a fuel reprocessing plant with characteristics similar to the Barnwell Nuclear Fuel Plant. The main process building, spent fuel receiving and storage station, liquid radioactive waste storage tank system, and a conceptual high-level waste-solidification facility were postulated to be decommissioned. The plant was conceptually decommissioned to three decommissioning states or modes; layaway, protective storage, and dismantlement. Assuming favorable work performance, the elapsed time required to perform the decommissioning work in each mode following plant shutdown was estimated to be 2.4 years for layaway, 2.7 years for protective storage, and 5.2 years for dismantlement. In addition to these times, approximately 2 years of planning and preparation are required before plant shutdown. Costs, in constant 1975 dollars, for decommissioning were estimated to be $18 million for layaway, $19 million for protective storage and $58 million for dismantlement. Maintenance and surveillance costs were estimated to be $680,000 per year after layaway and $140,000 per year after protective storage. The combination mode of protective storage followed by dismantlement deferred for 10, 30, and 100 years was estimated to cost $64 million, $67 million and $77 million, respectively, in nondiscounted total 1975 dollars. Present values of these costs give reduced costs as dismantlement is deferred. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year radiation dose commitment to the members of the public from airborne releases from normal decommissioning activities were estimated to be less than 11 man-rem

Technology, safety, and costs of decommissioning a reference nuclear fuel reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Schneider, K.J.; Jenkins, C.E.; Rhoads, R.E.

1977-09-01

Safety and cost information were developed for the conceptual decommissioning of a fuel reprocessing plant with characteristics similar to the Barnwell Nuclear Fuel Plant. The main process building, spent fuel receiving and storage station, liquid radioactive waste storage tank system, and a conceptual high-level waste-solidification facility were postulated to be decommissioned. The plant was conceptually decommissioned to three decommissioning states or modes; layaway, protective storage, and dismantlement. Assuming favorable work performance, the elapsed time required to perform the decommissioning work in each mode following plant shutdown was estimated to be 2.4 years for layaway, 2.7 years for protective storage, and 5.2 years for dismantlement. In addition to these times, approximately 2 years of planning and preparation are required before plant shutdown. Costs, in constant 1975 dollars, for decommissioning were estimated to be $18 million for layaway, $19 million for protective storage and $58 million for dismantlement. Maintenance and surveillance costs were estimated to be $680,000 per year after layaway and $140,000 per year after protective storage. The combination mode of protective storage followed by dismantlement deferred for 10, 30, and 100 years was estimated to cost $64 million, $67 million and $77 million, respectively, in nondiscounted total 1975 dollars. Present values of these costs give reduced costs as dismantlement is deferred. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year radiation dose commitment to the members of the public from airborne releases from normal decommissioning activities were estimated to be less than 11 man-rem.

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)

Nuclear criticality safety guide

International Nuclear Information System (INIS)

Pruvost, N.L.; Paxton, H.C.

1996-09-01

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators

Nuclear criticality safety guide

Energy Technology Data Exchange (ETDEWEB)

Pruvost, N.L.; Paxton, H.C. [eds.

1996-09-01

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

Main technical options of the Jules Horowitz reactor project to achieve high flux performances and high safety level

International Nuclear Information System (INIS)

Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

2003-01-01

Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 kW/I for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: - A safety approach based upon the defence-in-depth principle. - A strategy of generic approaches to assess experimental risks in the facility. - Internal events analysis taking into account risks linked to reactor and experiments (e.g., radioactive source-term). - Systematic consideration of external hazards (e.g., earthquake, airplane crash) and internal hazards. - Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (authors)

Main technical options of the Jules Horowitz Reactor project to achieve high flux performances and high safety level

International Nuclear Information System (INIS)

Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

2003-01-01

Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 KW/l for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: 1) A safety approach based upon the defence-in-depth principle. 2) A strategy of generic approaches to assess experimental risks in the facility. 3) Internal events analysis taking into account risks linked to reactor and experiments (eg., radioactive source-term). 4) Systematic consideration of external hazards (eg., earthquake, airplane crash) and internal hazards. 5) Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (author)

Age-specific reference values for serum FSH and estradiol levels throughout the reproductive period.

Science.gov (United States)

Grisendi, Valentina; Spada, Elena; Argento, Cindy; Plebani, Maddalena; Milani, Silvano; Seracchioli, Renato; Volpe, Annibale; La Marca, Antonio

2014-06-01

High serum day 3 FSH levels are associated with poor ovarian reserve and reduced fertility, but the interpretation of FSH values according to age is still not univocal. The purpose of this study was to determine age-dependent reference values in women with regular menstrual cycles and FSH as a guide for specialists. The study was performed at the Department of Mother-Infant of a University-based tertiary care centre. One-hundred ninety-two healthy normal menstruating women were recruited for the study. All patients attended the department on menstrual cycle day 3 for a blood sample for FSH and estradiol determination. A linear relationship between FSH or estradiol serum levels and age was observed. The FSH level increased by 0.11 IU for every year of age (1 IU for every 9 years of age). The values of FSH and estradiol corresponding to the 5th, 25th, 50th, 75th, 95th centiles for any specific age have been calculated. Serum FSH levels need to be interpreted according to age-dependent reference values. Serum FSH levels on 95th centile for any age may represent a warning sign for reduced ovarian reserve.

On detecting reference level of acrolein content in children's blood

Directory of Open Access Journals (Sweden)

T.S. Ulanova

2017-03-01

Full Text Available The article gives the results of complex chemical-analytical and clinical-laboratory research in course of which biological media of children living in Perm region were examined. To study impacts exerted by exogenous acrolein we examined 156 children in 2014–2016, aged 5–10, attending pre-school facilities and schools, and living in Perm region. As we conducted this research we detected average annual acrolein concentration in atmosphere on the examined territory; this concentration was equal to 0.000024 mg/m3, and it was 1.2 times higher than reference acrolein concentration in the air for chronic inhalation exposure. Average group acrolein concentration in children's blood was 1.2 times authentically higher (р3.96, p≤0.05. We used increased content of delta-aminolevulinic acid in urine as a limiting marker for effects occurring at chronic inhalation exposure to acrolein. Basing on the results of the performed examination we recommend concentration equal to 0.10 mgr/dm 3 as a reference level of acrolein content in blood at chronic inhalation exposure.

PSA methodology including new design, operational and safety factors, 'Level of recognition of phenomena with a presumed dominant influence upon operational safety' (failures of conventional as well as non-conventional passive components, dependent failures, influence of operator, fires and external threats, digital control, organizational factors)

International Nuclear Information System (INIS)

Jirsa, P.

2001-10-01

The document represents a specific type of discussion of existing methodologies for the creation and application of probabilistic safety assessment (PSA) in light of the EUR document summarizing requirements placed by Western European NPP operators on the future design of nuclear power plants. A partial goal of this discussion consists in mapping, from the PSA point of view, those selected design, operational and/or safety factors of future NPPs that may be entirely new or, at least, newly addressed. Therefore, the terms of reference for this stage were formulated as follows: Assess current level of knowledge and procedures in the analysis of factors and phenomena with a dominant influence upon operational safety of new generation reactors, especially in the following areas: (1) Phenomenology of failure types and mechanisms and reliability of conventional passive safety system components; (2) Phenomenology of failure types and mechanisms and reliability of non-conventional passive components of newly designed safety systems; (3) Phenomenology of types and mechanisms of dependent failures; (4) Human factor role in new generation reactors and its effect upon safety; (5) Fire safety and other external threats to new nuclear installations; (6) Reliability of the digital systems of the I and C system and their effect upon safety; and (7) Organizational factors in new nuclear installations. (P.A.)

23 CFR Appendix A to Part 772 - National Reference Energy Mean Emission Levels as a Function of Speed

Science.gov (United States)

2010-04-01

... 23 Highways 1 2010-04-01 2010-04-01 false National Reference Energy Mean Emission Levels as a Function of Speed A Appendix A to Part 772 Highways FEDERAL HIGHWAY ADMINISTRATION, DEPARTMENT OF... NOISE Pt. 772, App. A Appendix A to Part 772—National Reference Energy Mean Emission Levels as a...

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

International Nuclear Information System (INIS)

2008-12-01

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the

Safety Training: places available in April 2014

CERN Multimedia

Safety Training team, HSE Unit

2014-01-01

There are places available in the forthcoming Safety courses. For updates and registrations, please refer to the Safety Training Catalogue (see here).   April 2014 (alphabetical order) ATEX Habilitation - Level 2 03-APR-14 to 04-APR-14, 9.00 – 17.30, in French Cryogenic Safety - Level 1 10-APR-14, 10.00 – 12.00, in English Electrical Palett Truck – Driving 15-APR-14, 8.30 – 12.30, in French (hand-outs in English for non-French-speaking participants) Fire Extinguisher 09-APR-14, 10.30 – 12.00, in French 16-APR-14, 10.30 – 12.00, in English 23-APR-14, 10.30 – 12.00, in English First Aider - Level 1 – Initial 03-APR-14, 8.30 – 17.30, in French 10-APR-14, 8.30 – 17.30, in English 16-APR-14, 8.30 – 17.30, in English 24-APR-14, 8.30 – 17.30, in English First Aider - Level 1 – Refresher 17-APR-14, 8.30 – 12.30, in English 17-APR-14, 13.30 – 17.30, in...

Planning exercise for the resolution of high level waste tank safety issues

International Nuclear Information System (INIS)

Bunting, J.; Saveland, J.

1992-01-01

Several conditions have been found to exist within high level radioactive waste storage tanks at the Hanford site which could lead to uncontrolled exothermic reactions and/or to the release of tank contents into the environment. These conditions have led to the establishment of four priority 1 safety issues for the Hanford tanks. Resolution of these safety issues will require the coordinated efforts of professionals in chemical, nuclear, operations, safety, and other technical areas. A coordinated and integrated approach is necessary in order to achieve resolution in the shortest possible time, while ensuring that the steps taken do not complicate the later jobs of vitrification and ultimate disposal. This paper describes the purpose, process, and results of an effort to develop a suggested approach. (author)

Patient safety climate profiles across time: Strength and level of safety climate associated with a quality improvement program in Switzerland—A cross-sectional survey study

Science.gov (United States)

Mascherek, Anna C.

2017-01-01

Safety Climate has been acknowledged as an unspecific factor influencing patient safety. However, studies rarely provide in-depth analysis of climate data. As a helpful approach, the concept of “climate strength” has been proposed. In the present study we tested the hypotheses that even if safety climate remains stable on mean-level across time, differences might be evident in strength or shape. The data of two hospitals participating in a large national quality improvement program were analysed for differences in climate profiles at two measurement occasions. We analysed differences on mean-level, differences in percent problematic response, agreement within groups, and frequency histograms in two large hospitals in Switzerland at two measurement occasions (2013 and 2015) applying the Safety Climate Survey. In total, survey responses of 1193 individuals were included in the analyses. Overall, small but significant differences on mean-level of safety climate emerged for some subgroups. Also, although agreement was strong at both time-points within groups, tendencies of divergence or consensus were present in both hospitals. Depending on subgroup and analyses chosen, differences were more or less pronounced. The present study illustrated that taking several measures into account and describing safety climate from different perspectives is necessary in order to fully understand differences and trends within groups and to develop interventions addressing the needs of different groups more precisely. PMID:28753633

Arterial blood gas reference values for sea level and an altitude of 1,400 meters.

Science.gov (United States)

Crapo, R O; Jensen, R L; Hegewald, M; Tashkin, D P

1999-11-01

Blood gas measurements were collected on healthy lifetime nonsmokers at sea level (n = 96) and at an altitude of 1,400 meters (n = 243) to establish reference equations. At each study site, arterial blood samples were analyzed in duplicate on two separate blood gas analyzers and CO-oximeters. Arterial blood gas variables included Pa(O(2)), Pa(CO(2)), pH, and calculated alveolar-arterial PO(2) difference (AaPO(2)). CO-oximeter variables were Hb, COHb, MetHb, and Sa(O(2)). Subjects were 18 to 81 yr of age with 166 male and 173 female. Outlier data were excluded from multiple regression analysis, and reference equations were fitted to the data in two ways: (1) best fit using linear, squared, and cross-product terms; (2) simple equations, including only the variables that explained at least 3% of the variance. Two sets of equations were created: (1) using only the sea level data and (2) using the combined data with barometric pressure as an independent variable. Comparisons with earlier studies revealed small but significant differences; the decline in Pa(O(2)) with age at each altitude was consistent with most previous studies. At sea level, the equation that included barometric pressure predicted Pa(O(2)) slightly better than the sea level specific equation. The inclusion of barometric pressure in the equations allows better prediction of blood gas reference values at sea level and at altitudes as high as 1,400 meters.

Pilot Study on Harmonisation of Reactor Safety in WENRA Countries

International Nuclear Information System (INIS)

2003-03-01

Most of the objectives, set for the Pilot Study, were met. It can be concluded that the methodology was adequate for its purpose. National requirements on selected safety issues have been systematically compared and the major gaps and differences have been identified. Convenient overviews have been provided of differences and similarities between the countries. Furthermore, the conclusions are based on a safety justification and are detailed enough to provide input to a further more detailed analysis on the national level. It was not possible, however, to provide fully verified conclusions about the implementation of the reference levels in the different countries. This has to do with the following constraints on the study: In line with the Terms of Reference, the comparison of formal requirements did not address the more detailed use of criteria and methods to verify compliance. The same requirement could be enforced differently in different regulatory systems, and hence lead to different implementation. The Pilot Study also assessed the implementation, but it was not possible to do this in sufficient detail to identify such differences. The implementation was assessed on the basis of current knowledge of the respective regulatory body, but it was not possible to provide the panels with evidence of the implementation. For these reasons, conclusions about implemented safety provisions in the different countries should be drawn with precaution. The introduction of the panel assessments greatly improved the quality and consistency of the comparison assessments. Uncertainties in the assessments are mainly connected with lack of time to make a detailed analysis in some cases. The reliability of the assessments seems to be sufficient for the objectives of the Pilot Study. The introduction of the IAEA safety standards in the study proved to be helpful and provided confidence in the scope and strictness of the reference levels. This Pilot Study has contributed to

Nanoscale reference materials for environmental, health and safety measurements: needs, gaps and opportunities.

Science.gov (United States)

Stefaniak, Aleksandr B; Hackley, Vincent A; Roebben, Gert; Ehara, Kensei; Hankin, Steve; Postek, Michael T; Lynch, Iseult; Fu, Wei-En; Linsinger, Thomas P J; Thünemann, Andreas F

2013-12-01

The authors critically reviewed published lists of nano-objects and their physico-chemical properties deemed important for risk assessment and discussed metrological challenges associated with the development of nanoscale reference materials (RMs). Five lists were identified that contained 25 (classes of) nano-objects; only four (gold, silicon dioxide, silver, titanium dioxide) appeared on all lists. Twenty-three properties were identified for characterisation; only (specific) surface area appeared on all lists. The key themes that emerged from this review were: 1) various groups have prioritised nano-objects for development as "candidate RMs" with limited consensus; 2) a lack of harmonised terminology hinders accurate description of many nano-object properties; 3) many properties identified for characterisation are ill-defined or qualitative and hence are not metrologically traceable; 4) standardised protocols are critically needed for characterisation of nano-objects as delivered in relevant media and as administered to toxicological models; 5) the measurement processes being used to characterise a nano-object must be understood because instruments may measure a given sample in a different way; 6) appropriate RMs should be used for both accurate instrument calibration and for more general testing purposes (e.g., protocol validation); 7) there is a need to clarify that where RMs are not available, if "(representative) test materials" that lack reference or certified values may be useful for toxicology testing and 8) there is a need for consensus building within the nanotechnology and environmental, health and safety communities to prioritise RM needs and better define the required properties and (physical or chemical) forms of the candidate materials.

Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

International Nuclear Information System (INIS)

Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

2014-01-01

Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

Annual plan of research on safety techniques against low level radioactive wastes, 1984-1988

International Nuclear Information System (INIS)

1984-01-01

The establishment of the countermeasures for treating and disposing radioactive wastes has become an important subject for promoting the utilization of atomic energy. Especially as to low level radioactive wastes, the cumulative quantity has reached about 460,000 in terms of 200 l drums as of the end of March, 1983, and accompanying the development of the utilization of atomic energy, its rapid increase is expected. So far, as for the disposal of low level radioactive wastes, the research and development and the preparation of safety criteria and safety evaluation techniques have been carried out, following the basic policy of the Atomic Energy Commission to execute land disposal and ocean disposal in combination, first to make the test disposal after preliminary safety evaluation, and to shift to the full scale disposal based on the results. The annual plan was decided on July 22, 1983, and the first revision was carried out this time, therefore, it is reported here. The basic policy of establishing this annual plan, and the annual plan for safety technique research are described. (Kako, I.)

Basic Restriction and Reference Level in Anatomically-based Japanese Models for Low-Frequency Electric and Magnetic Field Exposures

Science.gov (United States)

Takano, Yukinori; Hirata, Akimasa; Fujiwara, Osamu

Human exposed to electric and/or magnetic fields at low frequencies may cause direct effect such as nerve stimulation and excitation. Therefore, basic restriction is regulated in terms of induced current density in the ICNIRP guidelines and in-situ electric field in the IEEE standard. External electric or magnetic field which does not produce induced quantities exceeding the basic restriction is used as a reference level. The relationship between the basic restriction and reference level for low-frequency electric and magnetic fields has been investigated using European anatomic models, while limited for Japanese model, especially for electric field exposures. In addition, that relationship has not well been discussed. In the present study, we calculated the induced quantities in anatomic Japanese male and female models exposed to electric and magnetic fields at reference level. A quasi static finite-difference time-domain (FDTD) method was applied to analyze this problem. As a result, spatially averaged induced current density was found to be more sensitive to averaging algorithms than that of in-situ electric field. For electric and magnetic field exposure at the ICNIRP reference level, the maximum values of the induced current density for different averaging algorithm were smaller than the basic restriction for most cases. For exposures at the reference level in the IEEE standard, the maximum electric fields in the brain were larger than the basic restriction in the brain while smaller for the spinal cord and heart.

Nuclear power plant safety

International Nuclear Information System (INIS)

Otway, H.J.

1974-01-01

Action at the international level will assume greater importance as the number of nuclear power plants increases, especially in the more densely populated parts of the world. Predictions of growth made prior to October 1973 [9] indicated that, by 1980, 14% of the electricity would be supplied by nuclear plants and by the year 2000 this figure would be about 50%. This will make the topic of international co-operation and standards of even greater importance. The IAEA has long been active in providing assistance to Member States in the siting design and operation of nuclear reactors. These activities have been pursued through advisory missions, the publication of codes of practice, guide books, technical reports and in arranging meetings to promote information exchange. During the early development of nuclear power, there was no well-established body of experience which would allow formulation of internationally acceptable safety criteria, except in a few special cases. Hence, nuclear power plant safety and reliability matters often received an ad hoc approach which necessarily entailed a lack of consistency in the criteria used and in the levels of safety required. It is clear that the continuation of an ad hoc approach to safety will prove inadequate in the context of a world-wide nuclear power industry, and the international trade which this implies. As in several other fields, the establishment of internationally acceptable safety standards and appropriate guides for use by regulatory bodies, utilities, designers and constructors, is becoming a necessity. The IAEA is presently planning the development of a comprehensive set of basic requirements for nuclear power plant safety, and the associated reliability requirements, which would be internationally acceptable, and could serve as a standard frame of reference for nuclear plant safety and reliability analyses

Tank characterization reference guide

International Nuclear Information System (INIS)

De Lorenzo, D.S.; DiCenso, A.T.; Hiller, D.B.; Johnson, K.W.; Rutherford, J.H.; Smith, D.J.; Simpson, B.C.

1994-09-01

Characterization of the Hanford Site high-level waste storage tanks supports safety issue resolution; operations and maintenance requirements; and retrieval, pretreatment, vitrification, and disposal technology development. Technical, historical, and programmatic information about the waste tanks is often scattered among many sources, if it is documented at all. This Tank Characterization Reference Guide, therefore, serves as a common location for much of the generic tank information that is otherwise contained in many documents. The report is intended to be an introduction to the issues and history surrounding the generation, storage, and management of the liquid process wastes, and a presentation of the sampling, analysis, and modeling activities that support the current waste characterization. This report should provide a basis upon which those unfamiliar with the Hanford Site tank farms can start their research

Diagnostic reference levels in digital mammography: a systematic review

International Nuclear Information System (INIS)

Suleiman, Moayyad E.; Brennan, Patrick C.; McEntee, Mark F.

2015-01-01

This study aims to review the literature on existing diagnostic reference levels (DRLs) in digital mammography and methodologies for establishing them. To this end, a systematic search through Medline, Cinahl, Web of Science, Scopus and Google scholar was conducted using search terms extracted from three terms: DRLs, digital mammography and breast screen. The search resulted in 1539 articles of which 22 were included after a screening process. Relevant data from the included studies were summarised and analysed. Differences were found in the methods utilised to establish DRLs including test subjects types, protocols followed, conversion factors employed, breast compressed thicknesses and percentile values adopted. These differences complicate comparison of DRLs among countries; hence, an internationally accepted protocol would be valuable so that international comparisons can be made. (authors)

Evaluation of Patient and Family Engagement Strategies to Improve Medication Safety.

Science.gov (United States)

Kim, Julia M; Suarez-Cuervo, Catalina; Berger, Zackary; Lee, Joy; Gayleard, Jessica; Rosenberg, Carol; Nagy, Natalia; Weeks, Kristina; Dy, Sydney

2018-04-01

Patient and family engagement (PFE) is critical for patient safety. We systematically reviewed types of PFE strategies implemented and their impact on medication safety. We searched MEDLINE, EMBASE, reference lists and websites to August 2016. Two investigators independently reviewed all abstracts and articles, and articles were additionally reviewed by two senior investigators for selection. One investigator abstracted data and two investigators reviewed the data for accuracy. Study quality was determined by consensus. Investigators developed a framework for defining the level of patient engagement: informing patients about medications (Level 1), informing about engagement with health care providers (Level 2), empowering patients with communication tools and skills (Level 3), partnering with patients in their care (Level 4), and integrating patients as full care team members (Level 5). We included 19 studies that mostly targeted older adults taking multiple medications. The median level of engagement was 2, ranging from 2-4. We identified no level 5 studies. Key themes for patient engagement strategies impacting medication safety were patient education and medication reconciliation, with a subtheme of patient portals. Most studies (84%) reported implementation outcomes. The most commonly reported medication safety outcomes were medication errors, including near misses and discrepancies (47%), and medication safety knowledge (37%). Most studies (63%) were of medium to low quality, and risk of bias was generally moderate. Among the 11 studies with control groups, 55% (n = 6) reported statistically significant improvement on at least one medication safety outcome. Further synthesis of medication safety measures was limited due to intervention and outcome heterogeneity. Key strategies for engaging patients in medication safety are education and medication reconciliation. Patient engagement levels were generally low, as defined by a novel framework for determining

Nuclear safety

International Nuclear Information System (INIS)

Tarride, Bruno

2015-10-01

The author proposes an overview of methods and concepts used in the nuclear industry, at the design level as well as at the exploitation level, to ensure an acceptable safety level, notably in the case of nuclear reactors. He first addresses the general objectives of nuclear safety and the notion of acceptable risk: definition and organisation of nuclear safety (relationships between safety authorities and operators), notion of acceptable risk, deterministic safety approach and main safety principles (safety functions and confinement barriers, concept of defence in depth). Then, the author addresses the safety approach at the design level: studies of operational situations, studies of internal and external aggressions, safety report, design principles for important-for-safety systems (failure criterion, redundancy, failure prevention, safety classification). The next part addresses safety during exploitation and general exploitation rules: definition of the operation domain and of its limits, periodic controls and tests, management in case of incidents, accidents or aggressions

Development Trends in Nuclear Technology and Related Safety Aspects

International Nuclear Information System (INIS)

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

2002-01-01

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

Hospitality Services Reference Book.

Science.gov (United States)

Texas Tech Univ., Lubbock. Home Economics Curriculum Center.

This reference book provides information needed by employees in hospitality services occupations. It includes 29 chapters that cover the following topics: the hospitality services industry; professional ethics; organization and management structures; safety practices and emergency procedures; technology; property maintenance and repair; purchasing…

Modelling blood safety

NARCIS (Netherlands)

Janssen, M.P.

2010-01-01

This thesis describes the development and application of methods and models to support decision making on safety measures aimed at preventing the transmission of infections by blood donors. Safety measures refer to screening tests for blood donors, quarantine periods for blood plasma, or methods for

Derivation of a reference dose and drinking water equivalent level for 1,2,3-trichloropropane.

Science.gov (United States)

Tardiff, Robert G; Carson, M Leigh

2010-06-01

In some US potable water supplies, 1,2,3-trichloropropane (TCP) has been present at ranges of non-detect to less than 100 ppb, resulting from past uses. In subchronic oral studies, TCP produced toxicity in kidneys, liver, and other tissues. TCP administered by corn oil gavage in chronic studies produced tumors at multiple sites in rats and mice; however, interpretation of these studies was impeded by substantial premature mortality. Drinking water equivalent levels (DWELs) were estimated for a lifetime of consumption by applying biologically-based safety/risk assessment approaches, including Monte Carlo techniques, and with consideration of kinetics and modes of action, to possibly replace default assumptions. Internationally recognized Frameworks for human relevance of animal data were employed to interpret the findings. Calculated were a reference dose (=39 microg/kg d) for non-cancer and Cancer Values (CV) (=10-14 microg/kg d) based on non-linear dose-response relationships for mutagenicity as a precursor of cancer. Lifetime Average Daily Intakes (LADI) are 3130 and 790-1120 microg/person-d for non-cancer and cancer, respectively. DWELs, estimated by applying a relative source contribution (RSC) of 50% to the LADIs, are 780 and 200-280 microg/L for non-cancer and cancer, respectively. These DWELs may inform establishment of formal/informal guidelines and standards to protect public health. Copyright 2010 Elsevier Ltd. All rights reserved.

MP-Division health and safety reference handbook. [Contains glossary

Energy Technology Data Exchange (ETDEWEB)

Putnam, T.M.

1987-09-01

This report presents the objectives, organization, policies, and essential rules and procedures that have been adopted by MP Division and that form the basis of the Health and Safety Program of the Clinton P. Anderson Meson Physics Facility (LAMPF). The facility includes the beam-delivery systems for the Los Alamos Neutron Scattering Center and the Weapons Neutron Research Facility (LANSCE/WNR). The program is designed not only to assure the health and safety of all personnel, including users, in their work at LAMPF, and of MP-Division staff in their work on the LANSCE/WNR beam lines, but also to protect the facility (buildings and equipment) and the environment. 33 refs., 18 figs., 2 tabs.

2005 dossier: granite. Tome: safety analysis of the geologic disposal

International Nuclear Information System (INIS)

2005-01-01

This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

The application of the Appropriate Level of Protection (ALOP) and Food Safety Objective (FSO) concepts in food safety management, using Listeria monocytogenes in deli meats as a case study

NARCIS (Netherlands)

Gkogka, E.; Reij, M.W.; Gorris, L.G.M.; Zwietering, M.H.

2011-01-01

To establish a link between governmental food safety control and operational food safety management, the concepts of the Appropriate Level of Protection (ALOP) and the Food Safety Objective (FSO) have been suggested by international bodies as a means of making food safety control transparent and

IAEA safety standards and approach to safety of advanced reactors

International Nuclear Information System (INIS)

Gasparini, M.

2004-01-01

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

Deepening om Safety culture Auto evaluation; Profundizacion en cultura de seguridad. Autoevaluacion

Energy Technology Data Exchange (ETDEWEB)

Lopez Churruca, I.; Buedo Jimenez, J. L.

2009-07-01

The concept of safety culture used in nuclear forums refers to the series of actions aimed at guaranteeing that safety issues in nuclear power plants are adequately addressed. The activities to which this concept refers have been gradually extended over the years so that, today, they encompass the whole organizations structure. In other words, the safety culture implies that all positions in the organizational structure perform their tasks with a level of attention such that all their senses are focused on them. And this performance is what leads us to excellence in plant operation. In addition, the implementation of a self-assessment system in the Cofrentes Nuclear Power Plant of its activities and processes has resulted in the identification of expectations on the basis of which we can identify strengths and weaknesses, enabling us to leverage the former and correct the latter. (Author)

Reference levels at diagnosis (NRD) for explorations in TC of the university Hospital Donostia

International Nuclear Information System (INIS)

Iriondo Igerabide, U.; Puertolas Hernandez, J. R.; Masso Odriozola, A.; Alonso Espinaco, M. t.; Pino Leon, C.; Lozano Flores, F. J.; Larretxea Etxarri, R.

2013-01-01

The objective of this work is the establishment of diagnostic reference levels in TC, for the anatomical regions, in the University Hospital Donostia, in order to reduce the dose to patients and without prejudice to the required diagnosis. (Author)

Establishment of dose reference levels for mammography in Greece

International Nuclear Information System (INIS)

Kalathaki, M.; Hourdakis, C.J.; Economides, S.; Tritakis, P.; Manousaridis, G.; Kalyvas, N.; Simantirakis, G.; Kipouros, P.; Kamenopoulou, V.

2006-01-01

Full text of publication follows: Diagnostic Reference Levels (D.R.L.) are dose levels established in medical practices for typical x-ray examinations concerning groups of standard size patients or standard phantoms and broadly defined types of equipment. When good and normal practice is performed, these levels are not expected to be exceeded. This work is an attempt to establish for the first time the D.R.L. for mammography in Greece. At present, there are 402 mammographic systems in clinical use all over the country. This study that lasted 3 years (2000-2003), includes 117 of these systems, 85% of which are installed in private and 15% in public sector countrywide. Measurements of entrance surface dose (E.S.D.) were performed as a part of the regular inspections performed by the Licensing and Inspections Department of Greek Atomic Energy Commission on the basis of the laboratories licensing procedure. Moreover, the entire performance of the mammographic units was assessed by quantitative and qualitative measurements of specific parameters. In order to establish the national D.R.L., a standard phantom was used during the quality control of the mammographic units and E.S.D. measurements were performed based on the clinical practice of each laboratory. The D.R.L. for this type of examination was established according to the 75. percentile of the E.S.D. curve and found equal to 7 mGy per single view. The comparison of this value with the one reported by the European Commission (10 mGy per view), indicates that the D.R.L. for mammography is lower in Greece. However, the primary concern of a mammographic examination is to keep breast dose as low as reasonably achievable while providing images with the maximum amount of diagnostic information. The quality of the produced images was therefore assessed for all systems examined, regardless of meeting or exceeding the quality criteria reference surface entrance dose. The results showed that the average total score of the

Certification plan for safety and PRA codes

International Nuclear Information System (INIS)

Toffer, H.; Crowe, R.D.; Ades, M.J.

1990-05-01

A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA ampersand PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan's objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations

Reference levels in PTCA as a function of procedure complexity

International Nuclear Information System (INIS)

Peterzol, A.; Quai, E.; Padovani, R.; Bernardi, G.; Kotre, C. J.; Dowling, A.

2005-01-01

The multicentre assessment of a procedure complexity index (CI) for the introduction of reference levels (RLs) in percutaneous transluminal coronary angio-plasties (PTCA) is presented here. PTCAs were investigated based on methodology proposed by Bernardi et al. Multiple linear stepwise regression analysis, including clinical, anatomical and technical factors, was performed to obtain fluoroscopy time predictors. Based on these regression coefficients, a scoring system was defined and CI obtained. CI was used to classify dose values into three groups: low, medium and high complexity procedures, since there was good correlation (r = 0.41; P 2 , and 12, 20 and 27 min for fluoroscopy time, for the three CI groups. (authors)

Development of a low-level radon reference chamber; Entwicklung einer Low-Level-Radon-Referenzkammer

Energy Technology Data Exchange (ETDEWEB)

Linzmaier, Diana

2013-01-04

The naturally occurring, radioactive noble gas radon-222 exists worldwide in different activity concentrations in the air. During the decay of radon-222, decay products are generated which are electrically charged and attach to aerosols in the air. Together with the aerosols, the radon is inhaled and exhaled by humans. While the radon is nearly completely exhaled, ca. 20 % of the inhaled aerosols remain in the lungs in one breath cycle. Due to ionizing radiation, in a chain of events, lung cancer might occur. Consequently, radon and its decay products are according to the current findings the second leading cause of lung cancer. At the workplace and in the home measurements of radon activity concentration are performed to determine the radiation exposition of humans. All measurement devices for the determination of radon activity concentration are calibrated above 1000 Bq/m{sup 3}, even though the mean value of the present investigation in Germany shows only 50 Bq/m{sup 3}. For the calibration of measurement devices in the range below 1000 Bq/m{sup 3} over a long time period, the generation of a stable reference atmosphere is presented in this work. Due to a long term calibration (t>5 days) of the measurement devices, smaller uncertainties result for the calibration factor. For the calibration procedure, a so-called low-level radon reference chamber was set up and started operation. The generation of a stable reference atmosphere is effected by means of emanation sources which consist of a radium-226 activity standard. On the basis of {gamma}-spectrometry, the effective emanation coefficient ofthe emanation sources is determined. The traceability of the activity concentration in the reference volume is realized via the activity ofthe radium-226, the emanation coefficient and the volume. With the emanation sources produced, stable reference atmospheres within the range of 150 Bq/m{sup 3} to 1900 Bq/m{sup 3} are achieved. For the realization, maintenance and

Project Guarantee 1985. Final repository for high-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

Final disposal of radioactive waste involves preventing the waste from returning from the repository location into the biosphere by means of successively arranged containment measures known as safety barriers. In the present volume NGB 85-04 of the series of reports for Project 'Guarantee' 1985, the safety barrier system for the type C repository for high-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). Safety barriers take the form of both technically constructed containment measures and the siting of the repository in suitable geological formations. The technical safety barrier system in the case of high-level waste comprises: the waste solidification matrix (borosilicate glass), massive steel canisters, encasement of the waste canisters, encasement of the waste canisters in highly compacted bentonite, sealing of vacant storage space and access routes on repository closure. The natural geological safety barriers - the host rock and overlying formations provide sufficiently long deep groundwater flow times from the repository location to the earth's surface and for additional lengthening of radionuclide migration times by means of various chemical and physical retardation mechanisms. The stability of the geological formations is so great that hydrogeological system is protected for a sufficient length of time from deterioration caused, in particular, by erosion. Observations in the final section of the report indicate that input data for the type C repository safety

The role and impact of reference doses on diagnostic radiology, how to use them at the national level?

International Nuclear Information System (INIS)

Nikodemova, D.; Horvathova, M.; Karkus, R.

2003-01-01

Results of patient dose audits reported in this paper for several types of examinations and various technical units have shown the importance of applications of reference dose levels in radiological practice. On the basis of national surveys slightly lower or higher standard dose reference levels (DRL) values could be justified. Continuing revision of DRL values and their extension to other types of radiographic and fluoroscopic examinations is needed

Technical Insight of the High Level Safety Goal for the NPPs Built in China’S Thirteenth Five-Year Period (2016-2020)

Energy Technology Data Exchange (ETDEWEB)

Shi, G.; Zhan, W.; Mei, Q.; Sun, D., E-mail: shi@snerdi.com.cn [Shanghai Nuclear Engineering Research and Design Institute, SNPTC Shanghai (China)

2014-10-15

The “Nuclear Safety Planning” has been published in Oct. 2012 in China, which stipulates the safety goals for the NPPs which will be built in the future. As for the NPPs which will be built in China's Thirteenth Five-Year (2016-2020) and later, the high level safety goal is described as “the possibility of the large radioactive release should be practically eliminated by design”. A thorough investigation has been performed at SNERDI to explore the technical insights of this high level safety goal by using MEDP hierarchical safety goal approach. The definition of large release is proposed accordingly, DID requirements and probabilistic requirements are derived from this high level safety goal. (author)

Food Safety and Sustainability ‒ An Exploratory Approach at the Level of the Romanian Wine Production Companies

Directory of Open Access Journals (Sweden)

Gheorghe Epuran

2018-02-01

Full Text Available This article aims at identifying and analyzing opinions referring to the necessity, opportunity and efficiency of the implementation of the strategies and sustainable development practices in the wine industry as food safety is considered a component of sustainability. The main objective of the paper is to create a clear picture of the existing willingness of companies to adopt marketing strategies that should contribute to increasing food security by increasing the quality of wine and which should be based on the principles of sustainability. In this sense, the authors conducted some qualitative marketing research in which 15 managers from wine producing companies from Romania's wine-growing areas were interviewed. The research highlights the relationship between responsible consumption, food safety and sustainability, and the authors believe that such an association is relevant to the economic and social context of the European Union and other parts of the world. The novelty of this research refers to the analysis of the relationship between the responsible wine consumption and sustainability, to the identification of the opinions of the wine companies managers regarding the benefits and the limits of M2M and IoT technologies, to the emphasis of the marketing strategies and techniques considered innovative in the field of viticulture, as well as to the association of the concept of food safety by wine consumption to the aspects related to education and culture.

Dose limits, constraints, reference levels. What does it mean for radiation protection?

International Nuclear Information System (INIS)

Breckow, J.

2016-01-01

The established concept of radiation protection with its basic principles justification, optimization, and limitation has proved its value and is going to be continued. In its deeper meaning, however, the concept is rather subtle and complex. Furthermore, in some aspects there remain some breaches or inconsistencies. This is just true for the terms dose limit, reference lever, and constraint that are tightly associated with the radiation protection principles. In order to guarantee the ability of radiation protection in whole extent, the subtle differences of meaning have to be communicated. There is a permanent need to defend the conceptual function of these terms against deliberate or undeliberate misinterpretations. Reference levels are definitely not the same as dose limits and they may not be misused as such. Any attempt to misinterpret fundamental radiation protection principles for selfish purposes should discouraged vigorously.

Nuclear Safety

International Nuclear Information System (INIS)

1978-09-01

In this short paper it has only been possible to deal in a rather general way with the standards of safety used in the UK nuclear industry. The record of the industry extending over at least twenty years is impressive and, indeed, unique. No other industry has been so painstaking in protection of its workers and in its avoidance of damage to the environment. Headings are: introduction; how a nuclear power station works; radiation and its effects (including reference to ICRP, the UK National Radiological Protection Board, and safety standards); typical radiation doses (natural radiation, therapy, nuclear power programme and other sources); safety of nuclear reactors - design; key questions (matters of concern which arise in the public mind); safety of operators; safety of people in the vicinity of a nuclear power station; safety of the general public; safety bodies. (U.K.)

Standardization of reference radiation field of beta for 85Kr using extrapolation chamber

International Nuclear Information System (INIS)

Nazaroh; Fendinugroho

2013-01-01

Standardization of reference radiation field of beta for 85 Kr in PTKMR-BATAN Laboratory has been performed at the SDD's 30 cm by using extrapolation chamber detector, coupled with Uni dose electrometer. The result was : (8.98±3 %) mGy/h, at 95 % confidence level. The aim of standardization of reference radiation field is to support radiation protection and safety program, provided by the International Atomic Energy Agency to its Member States, included BATAN-Indonesia, especially, PTKMR. The aim of radiation protection program and safety program is to promote an internationally harmonized approach for radiation measurement in protection level, besides for calibration of radiation measuring instrument, which users spread across Indonesia, with the number of about 795 firms in the year of 2012. These benefits can be felt by workers, communities and the environment, because by calibration, measurement survey meter, pocket dosimeter and TLD to be more accurate so that the radiation dose received by radiation workers is accurate and can be ascertained in a specified period, not to exceed a predetermined NBD by BAPETEN. The aim of this calibration is appropriate with the primary objective of calibration on IAEA/TRS16:2000. (author)

Probabilistic safety assessment for Hanford high-level waste tanks

International Nuclear Information System (INIS)

MacFarlane, D.R.; Stack, D.S.; Kindinger, J.P.; Deremer, R.K.

1995-01-01

This paper gives results from the first comprehensive level-3 probabilistic safety assessment (PSA), including consideration of external events, for the Hanford tank farm (HTF). This work was sponsored by the U.S. Department of Energy/Environmental Restoration and Waste Management Division (DOE/EM). At the HTF, there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/saltcake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is ∼60 million gal, containing ∼200 million Ci of radioactivity

Demonstration tests for low level radioactive waste packaging safety

International Nuclear Information System (INIS)

Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

1993-01-01

The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

Technological progress, safety, and the guardian role of inspection

Energy Technology Data Exchange (ETDEWEB)

Critchley, O H

1981-08-01

Technological innovation is accompanied by unforeseen human consequences as well as benefits, and progress has produced a public awareness of the potential for hazards that has led to efficient safety-inspection procedures. Because no safety procedure is foolproof, the public learns to tolerate certain levels of risk from technology if it concludes that the benefits are worthwhile. The perception of values often transcends simple cost/benefit analysis. Safety technology and regulation developed during the past 50 years has benefited from earlier disastrous accidents enough to give the nuclear power industry an unprecedented safety record. Efforts to understand and anticipate human error have refined the role of inspection without achieving absolute reliability. Well-directed inspections that accept human fallibility will achieve more than design and organizational improvements. 42 references. (DCK)

Optimum Safety Levels and Design Rules for the Icelandic-Type Berm Breakwater

DEFF Research Database (Denmark)

Sigurdarson, Sigurdur; van der Meer, Jentsje W.; Burcharth, Hans F.

2009-01-01

strategies and possible failure with corresponding downtime have been taken into account, as well as actual market prices (in Iceland and Norway) for rack material and construction. Calculations show that low stability numbers for the largest rock armour layer give the optimal safety level....

Establishment of multi-slice computed tomography (MSCT) reference level in Johor, Malaysia

Science.gov (United States)

Karim, M. K. A.; Hashim, S.; Bakar, K. A.; Muhammad, H.; Sabarudin, A.; Ang, W. C.; Bahruddin, N. A.

2016-03-01

Radiation doses from computed tomography (CT) are the highest and most hazardous compared to other imaging modalities. This study aimed to evaluate radiation dose in Johor, Malaysia to patients during computed tomography examinations of the brain, chest and abdomen and to establish the local diagnostic reference levels (DRLs) as are present with the current, state- of-art, multi-slice CT scanners. Survey forms were sent to five centres performing CT to obtain data regarding acquisition parameters as well as the dose information from CT consoles. CT- EXPO (Version 2.3.1, Germany) was used to validate the dose information. The proposed DRLs were indicated by rounding the third quartiles of whole dose distributions where mean values of CTDIw (mGy), CTDIvol (mGy) and DLP (mGy.cm) were comparable with other reference levels; 63, 63, and 1015 respectively for CT Brain; 15, 14, and 450 respectively for CT thorax and 16, 17, and 590 respectively for CT abdomen. The study revealed that the CT practice and dose output were revolutionised, and must keep up with the pace of introductory technology. We suggest that CTDIvol should be included in current national DRLs, as modern CTs are configured with a higher number of detectors and are independent of pitch factors.

Establishment of multi-slice computed tomography (MSCT) reference level in Johor, Malaysia

International Nuclear Information System (INIS)

Karim, M K A; Hashim, S; Ang, W C; Bahruddin, N A; Bakar, K A; Muhammad, H; Sabarudin, A

2016-01-01

Radiation doses from computed tomography (CT) are the highest and most hazardous compared to other imaging modalities. This study aimed to evaluate radiation dose in Johor, Malaysia to patients during computed tomography examinations of the brain, chest and abdomen and to establish the local diagnostic reference levels (DRLs) as are present with the current, state- of-art, multi-slice CT scanners. Survey forms were sent to five centres performing CT to obtain data regarding acquisition parameters as well as the dose information from CT consoles. CT- EXPO (Version 2.3.1, Germany) was used to validate the dose information. The proposed DRLs were indicated by rounding the third quartiles of whole dose distributions where mean values of CTDI w (mGy), CTDI vol (mGy) and DLP (mGy.cm) were comparable with other reference levels; 63, 63, and 1015 respectively for CT Brain; 15, 14, and 450 respectively for CT thorax and 16, 17, and 590 respectively for CT abdomen. The study revealed that the CT practice and dose output were revolutionised, and must keep up with the pace of introductory technology. We suggest that CTDI vol should be included in current national DRLs, as modern CTs are configured with a higher number of detectors and are independent of pitch factors. (paper)

Proposal for basic safety requirements regarding the disposal of high-level radioactive waste

International Nuclear Information System (INIS)

1980-04-01

A working group commissioned to prepare proposals for basic safety requirements for the storage and transport of radioactive waste prepared its report to the Danish Agency of Environmental Protection. The proposals include: radiation protection requirements, requirements concerning the properties of high-level waste units, the geological conditions of the waste disposal location, the supervision of waste disposal areas. The proposed primary requirements for safety evaluation of the disposal of high-level waste in deep geological formations are of a general nature, not being tied to specific assumptions regarding the waste itself, the geological and other conditions at the place of disposal, and the technical methods of disposal. It was impossible to test the proposals for requirements on a working repository. As no country has, to the knowledge of the working group, actually disposed of hifg-level radioactive waste or approved of plans for such disposal. Methods for evaluating the suitability of geological formations for waste disposal, and background material concerning the preparation of these proposals for basic safety requirements relating to radiation, waste handling and geological conditions are reviewed. Appended to the report is a description of the phases of the fuel cycle that are related to the storage of spent fuel and the disposal of high-level reprocessing waste in a salt formation. It should be noted that the proposals of the working group are not limited to the disposal of reprocessed fuel, but also include the direct disposal of spent fuel as well as disposal in geological formations other than salt. (EG)

Safety analysis for 'Fugen'

International Nuclear Information System (INIS)

1997-10-01

The improvement of safety in nuclear power stations is an important proposition. Therefore also as to the safety evaluation, it is important to comprehensively and systematically execute it by referring to the operational experience and the new knowledge which is important for the safety throughout the period of use as well as before the construction and the start of operation of nuclear power stations. In this report, the results when the safety analysis for ''Fugen'' was carried out by referring to the newest technical knowledge are described. As the result, it was able to be confirmed that the safety of ''Fugen'' has been secured by the inherent safety and the facilities which were designed for securing the safety. The basic way of thinking on the safety analysis including the guidelines to be conformed to is mentioned. As to the abnormal transient change in operation and accidents, their definition, the events to be evaluated and the standards for judgement are reported. The matters which were taken in consideration at the time of the analysis are shown. The computation programs used for the analysis were REACT, HEATUP, LAYMON, FATRAC, SENHOR, LOTRAC, FLOOD and CONPOL. The analyses of the abnormal transient change in operation and accidents are reported on the causes, countermeasures, protective functions and results. (K.I.)

Burnout, Perceived Stress, and Job Satisfaction Among Trauma Nurses at a Level I Safety-Net Trauma Center.

Science.gov (United States)

Munnangi, Swapna; Dupiton, Lynore; Boutin, Anthony; Angus, L D George

Nurses are at the forefront of our health care delivery system and have been reported to exhibit a high level of burnout. Burnout and stress in trauma nurses at a safety-net hospital can negatively impact patient care. Safety-net hospitals are confronted with unique social, financial, as well as resource problems that can potentially make the work environment frustrating. The purpose of this study was to explore the levels of burnout, stress, and job satisfaction in nurses providing care to trauma patients at a Level I safety-net trauma center. A cross-sectional survey design was used to investigate principal factors including personal and professional demographics, burnout, perceived stress, and job satisfaction. Trauma nurses working at a Level I safety-net trauma center are stressed and exhibited moderate degree of burnout. The extent of emotional exhaustion experienced by the nurses varied with work location and was highest in surgical intensive care unit nurses. The level of job satisfaction in terms of opportunities for promotion differed significantly by race and the health status of the nurses. Satisfaction with coworkers was lowest in those nurses between the ages of 60-69 years. Female nurses were more satisfied with their coworkers than male nurses. In addition, the study revealed that significant relationships exist among perceived stress, burnout, and job satisfaction. Work environment significantly impacts burnout, job satisfaction, and perceived stress experienced by trauma nurses in a safety-net hospital. Nursing administration can make an effort to understand the levels of burnout and strategically improve work environment for trauma nurses in order to minimize stressors leading to attrition and enhance job satisfaction.

Estimates of diagnostic reference levels for common radiographic x-ray examinations in some sudanese hospitals

International Nuclear Information System (INIS)

Awad, Ezdehar Mohammed Satti.

2016-03-01

The objective of this study were to estimate patient dose in some common diagnostic x-ray examinations in Sudan with the aim to propose national diagnostic reference levels (DRLs). Radiation doses were estimated for the patients in 23 public hospitals in different town in Sudan (Wad-Madani, Kasala, Atbara Al obaid, Al nhod, Khartoum). Entrance surface air kerma (ESAK) was estimated in a three steps protocol: First, x-ray unit output Y (d) was measured at a distance at a distance (d) for different peak tube voltages and tube loading (mAs). Next, incident air kerma (k) was calculated from Y (d) using inverse square law combined with patient exposure factors. ESAK was calculated from k using backscatter factor, B. Mean ESAK values were comparable to those reported in other countries and are below reference dose levels. The estimated mean ESAK values were: 0.4, 1.9, 1.8, 3.2, 2.4, 3.5, and 8.4 mGy for chests . The estimated mean ESAK values were 0.4, 1.9, 1.8, 3.2, 2.4, 3,5, and 8.4 mGy for chest PA, Skull AP/PA, Skull LAT, Abdomen Pelvis AP, Lumber Spine AP and Lumber Spine LAT examination respectively. The results are used for dose optimization and to propose national diagnostic reference levels. (Author)

Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

International Nuclear Information System (INIS)

2013-01-01

The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

A prediction model for the radiation safety management behavior of medical cyclotrons

International Nuclear Information System (INIS)

Jung, Ji Hye; Han, Eun Ok; Kim, Ssang Tae

2008-01-01

This study attempted to provide reference materials for improving the behavior level in radiation safety managements by drawing a prediction model that affects the radiation safety management behavior because the radiation safety management of medical Cyclotrons, which can be used to produce radioisotopes, is an important factor that protects radiation caused diseases not only for radiological operators but average users. In addition, this study obtained follows results through the investigation applied from January 2 to January 30, 2008 for the radiation safety managers employed in 24 authorized organizations, which have already installed Cyclotrons, through applying a specific form of questionnaire in which the validity was guaranteed by reference study, site investigation, and focus discussion by related experts. The radiation safety management were configured as seven steps: step 1 is a production preparation step, step 2 is an RI production step, step 3 is a synthesis step, step 4 is a distribution step, step 5 is a quality control step, step 6 is a carriage container packing step, and step 7 is a transportation step. It was recognized that the distribution step was the most exposed as 15 subjects (62.5%), the items of 'the sanction and permission related works' and 'the guarantee of installation facilities and production equipment' were the most difficult as 9 subjects (37.5%), and in the trouble steps in such exposure, the item of 'the synthesis and distribution' steps were 4 times, respectively (30.8%). In the score of the behavior level in radiation safety managements, the minimum and maximum scores were 2.42 and 4.00, respectively, and the average score was 3.46 ± 0.47 out of 4. Prosperity and well-being programs in the behavior and job in radiation safety managements (r=0.529) represented a significant correlation statistically. In the drawing of a prediction model based on the factors that affected the behavior in radiation safety managements, general

Central repository for low- and intermediate-level waste (ALMA) conceptual design, siting and safety study

International Nuclear Information System (INIS)

Kjellbert, N.; Haeggblom, H.; Cederstroem, M.; Lundgren, T.

1980-07-01

A generic design, siting and safety study of a proposed repository for low- and intermediate-level waste has been made. Special emphasis has been placed on safety characterostics. The conceptual design and the generic site, on which the study is based, are realistically chosen in accordance with present construction techniques and the existing geohydrological conditions in Sweden. (Auth.)

Safety analysis of the VLJ repository

International Nuclear Information System (INIS)

Vieno, T.; Nordman, H.

1991-05-01

The VLJ repository is an underground disposal facility for the low and medium level waste generated at the Olkiluoto nuclear power plant. The repository is located within 1 km from TVO I and TVO II (2 x 710 MWe) BWR's on the Olkiluoto island at the west coast of Finland. It contains two rock silos excavated at the depth of 60...100 meters in the bedrock. Low level waste will be disposed of in a shotcreted rock silo. For bituminized medium level waste, a separate silo of reinforced concrete has been built inside the shotcreted rock silo. The post-closure safety analysis has been done for the Final Safety Analysis Report (FSAR) of the VLJ repository. In addition to the normal evolution scenario, several disturbed evolution and accident scenarios have been analysed. In the reference scenario, radio-nuclides are assumed to be released from the bituminized waste within 500 years, the concrete silo is assumed to gradually disintegrate and finally to collapse at 5 000 years, all concrete in the silo is assumed to be also chemically depleted within 6 000 years, and all the seals of the repository are assumed to deteriorate within 12 000 years. The ability of alone natural barriers to restrict the release of radionuclides into the biosphere has been evaluated by means of scenarios where the degradation of engineered barriers has been assumed to take place at a still faster rate. In one of the disturbed evolution scenarios it has been assumed that the concrete silo for medium level waste is severely impaired immediately after sealing of the repository. Effects of gas generation and consequences of human intrusion have been evaluated, too. The results of the safety analysis show that radiation doses of any significance are caused only if a well is bored in the vicinity of the repository or if the groundwater discharge spot is inhabited and used for cultivation. In the reference scenario the maximum expectation value of the individual dose rate is 0.3 mSv/a

Preliminary safety concept for disposal of the very low level radioactive waste in Romania.

Science.gov (United States)

Niculae, O; Andrei, V; Ionita, G; Duliu, O G

2009-05-01

In Romania, there are certain nuclear installations in operation or under decommissioning, all of them representing an important source of very low level waste (VLLW). This paper presents an overview on the approach of the VLLW management in Romania, focused on those resulted from the nuclear power plants decommissioning. At the same time, the basic elements of safety concept, together with some safety evaluations concerning VLLW repository are presented and discussed too.

Guidance for the definition and application of probabilistic safety criteria

International Nuclear Information System (INIS)

Holmberg, J.-E.; Knochenhauer, M.

2011-05-01

The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

Guidance for the definition and application of probabilistic safety criteria

Energy Technology Data Exchange (ETDEWEB)

Holmberg, J.-E. (VTT Technical Research Centre of Finland (Finland)); Knochenhauer, M. (Scandpower AB (Sweden))

2011-05-15

The project 'The Validity of Safety Goals' has been financed jointly by NKS (Nordic Nuclear Safety Research), SSM (Swedish Radiation Safety Authority) and the Swedish and Finnish nuclear utilities. The national financing went through NPSAG, the Nordic PSA Group (Swedish contributions) and SAFIR2010, the Finnish research programme on NPP safety (Finnish contributions). The project has been performed in four phases during 2006-2010. This guidance document aims at describing, on the basis of the work performed throughout the project, issues to consider when defining, applying and interpreting probabilistic safety criteria. Thus, the basic aim of the document is to serve as a checklist and toolbox for the definition and application of probabilistic safety criteria. The document describes the terminology and concepts involved, the levels of criteria and relations between these, how to define a probabilistic safety criterion, how to apply a probabilistic safety criterion, on what to apply the probabilistic safety criterion, and how to interpret the result of the application. The document specifically deals with what makes up a probabilistic safety criterion, i.e., the risk metric, the frequency criterion, the PSA used for assessing compliance and the application procedure for the criterion. It also discusses the concept of subsidiary criteria, i.e., different levels of safety goals. The results from the project can be used as a platform for discussions at the utilities on how to define and use quantitative safety goals. The results can also be used by safety authorities as a reference for risk-informed regulation. The outcome can have an impact on the requirements on PSA, e.g., regarding quality, scope, level of detail, and documentation. Finally, the results can be expected to support on-going activities concerning risk-informed applications. (Author)

The Reference Scenarios for the Swiss Emergency Planning

International Nuclear Information System (INIS)

Hanspeter Isaak; Navert, Stephan B.; Ralph Schulz

2006-01-01

For the purpose of emergency planning and preparedness, realistic reference scenarios and corresponding accident source terms have been defined on the basis of common plant features. Three types of representative reference scenarios encompass the accident sequences expected to be the most probable. Accident source terms are assumed to be identical for all Swiss nuclear power plants, although the plants differ in reactor type and power. Plant-specific probabilistic safety analyses were used to justify the reference scenarios and the postulated accident source terms. From the full spectrum of release categories available, those categories were selected which would be covered by the releases and time frames assumed in the reference scenarios. For each nuclear power plant, the cumulative frequency of accident sequences not covered by the reference scenarios was determined. It was found that the cumulative frequency for such accident sequences does not exceed about 1 x 10 -6 per year. The Swiss Federal Nuclear Safety Inspectorate concludes that the postulated accident source terms for the reference scenarios are consistent with the current international approach in emergency planning, where one should concentrate on the most probable accident sequences. (N.C.)

Environmental Change in Post-closure Safety Assessment of Solid Radioactive Waste Repositories. Report of Working Group 3 Reference Models for Waste Disposal of EMRAS II Topical Heading Reference Approaches for Human Dose Assessment. Environmental Modelling for Radiation Safety (EMRAS II) Programme

International Nuclear Information System (INIS)

2016-08-01

Environmental assessment models are used for evaluating the radiological impact of actual and potential releases of radionuclides to the environment. They are essential tools for use in the regulatory control of routine discharges to the environment and also in planning measures to be taken in the event of accidental releases. They are also used for predicting the impact of releases which may occur far into the future, for example, from underground radioactive waste repositories. It is important to verify, to the extent possible, the reliability of the predictions of such models by a comparison with measured values in the environment or with predictions of other models. The IAEA has been organizing programmes of international model testing since the 1980s. These programmes have contributed to a general improvement in models, in the transfer of data and in the capabilities of modellers in Member States. IAEA publications on this subject over the past three decades demonstrate the comprehensive nature of the programmes and record the associated advances which have been made. From 2009 to 2011, the IAEA organized a programme entitled Environmental Modelling for Radiation Safety (EMRAS II), which concentrated on the improvement of environmental transfer models and the development of reference approaches to estimate the radiological impacts on humans, as well as on flora and fauna, arising from radionuclides in the environment. Different aspects were addressed by nine working groups covering three themes: reference approaches for human dose assessment, reference approaches for biota dose assessment and approaches for assessing emergency situations. This publication describes the work of the Reference Models for Waste Disposal Working Group

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

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

1977-05-01

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

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

International Nuclear Information System (INIS)

2004-01-01

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

Project Guarantee 1985. Final repository for low- and intermediate-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

The safety barrier system for the type B repository for low- and intermediate-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). In the case of low- and intermediate-level waste the technical safety barrier system comprises: waste solidification matrix (cement, bitumen and resin), immobilisation of the waste packages in containers using liquid cement, concrete repository containers, backfilling of remaining vacant storage space with special concrete, concrete lining of the repository caverns, sealing of access tunnels on final closure of the repository. Natural geological safety barriers - host rock and overlying formations - have the following important functions. Because of its stability, the host rock in the repository zone protects the technical safety barrier system from destruction caused by climatic effects and erosion for a sufficient length of time. It also provides for low water flow and favourable chemistry (reducing conditions)

LFR safety approach and main ELFR safety analysis results

International Nuclear Information System (INIS)

Bubelis, E.; Schikorr, M.; Frogheri, M.; Mansani, L.; Bandini, G.; Burgazzi, L.; Mikityuk, K.; Zhang, Y.; Lo Frano, R.; Forgione, N.

2013-01-01

LFR safety approach: → A global safety approach for the LFR reference plant has been assessed and the safety analyses methodology has been developed. → LFR follows the general guidelines of the Generation IV safety concept recommendations. Thus, improved safety and higher reliability are recognized as an essential priority. → The fundamental safety objectives and the Defence-in-Depth (DiD) approach, as described by IAEA Safety Guides, have been preserved. → The recommendations of the Risk and Safety Working Group (RSWG) of GEN-IV IF has been taken into account: • safety is to be “built-in” in the fundamental design rather than “added on”; • full implementation of the Defence-in-Depth principles in a manner that is demonstrably exhaustive, progressive, tolerant, forgiving and well-balanced; • “risk-informed” approach - deterministic approach complemented with a probabilistic one; • adoption of an integrated methodology that can be used to evaluate and document the safety of Gen IV nuclear systems - ISAM. In particular the OPT tool is the fundamental methodology used throughout the design process

The application of the Appropriate Level of Protection (ALOP) and Food Safety Objective (FSO) concepts in food safety management, using Listeria monocytogenes in deli meats as a case study

NARCIS (Netherlands)

Gkogka, E.; Reij, M.W.; Gorris, L.G.M.; Zwietering, M.H.

2013-01-01

To establish a link between governmental food safety control and operational food safety management, the concepts of the Appropriate Level of Protection (ALOP) and the Food Safety Objective (FSO) have been suggested by international governmental bodies as a means for competent authorities to make

A Systematic Analysis of Functional Safety Certification Practices in Industrial Robot Software Development

Directory of Open Access Journals (Sweden)

Tong Xie

2017-01-01

Full Text Available For decades, industry robotics have delivered on the promise of speed, efficiency and productivity. The last several years have seen a sharp resurgence in the orders of industrial robots in China, and the areas addressed within industrial robotics has extended into safety-critical domains. However, safety standards have not yet been implemented widely in academia and engineering applications, particularly in robot software development. This paper presents a systematic analysis of functional safety certification practices in software development for the safety-critical software of industrial robots, to identify the safety certification practices used for the development of industrial robots in China and how these practices comply with the safety standard requirements. Reviewing from Chinese academic papers, our research shows that safety standards are barely used in software development of industrial robot. The majority of the papers propose various solutions to achieve safety, but only about two thirds of the papers refer to non-standardized approaches that mainly address the systematic level rather than the software development level. In addition, our research shows that with the development of artificial intelligent, an emerging field is still on the quest for standardized and suitable approaches to develop safety-critical software.

Considerations on Applying the Method for Assessing the Level of Safety at Work

Directory of Open Access Journals (Sweden)

Costica Bejinariu

2017-07-01

Full Text Available The application of the method for assessing the level of safety at work starts with a document that contains the cover page, the description of the company (name, location, core business, organizational chart etc., description of the work system, a detailed list of its components, and a brief description of the assessment method. It continues with a Microsoft Excel document, which represents the actual application of the method and, finally, there is another document presenting conclusions, proposals, and prioritizations, which leads to the execution of the Prevention and Protection Plan. The present paper approaches the issue of developing the Microsoft Excel document, an essential part of the method for assessing the level of safety at work. The document is divided into a variable number of worksheets, showing the risk categories of general, specific, and management.

C-Band Airport Surface Communications System Engineering-Initial High-Level Safety Risk Assessment and Mitigation

Science.gov (United States)

Zelkin, Natalie; Henriksen, Stephen

2011-01-01

This document is being provided as part of ITT's NASA Glenn Research Center Aerospace Communication Systems Technical Support (ACSTS) contract: "New ATM Requirements--Future Communications, C-Band and L-Band Communications Standard Development." ITT has completed a safety hazard analysis providing a preliminary safety assessment for the proposed C-band (5091- to 5150-MHz) airport surface communication system. The assessment was performed following the guidelines outlined in the Federal Aviation Administration Safety Risk Management Guidance for System Acquisitions document. The safety analysis did not identify any hazards with an unacceptable risk, though a number of hazards with a medium risk were documented. This effort represents an initial high-level safety hazard analysis and notes the triggers for risk reassessment. A detailed safety hazards analysis is recommended as a follow-on activity to assess particular components of the C-band communication system after the profile is finalized and system rollout timing is determined. A security risk assessment has been performed by NASA as a parallel activity. While safety analysis is concerned with a prevention of accidental errors and failures, the security threat analysis focuses on deliberate attacks. Both processes identify the events that affect operation of the system; and from a safety perspective the security threats may present safety risks.

Do we need an integrative approach to food safety at the country level?

International Nuclear Information System (INIS)

Ristic, G.

2002-01-01

Scientific data show increasing evidence of relationship between food safety and food standards on one hand and public health concern on the other hand. In FR Yugoslavia in 1989 the system of reporting on food safety issues on federal and republic level was established. The system provides data on laboratory analysis of 22 food items (bread, milk, meat and meat products, vegetables, processed vegetables etc). Those items were and still are tested on food quality and safety parameters such as microbiological, chemical and radio nuclides. Seldom all required testing on chemical and radio nuclides are performed, so we lack exact risk assessment for those contaminants. Further, during war conflict in FR Yugoslavia and also due to industrial hazards in neighbouring countries (Rumania, Hungary) high quantities of PCBs, dioxins, heavy metals, arsenic compounds and other toxic compounds contaminated the environment. In the soil and in some food products (animal fats predominantly) radionuclides originating from Chernobyl hazard can still be detected. In order to identify the level of exposure to chemical and radio nuclide contaminants in the food chain it is essential to test intensively and systematically food from animal and from plant origin. In order to prevent entering the contaminants to the food chain new recommendations from WHO, FAO and EU suggest implementation of integrative approach to food safety and control over the whole chain of food production from 'farm to table'. This approach provides control of the contaminants in soil, water, air, control over primary food production (covering animal feed too), intensive control over processing with implementation of HACCP system, but also, over transportation, retail trade, street food and home made food too. In our country creation of the map of the polluted areas, and actions in order to treat the pollution should accompany implementation of this new food safety system. The need for assessment of the level of

Do we need an integrative approach to food safety at the country level?

Energy Technology Data Exchange (ETDEWEB)

Ristic, G E-mail:risticg@eunet yu [Department of Nutrition, Medical Faculty, Belgrade (Yugoslavia)

2002-05-01

Scientific data show increasing evidence of relationship between food safety and food standards on one hand and public health concern on the other hand. In FR Yugoslavia in 1989 the system of reporting on food safety issues on federal and republic level was established. The system provides data on laboratory analysis of 22 food items (bread, milk, meat and meat products, vegetables, processed vegetables etc). Those items were and still are tested on food quality and safety parameters such as microbiological, chemical and radio nuclides. Seldom all required testing on chemical and radio nuclides are performed, so we lack exact risk assessment for those contaminants. Further, during war conflict in FR Yugoslavia and also due to industrial hazards in neighbouring countries (Rumania, Hungary) high quantities of PCBs, dioxins, heavy metals, arsenic compounds and other toxic compounds contaminated the environment. In the soil and in some food products (animal fats predominantly) radionuclides originating from Chernobyl hazard can still be detected. In order to identify the level of exposure to chemical and radio nuclide contaminants in the food chain it is essential to test intensively and systematically food from animal and from plant origin. In order to prevent entering the contaminants to the food chain new recommendations from WHO, FAO and EU suggest implementation of integrative approach to food safety and control over the whole chain of food production from 'farm to table'. This approach provides control of the contaminants in soil, water, air, control over primary food production (covering animal feed too), intensive control over processing with implementation of HACCP system, but also, over transportation, retail trade, street food and home made food too. In our country creation of the map of the polluted areas, and actions in order to treat the pollution should accompany implementation of this new food safety system. The need for assessment of the level of

Management of safety and safety culture at the NPPs of Ukraine

International Nuclear Information System (INIS)

Koltakov, Vladimir

2002-01-01

The report contains general aspects of safety and safety culture. The brief description of operational characteristics and basic indexes of atomic power plants at the Ukraine are represented. The information referring to structure of NPPs of Operation organization license-holder, safety responsibility of both Regulatory and Utility Bodies also is given. The main part of the report include seven sections: 1. Practical application of safety management models; 2. erspective on the relationship between safety management and safety culture; 3. The role of leadership in achieving high standards of safety; 4. Current and future challengers that impact on safety culture and safety management (e.g. the impact of competition, changing, economic and political circumstances, workforce demographics, etc.); 5. Key lessons learned from major events; 6. Practical applications of safety culture concepts (e.g. learning organizations, training staff communications, etc.); 7. dvance in human performance. Some of the main pending safety and safety culture problems that are necessary to achieve in the near future are mentioned

Project Guarantee 1985. Final repository for low- and intermediate level radioactive wastes: Safety report

International Nuclear Information System (INIS)

Anon.

1985-01-01

Storage of radioactive waste must delay the return of radionuclides to the biosphere for a long period of time and must maintain the release rates at a sufficiently low level for all time. This is achieved with the aid of a series of safety barriers which consist, on the one hand, of technical barriers in the repository and, on the other hand , of natural geological barriers as they occur at the repository location. In order to assess the efficiency of the barriers, the working methods of the technical barriers and the host rock must be understood. This understanding is transferred into quantitative models in order to calculate the safety of the repository. The individual barriers and the methods used to modelling their functions were described in volume NGB 85-07 of the Project Guarantee 1985 report series and the data necessary for modelling were given. The models and data are used in the safety analysis, the results of which are contained in the present report. Safety considerations show that models are available in Switzerland which allow, in principle, an assessment of the long-term behaviour of a repository for low- and intermediate-level waste. The evaluation of earlier studies and experimental work, suitable laboratory measurements and results from field research enable compilation of a representative data-set so that the requirements for quantitative statements on safety of final disposal are met from this side also. The safety calculations show that the radiation doses calculated for a base case scenario with realistic/conservative parameter values are negligibly low. Also, radiation doses which are clearly under the protection standard of 10 mrem per year result for conservative values and the cumulation of several conservative assumptions. Even assuming exposure of the repository by erosion, a radiotoxicity of the soil formed results which is under natural values

Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

International Nuclear Information System (INIS)

2014-01-01

The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

Rational quantitative safety goals: a summary

International Nuclear Information System (INIS)

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

1984-08-01

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

Development of a new methodology for quantifying nuclear safety culture

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2017-01-01

The present study developed a Safety Culture Impact Assessment Model (SCIAM) which consists of a safety culture assessment methodology and a safety culture impact quantification methodology. The SCIAM uses a safety culture impact index (SCII) to monitor the status of safety culture of NPPs periodically and it uses relative core damage frequency (RCDF) to present the impact of safety culture on the safety of NPPs. As a result of applying the SCIAM to the reference plant (Kori 3), the standard for the healthy safety culture of the reference plant is suggested. SCIAM might contribute to improve the safety of NPPs (Nuclear Power Plants) by monitoring the status of safety culture periodically and presenting the standard of healthy safety culture.

Development of a new methodology for quantifying nuclear safety culture

Energy Technology Data Exchange (ETDEWEB)

Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

2017-01-15

The present study developed a Safety Culture Impact Assessment Model (SCIAM) which consists of a safety culture assessment methodology and a safety culture impact quantification methodology. The SCIAM uses a safety culture impact index (SCII) to monitor the status of safety culture of NPPs periodically and it uses relative core damage frequency (RCDF) to present the impact of safety culture on the safety of NPPs. As a result of applying the SCIAM to the reference plant (Kori 3), the standard for the healthy safety culture of the reference plant is suggested. SCIAM might contribute to improve the safety of NPPs (Nuclear Power Plants) by monitoring the status of safety culture periodically and presenting the standard of healthy safety culture.

Safety and radiation protection aspects of the management of radioactive wastes of high level activity

International Nuclear Information System (INIS)

Candes, P.; Pradel, J.

1977-01-01

Appropriate consideration is given in France to safety and protection problems to be solved from the production up to the final disposal of radioactive wastes of high level activity. The first stage of the work consisted in emphasizing the various technical options. Different strategies appear to be possible, taking into account technical, political, and psychological difficulties. This results in evaluating the safety problems to be solved in the framework of those strategies. In this field, the main safety and protection principles do not differ from those which apply to other nuclear facilities. Nevertheless, duration is in most cases a quite different factor (thousands or millions of years). The question is then raised of evaluating the importance to be given to very remote consequences, both at philosophical and scientific levels. As a first result of those considerations, the application of the ''barrier'' concept is recommended. This concept is familiar to safety specialists. Different barriers for which particular problems are listed and evaluated, are defined. Another results with regard to radiation protection principles is to consider that if safety provisions should lead to a containment of radioactive products as efficient as possible, it would not be realistic to consider such a containment as absolute, in particular for disposal durations arising to thousands of years. It is therefore assumed that a limited radioactivity transfer should be taken into account, and its consequences for environment and man be calculated. This is especially true in the study of an appropriate site for final storage, and the study should necessarily include a detailed investigation of the retention characteristics of soil layers, and the implementation of appropriate models giving a sufficiently accurate evaluation of the consequences of transfers, including those related to the effect of various elements after their arrival into the biosphere. The authors review the

Site safety requirements for high level waste disposal

International Nuclear Information System (INIS)

Chen Weiming; Wang Ju

2006-01-01

This paper outlines the content, status and trend of site safety requirements of International Atomic Energy Agency, America, France, Sweden, Finland and Japan. Site safety requirements are usually represented as advantageous vis-a-vis disadvantagous conditions, and potential advantage vis-a-vis disadvantage conditions, respectively in aspects of geohydrology, geochemistry, lithology, climate and human intrusion etc. Study framework and steps of site safety requirements for China are discussed under the view of systems science. (authors)

Improvement of the safety level of installations with the generalization of procedures

International Nuclear Information System (INIS)

Cornille, Y.; Dupraz, B.; Schektman, N.

1986-06-01

The generalization of control procedures to the largest possible spectra of accidental situations which is being developed on pressurized water reactor units will allow to increase the safety level of these installations. This improvement has been quantified for some situations pointing out an appreciable mitigation of meltdown risk which could result. A new improvement is aimed with the definition and the utilization of new procedures ''by states'' which will allow an optimized treatment of situations resulting from multiple failures, now treated in the procedures SPI - SPU - U1. The needs related to these procedures and their development led to joint research and development programs between Electricite de France and the Institute of Protection and Nuclear Safety [fr

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

Directory of Open Access Journals (Sweden)

Kenneth Edgar Hernandez-Ruiz

2016-01-01

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

16 CFR 1511.8 - Metric references.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Metric references. 1511.8 Section 1511.8 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION FEDERAL HAZARDOUS SUBSTANCES ACT REGULATIONS... parentheses for convenience and information only. ...

Radiological safety assessment of a reference INTOR facility

International Nuclear Information System (INIS)

Khan, T.A.; Stasko, R.R.; Watts, R.T.; Shaw, G.; Morrison, C.A.; Russell, S.; Kempe, T.; Zimmerman, R.

1985-03-01

This report consists of a number of separate studies all of which were performed in support of INTOR Critical Issue D: Tritium Containment and Personnel Access vs Remote Maintenance. The common thread running through these studies is the radiological safety element in the design and operation of the INTOR facility. The intent is to help establish a firm basis for comparisons between a reactor cell maintenance option which requires personnel access, and one which involves completely remote maintenance

System safety education focused on flight safety

Science.gov (United States)

Holt, E.

1971-01-01

The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.

ESRS guidelines for software safety reviews. Reference document for the organization and conduct of Engineering Safety Review Services (ESRS) on software important to safety in nuclear power plants

International Nuclear Information System (INIS)

2000-01-01

The IAEA provides safety review services to assist Member States in the application of safety standards and, in particular, to evaluate and facilitate improvements in nuclear power plant safety performance. Complementary to the Operational Safety Review Team (OSART) and the International Regulatory Review Team (IRRT) services are the Engineering Safety Review Services (ESRS), which include reviews of siting, external events and structural safety, design safety, fire safety, ageing management and software safety. Software is of increasing importance to safety in nuclear power plants as the use of computer based equipment and systems, controlled by software, is increasing in new and older plants. Computer based devices are used in both safety related applications (such as process control and monitoring) and safety critical applications (such as reactor protection). Their dependability can only be ensured if a systematic, fully documented and reviewable engineering process is used. The ESRS on software safety are designed to assist a nuclear power plant or a regulatory body of a Member State in the review of documentation relating to the development, application and safety assessment of software embedded in computer based systems important to safety in nuclear power plants. The software safety reviews can be tailored to the specific needs of the requesting organization. Examples of such reviews are: project planning reviews, reviews of specific issues and reviews prior final acceptance. This report gives information on the possible scope of ESRS software safety reviews and guidance on the organization and conduct of the reviews. It is aimed at Member States considering these reviews and IAEA staff and external experts performing the reviews. The ESRS software safety reviews evaluate the degree to which software documents show that the development process and the final product conform to international standards, guidelines and current practices. Recommendations are

Evolution of diagnostic reference levels in Spanish intraoral radiology

International Nuclear Information System (INIS)

Alcaraz, M.; Velasco, F.; Martinez-Beneyto, Y.; Alcaraz-Saura, M.; Velasco, E.; Achel, G. D.; Canteras, M.

2008-01-01

A total of 16 175 official reports of quality assurance on dental radiodiagnostic surgeries from 16 Spanish autonomous regions compiled during 2002-09 were studied to determine the evolution of diagnostic reference levels (DRLs) for obtaining a diagnostic image in the normal conditions of clinical practice in Spanish dental clinics. A DRL of 3.1 mGy was set in 2009, which represents a 35.4 % decrease compared with the dose determined in 2002 (4.8 mGy). During the same period, the mean dose fell by only 17.2 %. The DRL recommended by the European Union in 2004 for intraoral radiology is 4 mGy, and this study shows that 83.4 % of the installations used a dose below this. Of the installations using indirect or direct digital systems 1.1 and 1.2 %, respectively, used doses higher than those recommended, while 14.2 % of those using radiographic film exceeded this limit. (authors)

Provincial practice: adopting the new reference levels for Radon

International Nuclear Information System (INIS)

Kennedy, Christine; Johnson, Darryl

2008-01-01

Full text: In June 2007 new reference levels for radon gas were announced from Health Canada. These new levels brought new attention to the issue of radon gas exposure in the Province of Newfoundland and Labrador. Research in radon gas exposure has a long history in the fluorspar mines of the Burin Peninsula, indeed, occupational data from the 1950 's and 60 's had been included in Darby et al. large scale occupational meta analyses. Radon was also implicated in a royal commission of occupational mining hazards in Newfoundland in 1968. Although, the occupational exposures of miners have been well documented, very little is known about population exposures in indoor spaces. Geological maps are currently being composed for national areas, but data for this province is not yet published. Information about mining tailings used in construction materials or as fill is very poor (unlike the data known for uranium mining tailings in Ontario and Manitoba). The challenges of estimating radon exposures for the province are myriad and these are explored here in this narrative study of how new information has to be generated, and then incorporated into new environmental public health policies for a population. The process by which new scientific information informs public health policy is described in this study of radon in Newfoundland and Labrador. Anticipated regulations for new buildings are discussed. (author)

Nuclear Safety. 1997; Surete Nucleaire. 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-19

A quick review of the nuclear safety at EDF may be summarized as follows: - the nuclear safety at EDF maintains at a rather good standard; - none of the incidents that took place has had any direct impact upon safety; - the availability remained good; - initiation of the floor 4 reactor generation (N4 unit - 1450 MW) ensued without major difficulties (the Civaux 1 NPP has been coupled to the power network at 24 december 1997); - the analysis of the incidents interesting from the safety point of view presents many similarities with earlier ones. Significant progress has been recorded in promoting actively and directly a safe operation by making visible, evident and concrete the exertion of the nuclear operation responsibility and its control by the hierarchy. The report develops the following chapters and subjects: 1. An overview on 1997; 1.1. The technical issues of the nuclear sector; 1.2. General performances in safety; 1.3. The main incidents; 1.4. Wastes and radiation protection; 2. Nuclear safety management; 2.1. Dynamics and results; 2.2. Ameliorations to be consolidated; 3. Other important issues in safety; 3.1. Probabilistic safety studies; 3.2. Approach for safety re-evaluation; 3.3. The network safety; 3.4. Crisis management; 3.5. The Lifetime program; 3.6. PWR; 3.7. Documentation; 3.8. Competence; 4. Safety management in the future; 4.1. An open future; 4.2. The fast neutron NPP at Creys-Malville; 4.3. Stabilization of the PWR reference frame; 4.4. Implementing the EURATOM directive regarding the radiation protection standards; 4.5. Development of biomedical research and epidemiological studies; 4.6. New regulations concerning the liquid and gaseous effluents; 5. Visions of an open future; 5.1. Alternative views upon safety ay EDF; 5.2. Safety authority; 5.3. International considerations; 5.4. What happens abroad; 5.5. References from non-nuclear domain. Four appendices are added referring to policy of safety management, policy of human factors in NPPs

Evidence of dose saving in routine CT practice using iterative reconstruction derived from a national diagnostic reference level survey.

Science.gov (United States)

Thomas, P; Hayton, A; Beveridge, T; Marks, P; Wallace, A

2015-09-01

To assess the influence and significance of the use of iterative reconstruction (IR) algorithms on patient dose in CT in Australia. We examined survey data submitted to the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) National Diagnostic Reference Level Service (NDRLS) during 2013 and 2014. We compared median survey dose metrics with categorization by scan region and use of IR. The use of IR results in a reduction in volume CT dose index of between 17% and 44% and a reduction in dose-length product of between 14% and 34% depending on the specific scan region. The reduction was highly significant (p sum test) for all six scan regions included in the NDRLS. Overall, 69% (806/1167) of surveys included in the analysis used IR. The use of IR in CT is achieving dose savings of 20-30% in routine practice in Australia. IR appears to be widely used by participants in the ARPANSA NDRLS with approximately 70% of surveys submitted employing this technique. This study examines the impact of the use of IR on patient dose in CT on a national scale.

Progress report on safety research of high-level waste management for the period April 1986 to March 1987

International Nuclear Information System (INIS)

Nakamura, Haruto; Tashiro, Shingo

1987-08-01

Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1986 are reviewed in the report. Topics in the three sections are as follows: 1) Non-radioactive research has been continued on Synroc irradiation and modellings of waste form leaching. 2) Research results are described in the section of Safety Evaluation for Geological Disposal on engineered barriers, field tests, safety assessment models, migration, natural analogue, seabed disposal and conceptual design of a repository. 3) Adsorption behaviour of plutonium on leach-containers and migration of leached cesium in a rock column are described in the section of Safety Examination of Vitrified Forms in the Hot Cells of WASTEF. (author)

The safety case in support of the license application of the surface repository of low-level waste in Dessel, Belgium

International Nuclear Information System (INIS)

Wacquier, William; Cool, Wim

2014-01-01

The modern concept of the safety case, developed by the OECD/NEA for geological repositories of high- and medium-level waste has been successfully applied by ONDRAF/ NIRAS for a surface repository for Category A waste (i.e. low-level waste) in Belgium in the current project phase 2006-2012. This resulted in the submission on 31 January 2013 by ONDRAF/NIRAS of an application for a 'construction and operation license' to the safety authorities. The benefits of using the notion of the safety case have been that: i) safety has been incorporated in an integrated manner within all assessment basis, design and safety assessment activities; ii) the process of development of the license application has gained in clarity and traceability; iii) the documentation of the license application contains multiple lines of argumentation for safety rather than argumentation based only on quantitative radiological impact calculations. To offer a comprehensive view on the safety argumentation and its development, it has been found useful to develop the argumentation not only along a safety statements structure but also along the safety report structure. (authors)

Safety of nuclear fuel cycle facilities. Safety requirements

International Nuclear Information System (INIS)

2008-01-01

This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific reference include aspects of nuclear fuel generation, storage, reprocessing and disposal. Contents: 1. Introduction; 2. The safety objective, concepts and safety principles; 3. Legal framework and regulatory supervision; 4. The management system and verification of safety; 5. Siting of the facility; 6. Design of the facility; 7. Construction of the facility; 8. Commissioning of the facility; 9. Operation of the facility; 10. Decommissioning of the facility; Appendix I: Requirements specific to uranium fuel fabrication facilities; Appendix II: Requirements specific to mixed oxide fuel fabrication facilities; Appendix III: Requirements specific to conversion facilities and enrichment facilities

A prediction model for the radiation safety management behavior of medical cyclotrons

Energy Technology Data Exchange (ETDEWEB)

Jung, Ji Hye; Han, Eun Ok [Daegu Health College, Daegu (Korea, Republic of); Kim, Ssang Tae [CareCamp Inc., Seoul (Korea, Republic of)

2008-06-15

This study attempted to provide reference materials for improving the behavior level in radiation safety managements by drawing a prediction model that affects the radiation safety management behavior because the radiation safety management of medical Cyclotrons, which can be used to produce radioisotopes, is an important factor that protects radiation caused diseases not only for radiological operators but average users. In addition, this study obtained follows results through the investigation applied from January 2 to January 30, 2008 for the radiation safety managers employed in 24 authorized organizations, which have already installed Cyclotrons, through applying a specific form of questionnaire in which the validity was guaranteed by reference study, site investigation, and focus discussion by related experts. The radiation safety management were configured as seven steps: step 1 is a production preparation step, step 2 is an RI production step, step 3 is a synthesis step, step 4 is a distribution step, step 5 is a quality control step, step 6 is a carriage container packing step, and step 7 is a transportation step. It was recognized that the distribution step was the most exposed as 15 subjects (62.5%), the items of 'the sanction and permission related works' and 'the guarantee of installation facilities and production equipment' were the most difficult as 9 subjects (37.5%), and in the trouble steps in such exposure, the item of 'the synthesis and distribution' steps were 4 times, respectively (30.8%). In the score of the behavior level in radiation safety managements, the minimum and maximum scores were 2.42 and 4.00, respectively, and the average score was 3.46 {+-} 0.47 out of 4. Prosperity and well-being programs in the behavior and job in radiation safety managements (r=0.529) represented a significant correlation statistically. In the drawing of a prediction model based on the factors that affected the behavior in

The storage of liquid high level waste at BNFL, Sellafield. Addendum to February 2000 report

International Nuclear Information System (INIS)

2001-08-01

On 18 February 2000 the Health and Safety Executive (HSE) published a report on the work of its Nuclear Installations Inspectorate (NIl) in regulating the storage of liquid high level waste at the BNFL Sellafield site. Within the report NIl gave two undertakings. One was to publish an addendum around 1 year later covering its assessment of the new safety case for the storage plant and the second was to publish a further addendum when progress had been made with options studies for reducing the stocks of liquid high level waste (HLW), also referred to as highly active liquor (HAL), to a buffer level. A progress report was published in February 2001 which included a summary of the assessment of the new safety case and NIl's regulatory action to enforce liquid HLW stock reductions. This addendum provides a more detailed update on the position reached based on consideration of BNFL's responses to the recommendations from the February 2000 HLW report since its publication. It embodies the two addenda referred to above integrated into a single document for publication

Bohunice Nuclear Power Plant Safety Upgrading Program

International Nuclear Information System (INIS)

Toth, A.; Fagula, L.

1996-01-01

Bohunice nuclear Power Plant generation represents almost 50% of the Slovak republic electric power production. Due to such high level of commitment to nuclear power in the power generation system, a special attention is given to safe and reliable operation of NPPs. Safety upgrading and operational reliability improvement of Bohunice V-1 NPP was carried out by the Bohunice staff continuously since the plant commissioning. In the 1990 - 1993 period extensive projects were realised. As a result of 'Small Reconstruction of the Bohunice V-1 NPP', the standards of both the nuclear safety and operational reliability have been significantly improved. The implementation of another modifications that will take place gradually during extended refuelling outages and overhauls in the course of 1996 through 1999, is referred to as the Gradual Reconstruction of the Bohunice V-1 Plant. The general goal of the V-1 NPP safety upgrading is the achievement of internationally acceptable level of nuclear safety. Extensive and financially demanding modification process of Bohunice V-2 NPP is likely to be implemented after a completion of the Gradual Reconstruction of the Bohunice V-1 NPP, since the year 1999. With this in mind, a first draft of the strategy of the Bohunice V-2 NPP upgrading program based on Probabilistic Safety assessment consideration was developed. A number of actions with a general effect on Bohunice site safety is evident. All these activities are aimed at reaching the essential objective of Bohunice NPP Management - to ensure a safe, reliable and effective electric energy and heat generation at the Bohunice site. (author)

On safety management and nuclear safety - A frame of reference for studies of safety management with examples from non-nuclear contects of relevance for nuclear safety

International Nuclear Information System (INIS)

Svenson, O.; Allwin, P.; Salo, I.

2004-03-01

The report includes three case studies of safety management. The studies are presented as chapters, but are written in a format that makes them easy to read separately. Two of the studies cover regulators (the Swedish Civil Aviation Safety Authority, Luftfartsinspektionen) and the Norwegian Petroleum Directorate) and one a regulated activity/industry (a car manufacturer, Volvo Car). The introduction outlines a living system framework and relates this to concepts used in organizational management. The report concludes with some findings with potential relevance for safety management in the nuclear power domain. In the next phase of the work, the regulated counterparts of the regulators here will be investigated in addition to a fourth case study of a regulated activity/industry. (au)

Monitoring road safety development at regional level: A case study in the ASEAN region.

Science.gov (United States)

Chen, Faan; Wang, Jianjun; Wu, Jiaorong; Chen, Xiaohong; Zegras, P Christopher

2017-09-01

Persistent monitoring of progress, evaluating the results of interventions and recalibrating to achieve continuous improvement over time is widely recognized as being crucial towards the successful development of road safety. In the ASEAN (Association of Southeast Asian Nations) region there is a lack of well-resourced teams that contain multidisciplinary safety professionals, and specialists in individual countries, who are able to carry out this work effectively. In this context, not only must the monitoring framework be effective, it must also be easy to use and adapt. This paper provides a case study that can be easily reproduced; based on an updated and refined Road Safety Development Index (RSDI), by means of the RSR (Rank-sum ratio)-based model, for monitoring/reporting road safety development at regional level. The case study was focused on the road safety achievements in eleven Southeast Asian countries; identifying the areas of poor performance, potential problems and delays. These countries are finally grouped into several classes based on an overview of their progress and achievements regarding to road safety. The results allow the policymakers to better understand their own road safety progress toward their desired impact; more importantly, these results enable necessary interventions to be made in a quick and timely manner. Keeping action plans on schedule if things are not progressing as desired. This would avoid 'reinventing the wheel' and trial and error approaches to road safety, making the implementation of action plans more effective. Copyright © 2017 Elsevier Ltd. All rights reserved.

Endogenous steroid hormone levels in early pregnancy and risk of testicular cancer in the offspring: a nested case-referent study.

Science.gov (United States)

Holl, Katsiaryna; Lundin, Eva; Surcel, Heljä-Marja; Grankvist, Kjell; Koskela, Pentti; Dillner, Joakim; Hallmans, Göran; Wadell, Göran; Olafsdottir, Gudridur H; Ogmundsdottir, Helga M; Pukkala, Eero; Lehtinen, Matti; Stattin, Pär; Lukanova, Annekatrin

2009-06-15

According to the leading hypothesis on testicular cancer (TC) etiology exposure to a specific pattern of steroid hormones in utero, in particular, to high levels of estrogens and low levels of androgens is the major determinant of TC risk in the offspring. We performed a case-referent study nested within Finnish, Swedish and Icelandic maternity cohorts exploiting early pregnancy serum samples to evaluate the role of maternal endogenous steroid hormones with regard to the risk of TC. TC cases and referents were aged between 0 and 25 years. For each case-index mother pair, three or four matched referent-referent mother pairs were identified using national population registries. First trimester or early second trimester sera were retrieved from the index mothers of 73 TC cases and 286 matched referent mothers, and were tested for dehydroepiandrosterone sulfate (DHEAS), androstenedione, testosterone, estradiol, estrone, and sex hormone binding globulin (SHBG). Offspring of mothers with high DHEAS levels had a significantly decreased risk of TC (OR for highest vs. lowest DHEAS quartile, 0.18 (95% CI 0.06-0.58). In contrast, offspring of mothers with high androstenedione levels had an increased risk of TC (OR 4.1; 95% CI 1.2-12.0). High maternal total estradiol level also tended to be associated with an increased risk of TC in the offspring (OR 32; 95% CI 0.98-1,090). We report the first direct evidence that interplay of maternal steroid hormones in the early pregnancy is important in the etiology of TC in the offspring. Copyright 2008 UICC.

Countermeasures that work : a highway safety countermeasure guide for state highway safety offices : eighth edition : 2015

Science.gov (United States)

2015-11-01

The guide is a basic reference to assist State Highway Safety Offices in selecting effective, evidence- based : countermeasures for traffic safety problem areas. These areas include: : - Alcohol-and Drug-Impaired Driving; : - Seat Belts and Child Res...

Safety culture: modern slogan or effective contribution to safety?

International Nuclear Information System (INIS)

Salm, M.

1994-01-01

Safety culture is defined and its impact on nuclear power plants is documented using the words of the INSAG of IAEA. Two examples from the field of aviation and space flight testify, that the upper management, by its sheer image, may considerably influence actions of the lower levels of the hierarchy. Management therefore can do a lot more for safety than is commonly assumed. Two examples, although separated by 57 years, show that the mentioned influence remains unchanged inspire of progress in management- and organisation-methods as well as in safety-engineering. Safety culture is an overriding element of safety, acting at all levels of a hierarchy. Its action is most important on those levels, for which precise reglementation is hardly possible. The chain of technical and organisational measures guarantees safety only under the condition, that it is embedded in 'safety culture'. Safety culture therefore merits our full attention. (author) 1 fig

A relational leadership perspective on unit-level safety climate.

Science.gov (United States)

Thompson, Debra N; Hoffman, Leslie A; Sereika, Susan M; Lorenz, Holly L; Wolf, Gail A; Burns, Helen K; Minnier, Tamra E; Ramanujam, Rangaraj

2011-11-01

This study compared nursing staff perceptions of safety climate in clinical units characterized by high and low ratings of leader-member exchange (LMX) and explored characteristics that might account for differences. Frontline nursing leaders' actions are critical to ensure patient safety. Specific leadership behaviors to achieve this goal are underexamined. The LMX perspective has shown promise in nonhealthcare settings as a means to explain safety climate perceptions. Cross-sectional survey of staff (n = 711) and unit directors from 34 inpatient units in an academic medical center was conducted. Significant differences were found between high and low LMX scoring units on supervisor safety expectations, organizational learning-continuous improvement, total communication, feedback and communication about errors, and nonpunitive response to errors. The LMX perspective can be used to identify differences in perceptions of safety climate among nursing staff. Future studies are needed to identify strategies to improve staff safety attitudes and behaviors. Copyright © 2011 Wolters Kluwer Health | Lippincott Williams & Wilkins

Evaluation of the uncertainty around the mean level of 137Cs fallout at undisturbed reference site: A simple statistical approach

International Nuclear Information System (INIS)

Mabit, L.; Gonsalves, B.C.; Chen, X.; Toloza, A.; Weltin, G.; Darby, I.G.; Padilla-Alvarez, R.

2015-01-01

One of the major issues related to the use of 137 Cs as a soil erosion/sedimentation tracer is the selection of the reference site which is used to estimate the initial 137 Cs fallout input (also termed reference inventory). The initial 137 Cs fallout input is a key component of the conversion models used to estimate erosion and sedimentation rates from the 137 Cs data set. The selection and evaluation of the validity of reference sites have been explained in detail in the recent IAEA TECDOC 1741 ‘‘Guidelines for using Fallout radionuclides to assess erosion and effectiveness of soil conservation strategies’’. An investigation was carried out at the experimental research station of the Austrian Agency for Health and Food Safety (AGES), in Austria, Grabenegg (48°07'40 N , 15°13'16 E ). Located at an altitude of 260 m a.s.l, with an annual average temperature of 8.4 °C and annual precipitation of 686 mm, the soil of this area has been classified as Gleyic Cambisol with a silt-loamy texture

French nuclear safety authorities: for a harmonization of nuclear safety at the European level

International Nuclear Information System (INIS)

Anon.

2004-01-01

The European Commission is working on 2 directives concerning nuclear energy: the first one is dedicated to nuclear safety and the second to the management of radioactive wastes and spent fuels. In the context of the widening of the European Union and of the inter-connection of the different electric power grids throughout Europe, the harmonization of the rules in the nuclear safety field is seen by manufacturers as a mean to achieve a fair competition between nuclear equipment supplying companies and by the French nuclear safety authorities (FNSA) as a mean to keep on improving nuclear safety and to be sure that competitiveness does not drive safety standards down. According to FNSA the 2 European directives could give a legal framework to the harmonization and should contain principles that reinforce the responsibility of each state. FNSA considers that the EPR (European pressurized water reactor) may be an efficient tool for the harmonization because of existing industrial cooperation programs between France and Germany and between France and Finland. (A.C.)

Safety research program of NUCEF

International Nuclear Information System (INIS)

Naito, Y.

1996-01-01

To contribute the safety and establishment of advanced technologies in the area of nuclear fuel cycle, Japan Atomic Energy Research Institute (JAERI) has constructed a new research facility NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) as the center for the research and development, particularly on the reprocessing technology and transuranium (TRU) waste management. NUCEF consist of three buildings, administration building, experiment building A and B. Building A has two experiment facilities STACY (Static Experiment Critical Facility) and TRACY (Transient Experiment Critical Facility). The experiment building B is referred to as BECKY (Back-end Fuel Cycle Key Elements Research Facility). Researches on the reprocessing and the waste management are carried out with spent fuels, high-level liquid waste, TRU etc. in the α γ cell and glove boxes. NUCEF was constructed with the following aims. Using STACY and TRACY, are aimed, (1) research on advanced technology for criticality safety control, (2) reconfirmation of criticality safety margin of the Rokkasho reprocessing plant. Using BECKY, are aimed, (1) research on advanced technology of reprocessing process, (2) contribution to develop the scenario for TRU waste disposal, (3) development of new technology for TRU partitioning and volume reduction of radioactive waste. To realize the above aims, following 5 research subjects are settled in NUCEF, (1) Criticality safety research, (2) Research on safety and advanced technology of fuel reprocessing, (3) Research on TRU waste management, (4) Fundamental research on TRU chemistry, (5) Key technology development for TRU processing. (author)

The safety relief valve handbook design and use of process safety valves to ASME and International codes and standards

CERN Document Server

Hellemans, Marc

2009-01-01

The Safety Valve Handbook is a professional reference for design, process, instrumentation, plant and maintenance engineers who work with fluid flow and transportation systems in the process industries, which covers the chemical, oil and gas, water, paper and pulp, food and bio products and energy sectors. It meets the need of engineers who have responsibilities for specifying, installing, inspecting or maintaining safety valves and flow control systems. It will also be an important reference for process safety and loss prevention engineers, environmental engineers, and plant and process designers who need to understand the operation of safety valves in a wider equipment or plant design context. . No other publication is dedicated to safety valves or to the extensive codes and standards that govern their installation and use. A single source means users save time in searching for specific information about safety valves. . The Safety Valve Handbook contains all of the vital technical and standards informat...

Management of safety, safety culture and self assessment

International Nuclear Information System (INIS)

Carnino, A.

2000-01-01

Safety management is the term used for the measures required to ensure that an acceptable level of safety is maintained throughout the life of an installation, including decommissioning. The safety culture concept and its implementation are described in part one of the paper. The principles of safety are now quite well known and are implemented worldwide. It leads to a situation where harmonization is being achieved as indicated by the entry into force of the Convention on Nuclear Safety. To go beyond the present nuclear safety levels, management of safety and safety culture will be the means for achieving progress. Recent events which took place in major nuclear power countries have shown the importance of the management and the consequences on safety. At the same time, electricity deregulation is coming and will impact on safety through reductions in staffing and in operation and maintenance cost at nuclear installations. Management of safety as well as its control and monitoring by the safety authorities become a key to the future of nuclear energy.(author)

Comprehensive development plans for the low- and intermediate-level radioactive waste disposal facility in Korea and preliminary safety assessment

Energy Technology Data Exchange (ETDEWEB)

Jung, Kang Il; Kim, Jin Hyeong; Kwon, Mi Jin; Jeong, Mi Seon; Hong, Sung Wook; Park, Jin Beak [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

The disposal facility in Gyeongju is planning to dispose of 800,000 packages of low- and intermediate- level radioactive waste. This facility will be developed as a complex disposal facility that has various types of disposal facilities and accompanying management. In this study, based on the comprehensive development plan of the disposal facility, a preliminary post-closure safety assessment is performed to predict the phase development of the total capacity for the 800,000 packages to be disposed of at the site. The results for each scenario meet the performance target of the disposal facility. The assessment revealed that there is a significant impact of the inventory of intermediate-level radionuclide waste on the safety evaluation. Due to this finding, we introduce a disposal limit value for intermediate-level radioactive waste. With stepwise development of safety case, this development plan will increase the safety of disposal facilities by reducing uncertainties within the future development of the underground silo disposal facilities.

Probabilistic safety assessment for a generic deep geological repository for high-level waste and long-lived intermediate-level waste in clay

International Nuclear Information System (INIS)

Resele, G.; Holocher, J.; Mayer, G.; Hubschwerlen, N.; Niemeyer, M.; Beushausen, M.; Wollrath, J.

2010-01-01

Document available in extended abstract form only. In the selection procedure for the search of a final site location for the disposal of radioactive wastes, the comparison and evaluation of different potentially suitable repository systems in different types of host rocks will be an essential and crucial step. Since internationally accepted guidelines on how to perform such quantitative comparisons between repository systems with regard to their long-term safety behaviour are still lacking, in 2007 the German Federal Office for Radiation Protection launched the project 'VerSi' (Vergleichende Sicherheitsanalysen - Comparing Safety Assessments) that aims at the development of a methodology for the comparison of long-term safety assessments. A vital part of the VerSi project is the performance of long-term safety assessments for the comparison of two repository systems. The comparison focuses on a future repository for heat-generating, i.e. high-level and long-lived intermediate-level radioactive wastes in Germany. Rock salt is considered as a potential host rock for such a repository, and one repository system in VerSi is defined similarly to the potential site located in the Gorleben salt dome. Another suitable host rock formation may be clay. A generic location within the lower Cretaceous clays in Northern Germany is therefore chosen for the comparison of safety assessments within the VerSi project. The long-term safety assessment of a repository system for heat-generating radioactive waste at the generic clay location comprises different steps, amongst others: - Identifying the relevant processes in the near-field, in the geosphere and in the biosphere which are relevant for the long-term safety behaviour. - Development of a safety concept for the repository system. - Deduction of scenarios of the long-term evolution of the repository system. - Definition of statistic weights, i. e. the likelihood of occurrence of the scenarios. - Performance of a

Formulation and use of reliability data for probabilistic safety assessment. Working material

International Nuclear Information System (INIS)

1991-01-01

In this case study the PSA had examined the Shoreham Nuclear Power Station, owned by Long Island Lightning Company, the PSA sponsoring Utility on the North Shore of Long Island, New York. The case study references other PSA literature on peer review which helps put this case study in perspective. One such reference, An Intensive Peer Review for PRA in fact contains a fully-documented case study of a peer review for a level 3 PSA. The reader is encouraged to refer to this Nuclear Safety Analysis Center report, available through the Electric Power Research Institute as the NSAC-67 report. 20 refs, 17 figs

Formulation and use of reliability data for probabilistic safety assessment. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-12-31

In this case study the PSA had examined the Shoreham Nuclear Power Station, owned by Long Island Lightning Company, the PSA sponsoring Utility on the North Shore of Long Island, New York. The case study references other PSA literature on peer review which helps put this case study in perspective. One such reference, An Intensive Peer Review for PRA in fact contains a fully-documented case study of a peer review for a level 3 PSA. The reader is encouraged to refer to this Nuclear Safety Analysis Center report, available through the Electric Power Research Institute as the NSAC-67 report. 20 refs, 17 figs.

Safety Study of Doxicycline in Broiler Chickens

Directory of Open Access Journals (Sweden)

Romeo Teodor Cristina

2010-05-01

Full Text Available Doxicycline is a structural isomer who meets the European reference standard (Ph.Eur. III 1997:0272 and respect EU Regulations for M.R.L. The good solubility permits high tissular absorption and distribution after administration in drinking water. The aim was to ascertain the clinical effects and tolerance consecutive to risen doses administration, as a part of safety study for this antibiotic in poultry in the respect of current drug testing methodology. The safety study revealed: good local and general tolerance to therapeutic doses (10 mg x kgb.w.-1 and to x2 dose, diarrhoea in lots which received x3 and x5 the therapeutic dose. Comparatively with the Control lot, haemoleucogram doesn’t suffer evident changes the registered values being between the references limits (exception of leucocitary formula. Also it was found an increased creatinin concentration consecutively to x3 and x5 greater doses administration and marked increasing of ASAT level and limited for ALAT’s, comparatively to control lot in the case of E4 lot, but light to reference values. Macro and microscopy revealed for the liver samples: hepatomegaly, diffuse hepatic degenerescence, nuclear heterochromatinization, vacuolar degenerescence; renal changes in nefrons and renal corpuscles, light splenomegaly, caecal sacs distension and brown-yellowish gaseous content to lots E3 and E4.

Nuclear ships and their safety

Energy Technology Data Exchange (ETDEWEB)

NONE

1961-04-15

Several aspects of nuclear ship propulsion, with special reference to nuclear safety, were discussed at an international symposium at Taormina, Italy, from 14-18 November 1960. Discussions on specific topics are conducted, grouped under the following headings: Economics and National Activities in Nuclear Ship Propulsion; International Problems and General Aspects of Safety for Nuclear Ships; Nuclear Ship Projects from the Angle of Safety; Ship Reactor Problems; Sea Motion and Hull Problems; Maintenance and Refuelling Problems; and Safety Aspects of Nuclear Ship Operation.

16 CFR Figure 3 to Part 1203 - Location of Reference Plane

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Location of Reference Plane 3 Figure 3 to Part 1203 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION CONSUMER PRODUCT SAFETY ACT... Plane ER10MR98.003 ...

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

International Nuclear Information System (INIS)

2008-01-01

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

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

International Nuclear Information System (INIS)

2004-01-01

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

The Research on the Impact of Management Level's Charismatic Leadership Style on Miners' Unsafe Behavior.

Science.gov (United States)

Li, Hongxia; Di, Hongxi; Tian, Shuicheng; Li, Jian

2015-01-01

The aim of this study is research the impact of management level's charismatic leadership style on miners' unsafe behavior by using the questionnaires on charismatic leadership style, safety attitude and the miners' unsafe behavior measurement to investigate 200 employees in Shen Dong Company. The research results suggest that management level's charismatic leadership style have very important influence on miners' unsafe behavior and the influence is affected by the safety attitude which is the intermediary function. In the end, this study propose advice on how to improve the coal mine enterprise managers charismatic leadership style in the coal mine enterprise's safety management work, including attach great importance to a variety of incentive methods, set up safety moral models, practice of inductive leadership concept, create a good atmosphere of safety, etc for reference for coal mining enterprises.

Features, events, processes, and safety factor analysis applied to a near-surface low-level radioactive waste disposal facility

Energy Technology Data Exchange (ETDEWEB)

Stephens, M.E.; Dolinar, G.M.; Lange, B.A. [Atomic Energy of Canada Limited, Ontario (Canada)] [and others

1995-12-31

An analysis of features, events, processes (FEPs) and other safety factors was applied to AECL`s proposed IRUS (Intrusion Resistant Underground Structure) near-surface LLRW disposal facility. The FEP analysis process which had been developed for and applied to high-level and transuranic disposal concepts was adapted for application to a low-level facility for which significant efforts in developing a safety case had already been made. The starting point for this process was a series of meetings of the project team to identify and briefly describe FEPs or safety factors which they thought should be considered. At this early stage participants were specifically asked not to screen ideas. This initial list was supplemented by selecting FEPs documented in other programs and comments received from an initial regulatory review. The entire list was then sorted by topic and common issues were grouped, and issues were classified in three priority categories and assigned to individuals for resolution. In this paper, the issue identification and resolution process will be described, from the initial description of an issue to its resolution and inclusion in the various levels of the safety case documentation.

Nuclear criticality safety: 3-day training course

International Nuclear Information System (INIS)

Schlesser, J.A.

1993-06-01

The open-quotes 3-Day Training Courseclose quotes is an intensive course in criticality safety consisting of lectures and laboratory sessions, including active student participation in actual critical experiments, a visit to a plutonium processing facility, and in-depth discussions on safety philosophy. The program is directed toward personnel who currently have criticality safety responsibilities in the capacity of supervisory staff and/or line management. This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course's primary instructor. It should be noted that when chapters were extracted, an attempt was made to maintain footnotes and references as originally written. Photographs and illustrations are numbered sequentially

Patient safety principles in family medicine residency accreditation standards and curriculum objectives

Science.gov (United States)

Kassam, Aliya; Sharma, Nishan; Harvie, Margot; O’Beirne, Maeve; Topps, Maureen

2016-01-01

Abstract Objective To conduct a thematic analysis of the College of Family Physicians of Canada’s (CFPC’s) Red Book accreditation standards and the Triple C Competency-based Curriculum objectives with respect to patient safety principles. Design Thematic content analysis of the CFPC’s Red Book accreditation standards and the Triple C curriculum. Setting Canada. Main outcome measures Coding frequency of the patient safety principles (ie, patient engagement; respectful, transparent relationships; complex systems; a just and trusting culture; responsibility and accountability for actions; and continuous learning and improvement) found in the analyzed CFPC documents. Results Within the analyzed CFPC documents, the most commonly found patient safety principle was patient engagement (n = 51 coding references); the least commonly found patient safety principles were a just and trusting culture (n = 5 coding references) and complex systems (n = 5 coding references). Other patient safety principles that were uncommon included responsibility and accountability for actions (n = 7 coding references) and continuous learning and improvement (n = 12 coding references). Conclusion Explicit inclusion of patient safety content such as the use of patient safety principles is needed for residency training programs across Canada to ensure the full spectrum of care is addressed, from community-based care to acute hospital-based care. This will ensure a patient safety culture can be cultivated from residency and sustained into primary care practice. PMID:27965349

AGI Safety Literature Review

OpenAIRE

Everitt, Tom; Lea, Gary; Hutter, Marcus

2018-01-01

The development of Artificial General Intelligence (AGI) promises to be a major event. Along with its many potential benefits, it also raises serious safety concerns (Bostrom, 2014). The intention of this paper is to provide an easily accessible and up-to-date collection of references for the emerging field of AGI safety. A significant number of safety problems for AGI have been identified. We list these, and survey recent research on solving them. We also cover works on how best to think of ...

Uncertainty and sensitivity analysis in a Probabilistic Safety Analysis level-1

International Nuclear Information System (INIS)

Nunez Mc Leod, Jorge E.; Rivera, Selva S.

1996-01-01

A methodology for sensitivity and uncertainty analysis, applicable to a Probabilistic Safety Assessment Level I has been presented. The work contents are: correct association of distributions to parameters, importance and qualification of expert opinions, generations of samples according to sample sizes, and study of the relationships among system variables and systems response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well as different graphical visualization for the control of the study. (author)

Safety Assessment of the New Very Low-Level Waste Disposal Installation at El Cabril, Spain

International Nuclear Information System (INIS)

Lopez, I.; Navarro, M.; Zuloaga, P.

2009-01-01

The sixth General Radioactive Waste Plan approved by the Spanish government in 2006, foresees important volumes of wastes with a very low content of radioactivity mainly coming from the dismantling of nuclear power plants, along with the occurrence of some radiological industrial incidents in the past. This fact has boosted the construction of a new disposal installation, specifically designed for this category of waste. This new installation is part of the existing low and intermediate level waste (LILW) disposal facility at El Cabril, and includes four cells with a total capacity of around 130,000 m 3 . The design of the cells is consistent with the European Directive for the disposal of hazardous waste and fulfils the same basic safety criteria as the present facility for LILW. The safety assessment methodology applied for the very low level waste (VLLW) installation is fully coherent with the approach adopted for the existing disposal facility for low and intermediate level waste (concrete vaults disposal system) and takes into account the potential impact of the new installation during both the operational and long-term periods. The license for the VLLW installation was granted by the Spanish Ministry of Industry, Tourism and Commerce (MITYC) in July 2008, following technical approval by the Nuclear Safety Council (CSN), and the first disposal operation occurred in October 2008. (authors)

On personal safety culture

International Nuclear Information System (INIS)

Chen Zigen

1996-01-01

The paper mainly expounds the personal safety culture, including the following aspects: the attitude to exploration, strict methods and the habit of exchange etc. It points out that straightening the education of safety culture and heightening the level of personal safety culture can get not only high-level safety but also high-level quality

Safety training: places available in September 2014

CERN Multimedia

HSE Unit

2014-01-01

There are places available in the forthcoming Safety courses. For updates and registrations, please refer to the Safety Training Catalogue (see here).   Please note that there are 7 places left on the “Territorial Safety Officer (TSO) – Initial” course on September 23-25, 2014 (in French).

Improving construction site safety through leader-based verbal safety communication.

Science.gov (United States)

Kines, Pete; Andersen, Lars P S; Spangenberg, Soren; Mikkelsen, Kim L; Dyreborg, Johnny; Zohar, Dov

2010-10-01

The construction industry is one of the most injury-prone industries, in which production is usually prioritized over safety in daily on-site communication. Workers have an informal and oral culture of risk, in which safety is rarely openly expressed. This paper tests the effect of increasing leader-based on-site verbal safety communication on the level of safety and safety climate at construction sites. A pre-post intervention-control design with five construction work gangs is carried out. Foremen in two intervention groups are coached and given bi-weekly feedback about their daily verbal safety communications with their workers. Foremen-worker verbal safety exchanges (experience sampling method, n=1,693 interviews), construction site safety level (correct vs. incorrect, n=22,077 single observations), and safety climate (seven dimensions, n=105 questionnaires) are measured over a period of up to 42 weeks. Baseline measurements in the two intervention and three control groups reveal that foremen speak with their workers several times a day. Workers perceive safety as part of their verbal communication with their foremen in only 6-16% of exchanges, and the levels of safety at the sites range from 70-87% (correct observations). Measurements from baseline to follow-up in the two intervention groups reveal that safety communication between foremen and workers increases significantly in one of the groups (factor 7.1 increase), and a significant yet smaller increase is found when the two intervention groups are combined (factor 4.6). Significant increases in the level of safety are seen in both intervention groups (7% and 12% increases, respectively), particularly in regards to 'access ways' and 'railings and coverings' (39% and 84% increases, respectively). Increases in safety climate are seen in only one of the intervention groups with respect to their 'attention to safety.' No significant trend changes are seen in the three control groups on any of the three measures

Safety analysis and synthesis using fuzzy sets and evidential reasoning

International Nuclear Information System (INIS)

Wang, J.; Yang, J.B.; Sen, P.

1995-01-01

This paper presents a new methodology for safety analysis and synthesis of a complex engineering system with a structure that is capable of being decomposed into a hierarchy of levels. In this methodology, fuzzy set theory is used to describe each failure event and an evidential reasoning approach is then employed to synthesise the information thus produced to assess the safety of the whole system. Three basic parameters--failure likelihood, consequence severity and failure consequence probability, are used to analyse a failure event. These three parameters are described by linguistic variables which are characterised by a membership function to the defined categories. As safety can also be clearly described by linguistic variables referred to as the safety expressions, the obtained fuzzy safety score can be mapped back to the safety expressions which are characterised by membership functions over the same categories. This mapping results in the identification of the safety of each failure event in terms of the degree to which the fuzzy safety score belongs to each of the safety expressions. Such degrees represent the uncertainty in safety evaluations and can be synthesised using an evidential reasoning approach so that the safety of the whole system can be evaluated in terms of these safety expressions. Finally, a practical engineering example is presented to demonstrate the proposed safety analysis and synthesis methodology

Occurrence of lead-related symptoms below the current occupational safety and health act allowable blood lead levels.

Science.gov (United States)

Rosenman, Kenneth D; Sims, Amy; Luo, Zhehui; Gardiner, Joseph

2003-05-01

To determine the occurrence of symptoms of lead toxicity at levels below the current allowable Occupational Safety and Health Act blood lead level of 50 micrograms/dL, standardized telephone interviews were conducted of individuals reported to a statewide laboratory-based surveillance system. Four hundred and ninety-seven, or 75%, of the eligible participants were interviewed. Gastrointestinal, musculoskeletal, and nervous system symptoms increased with increasing blood lead levels. Nervous, gastrointestinal, and musculoskeletal symptoms all began to be increased in individuals with blood leads between 30-39 micrograms/dL and possibly at levels as low as 25-30 micrograms/dL for nervous system symptoms. The results of this study of increased symptoms are consistent with and provide added weight to previous results showing subclinical changes in the neurologic and renal systems and sperm counts at blood lead levels currently allowed by the Occupational Safety and Health Act.

Low-level violence in schools: is there an association between school safety measures and peer victimization?

Science.gov (United States)

Blosnich, John; Bossarte, Robert

2011-02-01

Low-level violent behavior, particularly school bullying, remains a critical public health issue that has been associated with negative mental and physical health outcomes. School-based prevention programs, while a valuable line of defense to stave off bullying, have shown inconsistent results in terms of decreasing bullying. This study explored whether school safety measures (eg, security guards, cameras, ID badges) were associated with student reports of different forms of peer victimization related to bullying. Data came from the 2007 School Crime Supplement of the National Crime Victimization Survey. Chi-square tests of independence were used to examine differences among categorical variables. Logistic regression models were constructed for the peer victimization outcomes. A count variable was constructed among the bullying outcomes (0-7) with which a Poisson regression model was constructed to analyze school safety measures' impacts on degree of victimization. Of the various school safety measures, only having adults in hallways resulted in a significant reduction in odds of being physically bullied, having property vandalized, or having rumors spread. In terms of degree of victimization, having adults and/or staff supervising hallways was associated with an approximate 26% decrease in students experiencing an additional form of peer victimization. Results indicated that school safety measures overall were not associated with decreased reports of low-level violent behaviors related to bullying. More research is needed to further explore what best promotes comprehensive safety in schools. © 2011, American School Health Association.

SGHWR fuel performance, safety and reliability

International Nuclear Information System (INIS)

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

1977-01-01

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

Construction Safety And Health Factors At The Industry Level: The Case Of Singapore

Directory of Open Access Journals (Sweden)

Charles Y.J. Cheah

2007-12-01

Full Text Available The construction industry is one of the most hazardous industries due to the unique nature of its products and the processes involved. Recent occurrences of highly publicized construction site accidents in Singapore have highlighted the immediate needs for the local industry to address safety and attention at the industry level. The objective of this paper is to examine issues and critical factors affecting S&H standards in Singapore. Clearly, collective efforts should be pursued at the industry level as the country moves towards the ultimate safety management strategy of self-regulation. The findings also indicate that the challenge of making worksites safe should not be placed solely on the contractors but should be shared by all parties affecting the value chain of construction, including the developers, the consultants and the government. The factors identified through factor analysis may inform legislators and industry practitioners in terms of the sources of problems and help develop effective strategies for improvement. Some of the experiences mentioned in the paper could also be relevant to other countries facing similar circumstances.

Multiperiod Telser’s Safety-First Portfolio Selection with Regime Switching

Directory of Open Access Journals (Sweden)

Chuangwei Lin

2018-01-01

Full Text Available This paper investigates a multiperiod Telser’s safety-first portfolio selection model with regime switching where the returns of the assets are assumed to depend on the market states modulated by a discrete-time Markov chain. The investor aims to maximize the expected terminal wealth and does not want the probability of the terminal wealth to fall short of a disaster level to exceed a predetermined number called the risk control level. Referring to Tchebycheff inequality, we modify Telser’s safety-first model to the case that aims to maximize the expected terminal wealth subject to a constraint where the upper bound of the disaster probability is less than the risk control level. By the Lagrange multiplier technique and the embedding method, we study in detail the existence of the optimal strategy and derive the closed-form optimal strategy. Finally, by mathematical and numerical analysis, we analyze the effects of the disaster level, the risk control level, the transition matrix of the Markov chain, the expected excess return, and the variance of the risky return.

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

Directory of Open Access Journals (Sweden)

Husin Husrul Nizam

2014-01-01

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

The spread model of food safety risk under the supply-demand disturbance.

Science.gov (United States)

Wang, Jining; Chen, Tingqiang

2016-01-01

In this paper, based on the imbalance of the supply-demand relationship of food, we design a spreading model of food safety risk, which is about from food producers to consumers in the food supply chain. We use theoretical analysis and numerical simulation to describe the supply-demand relationship and government supervision behaviors' influence on the risk spread of food safety and the behaviors of the food producers and the food retailers. We also analyze the influence of the awareness of consumer rights protection and the level of legal protection of consumer rights on the risk spread of food safety. This model contributes to the explicit investigation of the influence relationship among supply-demand factors, the regulation behavioral choice of government, the behavioral choice of food supply chain members and food safety risk spread. And this paper provides a new viewpoint for considering food safety risk spread in the food supply chain, which has a great reference for food safety management.

Nuclear safety culture and nuclear safety supervision

International Nuclear Information System (INIS)

Chai Jianshe

2013-01-01

In this paper, the author reviews systematically and summarizes up the development process and stage characteristics of nuclear safety culture, analysis the connotation and characteristics of nuclear safety culture, sums up the achievements of our country's nuclear safety supervision, dissects the challenges and problems of nuclear safety supervision. This thesis focused on the relationship between nuclear safety culture and nuclear safety supervision, they are essential differences, but there is a close relationship. Nuclear safety supervision needs to introduce some concepts of nuclear safety culture, lays emphasis on humanistic care and improves its level and efficiency. Nuclear safety supervision authorities must strengthen nuclear safety culture training, conduct the development of nuclear safety culture, make sure that nuclear safety culture can play significant roles. (author)

Serum Inhibin B and follicle-stimulating hormone levels as tools in the evaluation of infertile men: significance of adequate reference values from proven fertile men

DEFF Research Database (Denmark)

Andersson, A.-M.; Petersen, Jørgen Holm; Jørgensen, N.

2004-01-01

Inhibin B and FSH levels in 289 idiopathic infertile men were compared with reference materials consisting of 303 proven fertile men (reference group 1) and 307 healthy men from the general population with unknown fertility status (reference group 2). The diagnostic power of these two serum markers...... of spermatogenesis was evaluated by the use of receiver operating characteristic plot analysis, and an example of how both markers can be used simultaneously in a bivariate reference chart is presented. Inhibin B levels were significantly lower and FSH levels were significantly higher in the infertile men, compared...

Determination of Natural Levels of Radionuclides in Proposed Mushroom Reference Material (A Proficiency Test Exercise)

International Nuclear Information System (INIS)

Waheed, S.; Rahman, A.; Siddique, N.; Ahmad, S.; Zaidi, J.H.

2006-08-01

A proficiency test (PT) was organized within the framework of international Atomic Energy Agency (IAEA) project INT/1/054, entitled 'Preparation' of Reference Materials and Organization of Proficiency Test Rounds'. This exercise served to estimate the proficiency of the analytical laboratories from participating countries. This report presents the results of the proficiency test exercise on the proposed Mushroom Reference Material for the determination of natural levels of radionuclides. Laboratories from 6 different countries submitted data on the following three radionuclides: /sup 134/Cs, /sup 137/Cs, /sup 40/K. Results for /sup 134/Cs, 137/sup 137/Cs, and /sup 40/K in the mushroom reference material were reported by three or more participating laboratories and could be subjected to statistical evaluation. The original data of these raionuclides was subjected to a computer program 'Histo Vession 2.1' provided by IAEA. The four outlier tests i.e. Dixon, Grubbs, Skewness and Kurtosis were applied to the data sets. All values for these three radionuclides were accepted by the software. Consensus (overall) mean value, absolute standard deviation, relative standard deviation, standard error, median and range of values for these three radionuclides have been are obtained (at significance level 0.05). the consensus mean values and confidence intervals are given./sup 134/Cs: 4.4 Bq/kg (3.4-5.3 Bq/kg) /sup 137/Cs: 2899 Bq/kg (2740-3058 Bq/kg) /sup 40/K: 1136 Bq/kg (1046-1226 Bq/kg). (author)

Maximum surface level and temperature histories for Hanford waste tanks

International Nuclear Information System (INIS)

Flanagan, B.D.; Ha, N.D.; Huisingh, J.S.

1994-01-01

Radioactive defense waste resulting from the chemical processing of spent nuclear fuel has been accumulating at the Hanford Site since 1944. This waste is stored in underground waste-storage tanks. The Hanford Site Tank Farm Facilities Interim Safety Basis (ISB) provides a ready reference to the safety envelope for applicable tank farm facilities and installations. During preparation of the ISB, tank structural integrity concerns were identified as a key element in defining the safety envelope. These concerns, along with several deficiencies in the technical bases associated with the structural integrity issues and the corresponding operational limits/controls specified for conduct of normal tank farm operations are documented in the ISB. Consequently, a plan was initiated to upgrade the safety envelope technical bases by conducting Accelerated Safety Analyses-Phase 1 (ASA-Phase 1) sensitivity studies and additional structural evaluations. The purpose of this report is to facilitate the ASA-Phase 1 studies and future analyses of the single-shell tanks (SSTs) and double-shell tanks (DSTs) by compiling a quantitative summary of some of the past operating conditions the tanks have experienced during their existence. This report documents the available summaries of recorded maximum surface levels and maximum waste temperatures and references other sources for more specific data

Progress report on safety research of high-level waste management for the period April, 1981 to March, 1982

International Nuclear Information System (INIS)

Tashiro, Shingo

1982-10-01

Main results obtained on Safety Research of High-Level Waste Management in 1981 were edited. The research tjeme are following. (1) Characterization of vitrified waste. (2) Alternative waste form development. (3) Durability tests for HLW storage facility. (4) Safety evaluation of geologic disposal. (5) Preparation for hot test. (author)

Development of a computer code for low-and intermediate-level radioactive waste disposal safety assessment

International Nuclear Information System (INIS)

Park, J. W.; Kim, C. L.; Lee, E. Y.; Lee, Y. M.; Kang, C. H.; Zhou, W.; Kozak, M. W.

2002-01-01

A safety assessment code, called SAGE (Safety Assessment Groundwater Evaluation), has been developed to describe post-closure radionuclide releases and potential radiological doses for low- and intermediate-level radioactive waste (LILW) disposal in an engineered vault facility in Korea. The conceptual model implemented in the code is focused on the release of radionuclide from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. The radionuclide transport equations are solved by spatially discretizing the disposal system into a series of compartments. Mass transfer between compartments is by diffusion/dispersion and advection. In all compartments, radionuclides are decayed either as a single-member chain or as multi-member chains. The biosphere is represented as a set of steady-state, radionuclide-specific pathway dose conversion factors that are multiplied by the appropriate release rate from the far field for each pathway. The code has the capability to treat input parameters either deterministically or probabilistically. Parameter input is achieved through a user-friendly Graphical User Interface. An application is presented, which is compared against safety assessment results from the other computer codes, to benchmark the reliability of system-level conceptual modeling of the code

Reference hearing threshold levels for chirp signals delivered by an ER-3A insert earphone

DEFF Research Database (Denmark)

Gøtsche-Rasmussen, Kristian; Poulsen, Torben; Elberling, Claus

2012-01-01

back from a Tucker Davies Technologies System II, and a Matlab program controlled the test setup. The results are specified in dB peak-to-peak equivalent threshold sound pressure levels (dB peETSPL). Study sample: The test group consisted of 25 otologically-normal young adults (age 18–25 years......Objective: To establish reference hearing threshold levels for chirps and frequency-specific chirps. Design: Hearing thresholds were determined monaurally for broad-band chirps and octave-band chirps using the Etymotic Research, ER-3A insert earphone. The chirps were presented using two repetition...

Standard digital reference images for inspection of aluminum castings

CERN Document Server

American Society for Testing and Materials. Philadelphia

2005-01-01

1.1 These digital reference images illustrate the types and degrees of discontinuities that may be found in aluminum-alloy castings. The castings illustrated are in thicknesses of 1/ 4 in. [6.35 mm] and 3/4 in. [19.1mm]. 1.2 All areas of this standard may be open to agreement between the cognizant engineering organization and the supplier, or specific direction from the cognizant engineering organization. These items should be addressed in the purchase order or the contract. 1.3 The values stated in inch-pound units are to be regarded as standard. 1.4 These digital reference images are not intended to illustrate the types and degrees of discontinuities found in aluminum-alloy castings when performing film radiography. If performing film radiography of aluminum-alloy castings, refer to Reference Radiographs E 155. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and he...

A comparison and cross-reference of commercial low-level radioactive waste acceptance criteria

International Nuclear Information System (INIS)

Kerr, T.A.

1997-04-01

This document, prepared by the National Low-Level Waste Management Program at the Idaho National Engineering and Environmental Laboratory, is a comparison and cross-reference of commercial low-level radioactive waste acceptance criteria. Many of these are draft or preliminary criteria as well as implemented criteria at operating low-level radioactive waste management facilities. Waste acceptance criteria from the following entities are included: US Nuclear Regulatory Commission, South Carolina, Washington, Utah, Nevada, California, illinois, Texas, North Carolina, Nebraska, Pennsylvania, New York, and the Midwest Compact Region. Criteria in the matrix include the following: physical form, chemical form, liquid limits, void space in packages, concentration averaging, types of packaging, chelating agents, solidification media, stability requirements, sorptive media, gas, oil, biological waste, pyrophorics, source material, special nuclear material, package dimensions, incinerator ash, dewatered resin, transuranics, and mixed waste. Each criterion in the matrix is cross-referenced to its source document so that exact requirements can be determined

Towards an International Approach to Nuclear Safety

International Nuclear Information System (INIS)

Tomihiro Taniguchi

2006-01-01

This document presents in a series of transparencies the different activities of the IAEA: Introduction of International Atomic Energy Agency, Changing world, Changing Technology, Changing Global Security, Developing Innovative Nuclear Energy Systems, Global Nuclear Safety Regime, IAEA Safety Standards: Hierarchy - Global Reference for Striving for Excellence, IAEA Safety Reviews and Services: Integrated Safety Approach, Global Knowledge Network - Asian Nuclear Safety Network, Safety Issues and Challenges, Synergy between Safety and Security, Recent Developments: Safety and Security of Radioactive Sources, Convention on Physical Protection of Nuclear Material (CPPNM), Incident and Emergency Preparedness and Response, Holistic Approach for Safety and Security, Sustainable Development. (J.S.)

A review of the status of nuclear safety in the Central and East European Countries with special reference to the evaluation of the situation in Romania

International Nuclear Information System (INIS)

Webster, Simon

2001-01-01

This paper briefly presents the status of nuclear safety in the candidate countries in the light of the progress towards accession to the European Union, and draws particular attention to the recent evaluation made by the Atomic Questions Group of the Council of Ministers by reference to the general recommendations and the specific recommendations addressed to Romania appearing in the evaluation report. (author)

Investigation of Relationship between Level of Awareness around Health, Safety and Environment Management System and Its Effects on Safety Climate and Risk Perception by Employees in an Iran Oil Refinery, 2015

Directory of Open Access Journals (Sweden)

Milad Ahmadi Marzaleh

2017-02-01

Full Text Available Special attention of and oil, gas and petrochemical large corporation in the world to the HSE management system is due to its importance in the designing and development of products, services and processes by considering its health, safety and environment requirements. Staff's perception of the existing job risks has a significant impact on their safe behavior at work. This study was conducted to determine the relationship between safety climate and staff's perception of risk with an awareness level among employees of the HSE management system in an oil refinery in Kermanshah. The study population was employees in one of the oil refineries in Iran. After designing and questionnaire psychometric assessment of staff knowledge of HSE management system (Cronbach's alpha was 0.9 and its validity was assessed by certified professionals, Loughborough safety climate questionnaire and Flin risk perception questionnaire were used. Data analysis was performed using SPSS V22 software. Results showed that the relationship between safety climate and awareness level of the HSE management system; also the relationship between safety climate and perception of risk was also getting significant. However, the relationship between perception of risk and awareness level of the HSE management system was not significant. The results of this study showed a moderate awareness of HSE Management System in refinery workers. In this regard, appropriate and proper management policy should be committed to improving the situation. The results of this study is a profile the situation of safety climate in the refinery, which can be used as an indicator for the development of preventive policies and evaluate the performance of the organization's safety and the results of the safety improvement organization.

A suggestion of reference data for flow distribution at ankle and foot level using quantitative 99Tc-HDP three-phase bone scintigraphy

DEFF Research Database (Denmark)

Tøndevold, Niklas; Reving, Sofie; Møller, Nette

2012-01-01

To determine reference intervals for quantitative 99mTc-hydroxymethylene diphosphonate (99mTc-HDP) three-phase bone scintigraphy regarding flow distribution at ankle and mid-foot level.......To determine reference intervals for quantitative 99mTc-hydroxymethylene diphosphonate (99mTc-HDP) three-phase bone scintigraphy regarding flow distribution at ankle and mid-foot level....

Mapping of radio frequency electromagnetic field exposure levels in outdoor environment and comparing with reference levels for general public health.

Science.gov (United States)

Cansiz, Mustafa; Abbasov, Teymuraz; Kurt, M Bahattin; Celik, A Recai

2018-03-01

In this study, radio frequency electromagnetic field exposure levels were measured on the main streets in the city center of Diyarbakır, Turkey. Measured electric field levels were plotted on satellite imagery of Diyarbakır and were compared with exposure guidelines published by the International Commission on Non-Ionizing Radiation Protection (ICNIRP). Exposure measurements were performed in dense urban, urban and suburban areas each day for 7 consecutive days. The measurement system consisted of high precision and portable spectrum analyzer, three-axis electric field antenna, connection cable and a laptop which was used to record the measurement samples as a data logger. The highest exposure levels were detected for two places, which are called Diclekent and Batıkent. It was observed that the highest instantaneous electric field strength value for Batıkent was 7.18 V/m and for Diclekent was 5.81 V/m. It was statistically determined that the main contributor band to the total exposure levels was Universal Mobile Telecommunications System band. Finally, it was concluded that all measured exposure levels were lower than the reference levels recommended by ICNIRP for general public health.

Safety level assessment in the production environment using Fuzzy logic

Directory of Open Access Journals (Sweden)

2012-05-01

Conclusion: The application of the proposed method can reveal which safety items and factors are most important in improving workers safety, and therefore decide where to concentrate resources in order to improve the safety of the work environment.

Multinational Design Evaluation Programme (MDEP) - Safety Goals

International Nuclear Information System (INIS)

Vaughan, G.J.

2011-01-01

goals and/or targets that can be seen to be clearly related to the higher level ones and set consistent requirements for different technologies. MDEP has, therefore, established a sub-committee to carry out this work. This paper is a review of the work of this sub-committee over the last eighteen months or so in attempting to outline a framework within which potential goals can be included, as a move towards harmonization. If the work is successful, and leads to an agreed MDEP approach, it will greatly assist in the process of harmonisation. It is important to emphasise that this work has not as yet attempted to derive specific safety goals per se, but to derive a framework, which can be used to understand how the deterministic and the probabilistic elements can be integrated in establishing reference level of safety

Level of knowledge among the population of radiation safety basic issues

Directory of Open Access Journals (Sweden)

S. A. Zelencova

2015-01-01

Full Text Available The goal of research was to determine the level of knowledge among the population on issues like sources of ionising radiation, methods of ionising radiation measurement, measures of self-protection in case of threating or actual radioactive pollution in the district, and to study self-estimation by the population of their knowledge of radiation safety issues. Research was carried out using the method of questioning of population groups in three regions close to the places of previous peaceful nuclear explosions (Arkhangelsk, Murmansk and Tyumen regions, and in five Far East regions of the Russian Federation (Kamchatka, Khabarovsk, Primorsky, Magadan and South-Sakhalin regions after radiation accident in Japan at "Fukushima-1" NPP. This research included processing of 243 questionnaires from the regions close to places of previous peaceful nuclear explosions and 216 questionnaires from the Far East regions.The analysis of obtained questioning results enabled to make the following conclusions: the level of knowledge among the population about the basic concepts of radiation safety appeared to be generally low among respondents of all eight territories. Considerable number of respondents in seven groups correctly mentioned the x-ray device as a source of ionising radiation (from 71 to 88 % of answers. In Murmansk region – only 52 % of the answers. Respondents of the same seven groups often correctly answered the question on how to detect ionising radiation (only with devices – from 68 to 98 % in different groups. The smallest number of correct answers to this question (42 % is also noted among respondents from the Murmansk region.Level of knowledge on self-protection measures at threating or actual radioactive pollution of the places of residence appeared a little higher among the Far East region population, who had actual concerns regarding the threat of radioactive pollution at the present time. However, in all eight investigated groups

A review of models relevant to road safety.

Science.gov (United States)

Hughes, B P; Newstead, S; Anund, A; Shu, C C; Falkmer, T

2015-01-01

It is estimated that more than 1.2 million people die worldwide as a result of road traffic crashes and some 50 million are injured per annum. At present some Western countries' road safety strategies and countermeasures claim to have developed into 'Safe Systems' models to address the effects of road related crashes. Well-constructed models encourage effective strategies to improve road safety. This review aimed to identify and summarise concise descriptions, or 'models' of safety. The review covers information from a wide variety of fields and contexts including transport, occupational safety, food industry, education, construction and health. The information from 2620 candidate references were selected and summarised in 121 examples of different types of model and contents. The language of safety models and systems was found to be inconsistent. Each model provided additional information regarding style, purpose, complexity and diversity. In total, seven types of models were identified. The categorisation of models was done on a high level with a variation of details in each group and without a complete, simple and rational description. The models identified in this review are likely to be adaptable to road safety and some of them have previously been used. None of systems theory, safety management systems, the risk management approach, or safety culture was commonly or thoroughly applied to road safety. It is concluded that these approaches have the potential to reduce road trauma. Copyright © 2014 Elsevier Ltd. All rights reserved.

Safety Culture Monitoring: How to Assess Safety Culture in Real Time?

International Nuclear Information System (INIS)

Zronek, B.; Maryska, J.; Treslova, L.

2016-01-01

Do you know what is current level of safety culture in your company? Are you able to follow trend changes? Do you know what your recent issues are? Since safety culture is understood as vital part of nuclear industry daily life, it is crucial to know what the current level is. It is common to perform safety culture survey or ad hoc assessment. This contribution shares Temelin NPP, CEZ approach how to assess safety culture level permanently. Using behavioral related outputs of gap solving system, observation program, dedicated surveys, regulatory assessment, etc., allows creating real time safety culture monitoring without the need to perform any other activities. (author)

Safety assessment and verification for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2001-01-01

verification' are used differently in different countries. The way that these terms have been used in this Safety Guide is explained in Section 2. The term 'design' as used here includes the specifications for the safe operation and management of the plant. This Safety Guide identifies the key recommendations for carrying out the safety assessment and the independent verification. It provides detailed guidance in support of IAEA, Safety of Nuclear Power Plants: Design, Safety Standards Series No. NS-R-1 (2000), particularly in the area of safety analysis. However, this does not include all the technical details which are available and reference is made to other IAEA publications on specific design issues and safety analysis methods. Specific deterministic or probabilistic safety targets or radiological limits can vary in different countries and are the responsibility of the regulatory body. This Safety Guide provides some references to targets and limits established by international organizations. Operators, and sometimes designers, may also set their own safety targets which may be more stringent than those set by the regulator or may address different aspects of safety. In some countries operators are expected to do this as part of their 'ownership' of the entire safety case. This Safety Guide does not include specific recommendations for the safety assessment of those plant systems for which dedicated Safety Guides exist. Section 2 defines the terms 'safety assessment', 'safety analysis' and 'independent verification' and outlines their relationship. Section 3 gives the key recommendations for the safety assessment of the principal and plant design requirements. Section 4 gives the key recommendations for safety analysis. It describes the identification of postulated initiating events (PIEs), which are used throughout the safety assessment including the safety analysis, the deterministic transient analysis and severe accident analysis, and the probabilistic safety analysis

Further activities of safety culture toward nuclear transportation industry

International Nuclear Information System (INIS)

Machida, Y.; Shimakura, D.

2004-01-01

On September 30, 1999, a criticality accident occurred at the uranium processing facility of the JCO Co. Ltd. (hereinafter referred to as ''JCO'') Tokai plant, located in Tokaimura, Ibaraki Prefecture. This was an unprecedented accident in Japan's history of peaceful use of nuclear power, resulting in three workers exposed to severe radiation, two of whom died, and the evacuation and enforced indoor confinement of local residents. Nuclear power suppliers must take personal responsibility for ensuring safety. In this connection, the electric power industry, heavy electric machinery manufacturers, fuel fabricators, and nuclear power research organizations gathered together to establish the Nuclear Safety Network (NSnet) in December 1999, based on the resolve to share and improve the level of the safety culture across the entire nuclear power industry and to assure that such an accident never occurs again. NSnet serves as a link between nuclear power enterprises, research organizations, and other bodies, based on the principles of equality and reciprocity. A variety of activities are pursued, such as diffusing a safety culture, implementing mutual evaluation among members, and exchanging safety-related information. Aiming to share and improve the safety culture throughout the entire nuclear power industry, NSnet thoroughly implements the principle of safety first, while at the same time making efforts to restore trust in nuclear power