Safety culture management and quantitative indicator evaluation

International Nuclear Information System (INIS)

Mandula, J.

2002-01-01

This report discuses a relationship between safety culture and evaluation of quantitative indicators. It shows how a systematic use of generally shared operational safety indicators may contribute to formation and reinforcement of safety culture characteristics in routine plant operation. The report also briefly describes the system of operational safety indicators used at the Dukovany plant. It is a PC database application enabling an effective work with the indicators and providing all users with an efficient tool for making synoptic overviews of indicator values in their links and hierarchical structure. Using color coding, the system allows quick indicator evaluation against predefined limits considering indicator value trends. The system, which has resulted from several-year development, was completely established at the plant during the years 2001 and 2002. (author)

Book Review: Qualitative-Quantitative Analyses of Dutch and ...

African Journals Online (AJOL)

Abstract. Book Title: Qualitative-Quantitative Analyses of Dutch and Afrikaans Grammar and Lexicon. Book Author: Robert S. Kirsner. 2014. John Benjamins Publishing Company ISBN 9789027215772, price ZAR481.00. 239 pages ...

Quantitative analyses of shrinkage characteristics of neem ...

African Journals Online (AJOL)

Quantitative analyses of shrinkage characteristics of neem (Azadirachta indica A. Juss.) wood were carried out. Forty five wood specimens were prepared from the three ecological zones of north eastern Nigeria, viz: sahel savanna, sudan savanna and guinea savanna for the research. The results indicated that the wood ...

Selected problems and results of the transient event and reliability analyses for the German safety study

International Nuclear Information System (INIS)

Hoertner, H.

1977-01-01

For the investigation of the risk of nuclear power plants loss-of-coolant accidents and transients have to be analyzed. The different functions of the engineered safety features installed to cope with transients are explained. The event tree analysis is carried out for the important transient 'loss of normal onsite power'. Preliminary results of the reliability analyses performed for quantitative evaluation of this event tree are shown. (orig.) [de

Accelerated safety analyses - structural analyses Phase I - structural sensitivity evaluation of single- and double-shell waste storage tanks

International Nuclear Information System (INIS)

Becker, D.L.

1994-11-01

Accelerated Safety Analyses - Phase I (ASA-Phase I) have been conducted to assess the appropriateness of existing tank farm operational controls and/or limits as now stipulated in the Operational Safety Requirements (OSRs) and Operating Specification Documents, and to establish a technical basis for the waste tank operating safety envelope. Structural sensitivity analyses were performed to assess the response of the different waste tank configurations to variations in loading conditions, uncertainties in loading parameters, and uncertainties in material characteristics. Extensive documentation of the sensitivity analyses conducted and results obtained are provided in the detailed ASA-Phase I report, Structural Sensitivity Evaluation of Single- and Double-Shell Waste Tanks for Accelerated Safety Analysis - Phase I. This document provides a summary of the accelerated safety analyses sensitivity evaluations and the resulting findings

Application of geostatistical methods to long-term safety analyses for radioactive waste repositories

International Nuclear Information System (INIS)

Roehlig, K.J.

2001-01-01

Long-term safety analyses are an important part of the design and optimisation process as well as of the licensing procedure for final repositories for radioactive waste in deep geological formations. For selected scenarios describing possible evolutions of the repository system in the post-closure phase, quantitative consequence analyses are performed. Due to the complexity of the phenomena of concern and the large timeframes under consideration, several types of uncertainties have to be taken into account. The modelling work for the far-field (geosphere) surrounding or overlaying the repository is based on model calculations concerning the groundwater movement and the resulting migration of radionuclides which possibly will be released from the repository. In contrast to engineered systems, the geosphere shows a strong spatial variability of facies, materials and material properties. The paper presented here describes the first steps towards a quantitative approach for an uncertainty assessment taking into account this variability. Due to the availability of a large amount of data and information of several types, the Gorleben site (Germany) has been used for a case study in order to demonstrate the method. (orig.)

Implementing partnerships in nonreactor facility safety analyses

International Nuclear Information System (INIS)

Courtney, J.C.; Perry, W.H.; Phipps, R.D.

1996-01-01

Faculty and students from LSU have been participating in nuclear safety analyses and radiation protection projects at ANL-W at INEL since 1973. A mutually beneficial relationship has evolved that has resulted in generation of safety-related studies acceptable to Argonne and DOE, NRC, and state regulatory groups. Most of the safety projects have involved the Hot Fuel Examination Facility or the Fuel Conditioning Facility; both are hot cells that receive spent fuel from EBR-II. A table shows some of the major projects at ANL-W that involved LSU students and faculty

Safety analyses for high-temperature reactors

International Nuclear Information System (INIS)

Mueller, A.

1978-01-01

The safety evaluation of HTRs may be based on the three methods presented here: The licensing procedure, the probabilistic risk analysis, and the damage extent analysis. Thereby all safety aspects - from normal operation to the extreme (hypothetical) accidents - of the HTR are covered. The analyses within the licensing procedure of the HTR-1160 have shown that for normal operation and for the design basis accidents the radiation exposures remain clearly below the maximum permissible levels as prescribed by the radiation protection ordinance, so that no real hazard for the population will avise from them. (orig./RW) [de

Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

Energy Technology Data Exchange (ETDEWEB)

Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-10-01

The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort is to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis

Quantitative metagenomic analyses based on average genome size normalization

DEFF Research Database (Denmark)

Frank, Jeremy Alexander; Sørensen, Søren Johannes

2011-01-01

provide not just a census of the community members but direct information on metabolic capabilities and potential interactions among community members. Here we introduce a method for the quantitative characterization and comparison of microbial communities based on the normalization of metagenomic data...... marine sources using both conventional small-subunit (SSU) rRNA gene analyses and our quantitative method to calculate the proportion of genomes in each sample that are capable of a particular metabolic trait. With both environments, to determine what proportion of each community they make up and how......). These analyses demonstrate how genome proportionality compares to SSU rRNA gene relative abundance and how factors such as average genome size and SSU rRNA gene copy number affect sampling probability and therefore both types of community analysis....

EFFICIENT QUANTITATIVE RISK ASSESSMENT OF JUMP PROCESSES: IMPLICATIONS FOR FOOD SAFETY

OpenAIRE

Nganje, William E.

1999-01-01

This paper develops a dynamic framework for efficient quantitative risk assessment from the simplest general risk, combining three parameters (contamination, exposure, and dose response) in a Kataoka safety-first model and a Poisson probability representing the uncertainty effect or jump processes associated with food safety. Analysis indicates that incorporating jump processes in food safety risk assessment provides more efficient cost/risk tradeoffs. Nevertheless, increased margin of safety...

Development of the evaluation methods in reactor safety analyses and core characteristics

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

Quantitative software-reliability analysis of computer codes relevant to nuclear safety

International Nuclear Information System (INIS)

Mueller, C.J.

1981-12-01

This report presents the results of the first year of an ongoing research program to determine the probability of failure characteristics of computer codes relevant to nuclear safety. An introduction to both qualitative and quantitative aspects of nuclear software is given. A mathematical framework is presented which will enable the a priori prediction of the probability of failure characteristics of a code given the proper specification of its properties. The framework consists of four parts: (1) a classification system for software errors and code failures; (2) probabilistic modeling for selected reliability characteristics; (3) multivariate regression analyses to establish predictive relationships among reliability characteristics and generic code property and development parameters; and (4) the associated information base. Preliminary data of the type needed to support the modeling and the predictions of this program are described. Illustrations of the use of the modeling are given but the results so obtained, as well as all results of code failure probabilities presented herein, are based on data which at this point are preliminary, incomplete, and possibly non-representative of codes relevant to nuclear safety

Response surface use in safety analyses

International Nuclear Information System (INIS)

Prosek, A.

1999-01-01

When thousands of complex computer code runs related to nuclear safety are needed for statistical analysis, the response surface is used to replace the computer code. The main purpose of the study was to develop and demonstrate a tool called optimal statistical estimator (OSE) intended for response surface generation of complex and non-linear phenomena. The performance of optimal statistical estimator was tested by the results of 59 different RELAP5/MOD3.2 code calculations of the small-break loss-of-coolant accident in a two loop pressurized water reactor. The results showed that OSE adequately predicted the response surface for the peak cladding temperature. Some good characteristic of the OSE like monotonic function between two neighbor points and independence on the number of output parameters suggest that OSE can be used for response surface generation of any safety or system parameter in the thermal-hydraulic safety analyses.(author)

Regulatory support activities of JNES by thermal-hydraulic and safety analyses

International Nuclear Information System (INIS)

Kasahara, Fumio

2008-01-01

Current status and some related topics on regulatory support activities of Japan Nuclear Energy Safety Organization (JNES) by thermal-hydraulic and safety analyses are reported. The safety of nuclear facilities is secured primarily by plant owners and operators. However, the regulatory body NISA (Nuclear and Industrial Safety Agency) has conducted a strict regulation to confirm the adequacy of the site condition as well as the basic and detailed design. The JNES has been conducting independent analyses from applicants (audit analyses, etc.) by direction of NISA and supporting its review. In addition to the audit analysis, thermal-hydraulic and safety analyses are used in such areas as analytical evaluation for investigation of causes of accidents and troubles, level 2 PSA for risk informed regulation, etc. Recent activities of audit analyses are for the application of Tsuruga 3 and 4 (APWR), the spent fuel storage facility for the establishment, and the LMFBR Monju for the core change. For the trouble event evaluation, the criticality accident analysis of Sika1 was carried out and the evaluation of effectiveness of accident management (AM) measure for Tomari 3 (PWR) and Monju was performed. The analytical codes for these evaluations are continuously revised by reflecting the state-of-art technical information and validated using the information provided by the data from JAEA, OECD project, etc. (author)

SCALE Graphical Developments for Improved Criticality Safety Analyses

International Nuclear Information System (INIS)

Barnett, D.L.; Bowman, S.M.; Horwedel, J.E.; Petrie, L.M.

1999-01-01

New computer graphic developments at Oak Ridge National Ridge National Laboratory (ORNL) are being used to provide visualization of criticality safety models and calculational results as well as tools for criticality safety analysis input preparation. The purpose of this paper is to present the status of current development efforts to continue to enhance the SCALE (Standardized Computer Analyses for Licensing Evaluations) computer software system. Applications for criticality safety analysis in the areas of 3-D model visualization, input preparation and execution via a graphical user interface (GUI), and two-dimensional (2-D) plotting of results are discussed

Supporting Fernald Site Closure with Integrated Health and Safety Plans as Documented Safety Analyses

International Nuclear Information System (INIS)

Kohler, S.; Brown, T.; Fisk, P.; Krach, F.; Klein, B.

2004-01-01

At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 200 structures, including eight major nuclear production plants. There is one of twelve nuclear facilities still remaining (Silos containing uranium ore residues) with its own safety basis documentation. This paper presents the status of the FCP's safety basis documentation program, illustrating that all of the former nuclear facilities and activities have now replaced. Basis of Interim Operations (BIOs) with I-HASPs as their safety basis during the closure process

N reactor individual risk comparison to quantitative nuclear safety goals

International Nuclear Information System (INIS)

Wang, O.S.; Rainey, T.E.; Zentner, M.D.

1990-01-01

A full-scope level III probabilistic risk assessment (PRA) has been completed for N reactor, a US Department of Energy (DOE) production reactor located on the Hanford Reservation in the state of Washington. Sandia National Laboratories (SNL) provided the technical leadership for this work, using the state-of-the-art NUREG-1150 methodology developed for the US Nuclear Regulatory Commission (NRC). The main objectives of this effort were to assess the risks to the public and to the on-site workers posed by the operation of N reactor, to identify changes to the plant that could reduce the overall risk, and to compare those risks to the proposed NRC and DOE quantitative safety goals. This paper presents the methodology adopted by Westinghouse Hanford Company (WHC) and SNL for individual health risk evaluation, its results, and a comparison to the NRC safety objectives and the DOE nuclear safety guidelines. The N reactor results, are also compared with the five NUREG-1150 nuclear plants. Only internal events are compared here because external events are not yet reported in the current draft NUREG-1150. This is the first full-scope level III PRA study with a detailed quantitative safety goal comparison performed for DOE production reactors

Safety analyses of the electrical systems on VVER NPP

International Nuclear Information System (INIS)

Andel, J.

2004-01-01

Energoprojekt Praha has been the main entity responsible for the section on 'Electrical Systems' in the safety reports of the Temelin, Dukovany and Mochovce nuclear power plants. The section comprises 2 main chapters, viz. Offsite Power System (issues of electrical energy production in main generators and the link to the offsite transmission grid) and Onsite Power Systems (AC and DC auxiliary system, both normal and safety related). In the chapter on the off-site system, attention is paid to the analysis of transmission capacity of the 400 kV lines, analysis of transient stability, multiple fault analyses, and probabilistic analyses of the grid and NPP power system reliability. In the chapter on the on-site system, attention is paid to the power balances of the electrical sources and switchboards set for various operational and accident modes, checks of loading and function of service and backup sources, short circuit current calculations, analyses of electrical protections, and analyses of the function and sizing of emergency sources (DG sets and UPS systems). (P.A.)

Safety balance: Analysis of safety systems; Bilans de surete: analyse par les organismes de surete

Energy Technology Data Exchange (ETDEWEB)

Delage, M; Giroux, C

1990-12-01

Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses.

The role of CFD computer analyses in hydrogen safety management

International Nuclear Information System (INIS)

Komen, E.M.J; Visser, D.C; Roelofs, F.; Te Lintelo, J.G.T

2014-01-01

The risks of hydrogen release and combustion during a severe accident in a light water reactor have attracted considerable attention after the Fukushima accident in Japan. Reliable computer analyses are needed for the optimal design of hydrogen mitigation systems, like e.g. passive autocatalytic recombiners (PARs), and for the assessment of the associated residual risk of hydrogen combustion. Traditionally, so-called Lumped Parameter (LP) computer codes are being used for these purposes. In the last decade, significant progress has been made in the development, validation, and application of more detailed, three-dimensional Computational Fluid Dynamics (CFD) simulations for hydrogen safety analyses. The objective of the current paper is to address the following questions: - When are CFD computer analyses needed complementary to the traditional LP code analyses for hydrogen safety management? - What is the validation status of the CFD computer code for hydrogen distribution, mitigation, and combustion analyses? - Can CFD computer analyses nowadays be executed in practical and reliable way for full scale containments? The validation status and reliability of CFD code simulations will be illustrated by validation analyses performed for experiments executed in the PANDA, THAI, and ENACCEF facilities. (authors)

Swiss-Slovak cooperation program: a training strategy for safety analyses

International Nuclear Information System (INIS)

Husarcek, J.

2000-01-01

During the 1996-1999 period, a new training strategy for safety analyses was implemented at the Slovak Nuclear Regulatory Authority (UJD) within the Swiss-Slovak cooperation programme in nuclear safety (SWISSLOVAK). The SWISSLOVAK project involved the recruitment, training, and integration of the newly established team into UJD's organizational structure. The training strategy consisted primarily of the following two elements: a) Probabilistic Safety Analysis (PSA) applications (regulatory review and technical evaluation of Level-1/Level-2 PSAs; PSA-based operational events analysis, PSA applications to assessment of Technical Specifications; and PSA-based hardware and/or procedure modifications) and b) Deterministic accident analyses (analysis of accidents and regulatory review of licensee Safety Analysis Reports; analysis of severe accidents/radiological releases and the potential impact of the containment and engineered safety systems, including the development of technical bases for emergency response planning; and application of deterministic methods for evaluation of accident management strategies/procedure modifications). The paper discusses the specific aspects of the training strategy performed at UJD in both the probabilistic and deterministic areas. The integration of team into UJD's organizational structure is described and examples of contributions of the team to UJD's statutory responsibilities are provided. (author)

Method for quantitative assessment of nuclear safety computer codes

International Nuclear Information System (INIS)

Dearien, J.A.; Davis, C.B.; Matthews, L.J.

1979-01-01

A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

Reliability and safety analyses under fuzziness

International Nuclear Information System (INIS)

Onisawa, T.; Kacprzyk, J.

1995-01-01

Fuzzy theory, for example possibility theory, is compatible with probability theory. What is shown so far is that probability theory needs not be replaced by fuzzy theory, but rather that the former works much better in applications if it is combined with the latter. In fact, it is said that there are two essential uncertainties in the field of reliability and safety analyses: One is a probabilistic uncertainty which is more relevant for mechanical systems and the natural environment, and the other is fuzziness (imprecision) caused by the existence of human beings in systems. The classical probability theory alone is therefore not sufficient to deal with uncertainties in humanistic system. In such a context this collection of works will put a milestone in the arguments of probability theory and fuzzy theory. This volume covers fault analysis, life time analysis, reliability, quality control, safety analysis and risk analysis. (orig./DG). 106 figs

The profile of quantitative risk indicators in Krsko NPP

International Nuclear Information System (INIS)

Vrbanic, I.; Basic, I.; Bilic-Zabric, T.; Spiler, J.

2004-01-01

During the past decade strong initiative was observed which was aimed at incorporating information on risk into various aspects of operation of nuclear power plants. The initiative was observable in activities carried out by regulators as well as utilities and industry. It resulted in establishing the process, or procedure, which is often referred to as integrated decision making or risk informed decision making. In this process, engineering analyses and evaluations that are usually termed traditional and that rely on considerations of safety margins and defense in depth are supplemented by quantitative indicators of risk. Throughout the process, the plant risk was most commonly expressed in terms of likelihood of events involving damage to the reactor core and events with radiological releases to the environment. These became two commonly used quantitative indicators or metrics of plant risk (or, reciprocally, plant safety). They were evaluated for their magnitude (e.g. the expected number of events per specified time interval), as well as their profile (e.g. the types of contributing events). The information for quantitative risk indicators (to be used in risk informing process) is obtained from plant's probabilistic safety analyses or analyses of hazards. It is dependable on issues such as availability of input data or quality of model or analysis. Nuclear power plant Krsko has recently performed Periodic Safety Review, which was a good opportunity to evaluate and integrate the plant specific information on quantitative plant risk indicators and their profile. The paper discusses some aspects of quantitative plant risk profile and its perception.(author)

Nuclear power plants: Results of recent safety analyses

International Nuclear Information System (INIS)

Steinmetz, E.

1987-01-01

The contributions deal with the problems posed by low radiation doses, with the information currently available from analyses of the Chernobyl reactor accident, and with risk assessments in connection with nuclear power plant accidents. Other points of interest include latest results on fission product release from reactor core or reactor building, advanced atmospheric dispersion models for incident and accident analyses, reliability studies on safety systems, and assessment of fire hazard in nuclear installations. The various contributions are found as separate entries in the database. (DG) [de

Thermal hydraulic reactor safety analyses and experiments

International Nuclear Information System (INIS)

Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

1989-04-01

The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

International Nuclear Information System (INIS)

Aydogdu, K.; Boss, C. R.

2006-01-01

This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

European passive plant program preliminary safety analyses to support system design

International Nuclear Information System (INIS)

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

1999-01-01

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

Proposed quantitative approach to safety for nuclear power plants in Canada

International Nuclear Information System (INIS)

1995-07-01

A set of quantitative risk and frequency limits plus required processes is proposed to help ensure that a nuclear power plant in Canada meets the qualitative safety objectives defined in ACNS-2 and in IAEA 75-INSAG-3. As emphasized in this report, risks and hence doses are to be reduced below the limits using ALARA (As Low as Reasonably Achievable, economic and social factors being taken into account) or VIA (value-impact analysis) processes unless, in general, calculated risks and hence doses are below recommended de minimis levels. An updated version of ACNS-4, which will be issued as ACNS-21, will incorporate a statement of these limits and objectives as well as assessment criteria and procedures that will facilitate their application. The quantitative approach proposed here is consistent with a growing consensus on the need for, and the elements of, a quantitative approach to risk management of all major activities in an advanced industrial society. The ACNS recommends that the Atomic Energy Control Board adopt the proposed approach as a rational and coherent basis for nuclear power plant safety policy and requirements in Canada. (author). 68 refs., 4 tabs., 1 fig

Proposed quantitative approach to safety for nuclear power plants in Canada

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

A set of quantitative risk and frequency limits plus required processes is proposed to help ensure that a nuclear power plant in Canada meets the qualitative safety objectives defined in ACNS-2 and in IAEA 75-INSAG-3. As emphasized in this report, risks and hence doses are to be reduced below the limits using ALARA (As Low as Reasonably Achievable, economic and social factors being taken into account) or VIA (value-impact analysis) processes unless, in general, calculated risks and hence doses are below recommended de minimis levels. An updated version of ACNS-4, which will be issued as ACNS-21, will incorporate a statement of these limits and objectives as well as assessment criteria and procedures that will facilitate their application. The quantitative approach proposed here is consistent with a growing consensus on the need for, and the elements of, a quantitative approach to risk management of all major activities in an advanced industrial society. The ACNS recommends that the Atomic Energy Control Board adopt the proposed approach as a rational and coherent basis for nuclear power plant safety policy and requirements in Canada. (author). 68 refs., 4 tabs., 1 fig.

Quality assurance requirements for the computer software and safety analyses

International Nuclear Information System (INIS)

Husarecek, J.

1992-01-01

The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref

Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Eom, H.S.; Sung, T.Y.; Jeong, H.S.; Park, J.H.; Kang, H.G.; Lee, K

2001-03-01

As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software.

Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

International Nuclear Information System (INIS)

Eom, H. S.; Sung, T. Y.; Jeong, H. S.; Park, J. H.; Kang, H. G.; Lee, K.

2001-03-01

As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software

Quantitative assessment of safety barrier performance in the prevention of domino scenarios triggered by fire

International Nuclear Information System (INIS)

Landucci, Gabriele; Argenti, Francesca; Tugnoli, Alessandro; Cozzani, Valerio

2015-01-01

The evolution of domino scenarios triggered by fire critically depends on the presence and the performance of safety barriers that may have the potential to prevent escalation, delaying or avoiding the heat-up of secondary targets. The aim of the present study is the quantitative assessment of safety barrier performance in preventing the escalation of fired domino scenarios. A LOPA (layer of protection analysis) based methodology, aimed at the definition and quantification of safety barrier performance in the prevention of escalation was developed. Data on the more common types of safety barriers were obtained in order to characterize the effectiveness and probability of failure on demand of relevant safety barriers. The methodology was exemplified with a case study. The results obtained define a procedure for the estimation of safety barrier performance in the prevention of fire escalation in domino scenarios. - Highlights: • We developed a methodology for the quantitative assessment of safety barriers. • We focused on safety barriers aimed at preventing domino effect triggered by fire. • We obtained data on effectiveness and availability of the safety barriers. • The methodology was exemplified with a case study of industrial interest. • The results showed the role of safety barriers in preventing fired domino escalation

Qualitative and Quantitative Analyses of Glycogen in Human Milk.

Science.gov (United States)

Matsui-Yatsuhashi, Hiroko; Furuyashiki, Takashi; Takata, Hiroki; Ishida, Miyuki; Takumi, Hiroko; Kakutani, Ryo; Kamasaka, Hiroshi; Nagao, Saeko; Hirose, Junko; Kuriki, Takashi

2017-02-22

Identification as well as a detailed analysis of glycogen in human milk has not been shown yet. The present study confirmed that glycogen is contained in human milk by qualitative and quantitative analyses. High-performance anion exchange chromatography (HPAEC) and high-performance size exclusion chromatography with a multiangle laser light scattering detector (HPSEC-MALLS) were used for qualitative analysis of glycogen in human milk. Quantitative analysis was carried out by using samples obtained from the individual milks. The result revealed that the concentration of human milk glycogen varied depending on the mother's condition-such as the period postpartum and inflammation. The amounts of glycogen in human milk collected at 0 and 1-2 months postpartum were higher than in milk collected at 3-14 months postpartum. In the milk from mothers with severe mastitis, the concentration of glycogen was about 40 times higher than that in normal milk.

Sensitivity and uncertainty analyses applied to criticality safety validation. Volume 2

International Nuclear Information System (INIS)

Broadhead, B.L.; Hopper, C.M.; Parks, C.V.

1999-01-01

This report presents the application of sensitivity and uncertainty (S/U) analysis methodologies developed in Volume 1 to the code/data validation tasks of a criticality safety computational study. Sensitivity and uncertainty analysis methods were first developed for application to fast reactor studies in the 1970s. This work has revitalized and updated the existing S/U computational capabilities such that they can be used as prototypic modules of the SCALE code system, which contains criticality analysis tools currently in use by criticality safety practitioners. After complete development, simplified tools are expected to be released for general use. The methods for application of S/U and generalized linear-least-square methodology (GLLSM) tools to the criticality safety validation procedures were described in Volume 1 of this report. Volume 2 of this report presents the application of these procedures to the validation of criticality safety analyses supporting uranium operations where enrichments are greater than 5 wt %. Specifically, the traditional k eff trending analyses are compared with newly developed k eff trending procedures, utilizing the D and c k coefficients described in Volume 1. These newly developed procedures are applied to a family of postulated systems involving U(11)O 2 fuel, with H/X values ranging from 0--1,000. These analyses produced a series of guidance and recommendations for the general usage of these various techniques. Recommendations for future work are also detailed

Development of quantitative goals for inherent safety feature design and licensing

International Nuclear Information System (INIS)

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

1987-01-01

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

Concepts and examples of safety analyses for radioactive waste repositories in continental geological formations

International Nuclear Information System (INIS)

1983-01-01

This document is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of underground radioactive waste repositories. It is a companion to a general introductory document on the subject ''Safety Assessment for the Underground Disposal of Radioactive Wastes'', IAEA Safety Series No. 56, 1981, and reference to this earlier document will facilitate the reader's understanding of the present report. Since examples of safety analyses are summarized here, it is hoped that this document will contribute to providing a basis for a common understanding among authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing technical information, this document is also intended to stimulate further international discussion. The purposes of this report are: a) to identify the factors to be taken into account in radiological safety analyses of deep geological repositories, indicating as far as possible their relative importance during the various phases of system development; b) to show how these factors have been analysed in various safety assessment studies; and c) to comment on the merits of the selected and alternative approaches

Concepts and examples of safety analyses for radioactive waste repositories in continental geological formations

Energy Technology Data Exchange (ETDEWEB)

1983-01-01

This document is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of underground radioactive waste repositories. It is a companion to a general introductory document on the subject ''Safety Assessment for the Underground Disposal of Radioactive Wastes'', IAEA Safety Series No. 56, 1981, and reference to this earlier document will facilitate the reader's understanding of the present report. Since examples of safety analyses are summarized here, it is hoped that this document will contribute to providing a basis for a common understanding among authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing technical information, this document is also intended to stimulate further international discussion. The purposes of this report are: a) to identify the factors to be taken into account in radiological safety analyses of deep geological repositories, indicating as far as possible their relative importance during the various phases of system development; b) to show how these factors have been analysed in various safety assessment studies; and c) to comment on the merits of the selected and alternative approaches.

Best Estimate plus Uncertainty (BEPU) Analyses in the IAEA Safety Standards

International Nuclear Information System (INIS)

Dusic, Milorad; )

2013-01-01

The Safety Standards Series establishes an essential basis for safety and represents the broadest international consensus. Safety Standards Series publications are categorized into: Safety Fundamental (Present the overall objectives, concepts and principles of protection and safety, they are the policy documents of the safety standards), Safety Requirements (Establish requirements that must be met to ensure the protection and safety of people and the environment, both now and in the future), and Safety Guides (Provide guidance, in the form of more detailed actions, conditions or procedures that can be used to comply with the Requirements). The incorporation of more detailed requirements, in accordance with national practice, may still be necessary. There should be only one set of international safety standards. Each safety standard will be reviewed by the relevant committee or by the commission every five years. Best Estimate plus Uncertainty (BEPU) Analyses are approached in the following IAEA Safety Standards: - Safety Requirements SSR 2/1 - Safety of NPPs, Design (Revision of NS-R-1); - General Safety Requirement GSR Part 4: Safety Assessment for Facilities and Activities; - Safety Guide SSG-2 Deterministic Safety Analysis for Nuclear Power Plants. NUSSC suggested that new safety guides should be accompanied by documents like TECDOCs or Safety Reports describing in detail their recommendations where appropriate. Special review is currently underway to identify needs for revision in the light of the Fukushima accident. Revision will concern, first, the Safety Requirements, and then, the Selected Safety Guides

Quantitative numerical method for analysing slip traces observed by AFM

International Nuclear Information System (INIS)

Veselý, J; Cieslar, M; Coupeau, C; Bonneville, J

2013-01-01

Atomic force microscopy (AFM) is used more and more routinely to study, at the nanometre scale, the slip traces produced on the surface of deformed crystalline materials. Taking full advantage of the quantitative height data of the slip traces, which can be extracted from these observations, requires however an adequate and robust processing of the images. In this paper an original method is presented, which allows the fitting of AFM scan-lines with a specific parameterized step function without any averaging treatment of the original data. This yields a quantitative and full description of the changes in step shape along the slip trace. The strength of the proposed method is established on several typical examples met in plasticity by analysing nano-scale structures formed on the sample surface by emerging dislocations. (paper)

Joint analyses of open comments and quantitative data: Added value in a job satisfaction survey of hospital professionals.

Directory of Open Access Journals (Sweden)

Ingrid Gilles

Full Text Available To obtain a comprehensive understanding of the job opinions of hospital professionals by conducting qualitative analyses of the open comments included in a job satisfaction survey and combining these results with the quantitative results.A cross-sectional survey targeting all Lausanne University Hospital professionals was performed in the fall of 2013.The survey considered ten job satisfaction dimensions (e.g. self-fulfilment, workload, management, work-related burnout, organisational commitment, intent to stay and included an open comment section. Computer-assisted qualitative analyses were conducted on these comments. Satisfaction rates on the included dimensions and professional groups were entered as predictive variables in the qualitative analyses.Of 10 838 hospital professionals, 4978 participated in the survey and 1067 provided open comments. Data from 1045 respondents with usable comments constituted the analytic sample (133 physicians, 393 nurses, 135 laboratory technicians, 247 administrative staff, including researchers, 67 logistic staff, 44 psycho-social workers, and 26 unspecified.Almost a third of the comments addressed scheduling issues, mostly related to problems and exhaustion linked to shifts, work-life balance, and difficulties with colleagues' absences and the consequences for quality of care and patient safety. The other two-thirds related to classic themes included in job satisfaction surveys. Although some comments were provided equally by all professional groups, others were group specific: work and hierarchy pressures for physicians, healthcare quality and patient safety for nurses, skill recognition for administrative staff. Overall, respondents' comments were consistent with their job satisfaction ratings.Open comment analysis provides a comprehensive understanding of hospital professionals' job experiences, allowing better consideration of quality initiatives that match the needs of professionals with reality.

Joint analyses of open comments and quantitative data: Added value in a job satisfaction survey of hospital professionals.

Science.gov (United States)

Gilles, Ingrid; Mayer, Mauro; Courvoisier, Nelly; Peytremann-Bridevaux, Isabelle

2017-01-01

To obtain a comprehensive understanding of the job opinions of hospital professionals by conducting qualitative analyses of the open comments included in a job satisfaction survey and combining these results with the quantitative results. A cross-sectional survey targeting all Lausanne University Hospital professionals was performed in the fall of 2013. The survey considered ten job satisfaction dimensions (e.g. self-fulfilment, workload, management, work-related burnout, organisational commitment, intent to stay) and included an open comment section. Computer-assisted qualitative analyses were conducted on these comments. Satisfaction rates on the included dimensions and professional groups were entered as predictive variables in the qualitative analyses. Of 10 838 hospital professionals, 4978 participated in the survey and 1067 provided open comments. Data from 1045 respondents with usable comments constituted the analytic sample (133 physicians, 393 nurses, 135 laboratory technicians, 247 administrative staff, including researchers, 67 logistic staff, 44 psycho-social workers, and 26 unspecified). Almost a third of the comments addressed scheduling issues, mostly related to problems and exhaustion linked to shifts, work-life balance, and difficulties with colleagues' absences and the consequences for quality of care and patient safety. The other two-thirds related to classic themes included in job satisfaction surveys. Although some comments were provided equally by all professional groups, others were group specific: work and hierarchy pressures for physicians, healthcare quality and patient safety for nurses, skill recognition for administrative staff. Overall, respondents' comments were consistent with their job satisfaction ratings. Open comment analysis provides a comprehensive understanding of hospital professionals' job experiences, allowing better consideration of quality initiatives that match the needs of professionals with reality.

A methodology for a quantitative assessment of safety culture in NPPs based on Bayesian networks

International Nuclear Information System (INIS)

Kim, Young Gab; Lee, Seung Min; Seong, Poong Hyun

2017-01-01

Highlights: • A safety culture framework and a quantitative methodology to assess safety culture were proposed. • The relation among Norm system, Safety Management System and worker's awareness was established. • Safety culture probability at NPPs was updated by collecting actual organizational data. • Vulnerable areas and the relationship between safety culture and human error were confirmed. - Abstract: For a long time, safety has been recognized as a top priority in high-reliability industries such as aviation and nuclear power plants (NPPs). Establishing a safety culture requires a number of actions to enhance safety, one of which is changing the safety culture awareness of workers. The concept of safety culture in the nuclear power domain was established in the International Atomic Energy Agency (IAEA) safety series, wherein the importance of employee attitudes for maintaining organizational safety was emphasized. Safety culture assessment is a critical step in the process of enhancing safety culture. In this respect, assessment is focused on measuring the level of safety culture in an organization, and improving any weakness in the organization. However, many continue to think that the concept of safety culture is abstract and unclear. In addition, the results of safety culture assessments are mostly subjective and qualitative. Given the current situation, this paper suggests a quantitative methodology for safety culture assessments based on a Bayesian network. A proposed safety culture framework for NPPs would include the following: (1) a norm system, (2) a safety management system, (3) safety culture awareness of worker, and (4) Worker behavior. The level of safety culture awareness of workers at NPPs was reasoned through the proposed methodology. Then, areas of the organization that were vulnerable in terms of safety culture were derived by analyzing observational evidence. We also confirmed that the frequency of events involving human error

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)

Code development and analyses within the area of transmutation and safety

International Nuclear Information System (INIS)

Maschek, W.

2002-01-01

A strong code development is going on to meet various demands resulting from the development of dedicated reactors for transmutation and incineration. Code development is concerned with safety codes and general codes needed for assessing scenarios and transmutation strategies. Analyses concentrate on various ADS systems with solid and liquid molten salt fuels. Analyses deal with ADS Demo Plant (5th FP EU) and transmuters with advanced fuels

A quantitative method to analyse an open answer questionnaire: A case study about the Boltzmann Factor

International Nuclear Information System (INIS)

Battaglia, Onofrio Rosario; Di Paola, Benedetto

2015-01-01

This paper describes a quantitative method to analyse an openended questionnaire. Student responses to a specially designed written questionnaire are quantitatively analysed by not hierarchical clustering called k-means method. Through this we can characterise behaviour students with respect their expertise to formulate explanations for phenomena or processes and/or use a given model in the different context. The physics topic is about the Boltzmann Factor, which allows the students to have a unifying view of different phenomena in different contexts.

Criticality safety analyses in SKODA JS a.s

International Nuclear Information System (INIS)

Mikolas, P.; Svarny, J.

1999-01-01

This paper describes criticality safety analyses of spent fuel systems for storage and transport of spent fuel performed in SKODA JS s.r.o.. Analyses were performed for different systems both at NPP site including originally designed spent fuel pool with a large pitch between assemblies without any special absorbing material, high density spent fuel pool with an additional absorption by boron steel, depository rack for fresh fuel assemblies with a very large pitch between fuel assemblies, a container for transport of fresh fuel into the reactor pool and a cask for transport and storage of spent fuel and container for final storage depository. required subcriticality has been proven taking into account all possible unfavourable conditions, uncertainties etc. In two cases, burnup credit methodology is expected to be used. (Authors)

Integration of safety culture in transient analyses for nuclear power plants

International Nuclear Information System (INIS)

Stosic, Zoran V.; Stoll, Uwe

2009-01-01

In the nuclear field Safety Culture is the arrangement of attitudes and characteristics in individuals and organisations which determines first and foremost that nuclear power plant safety issues receive adequate attention due to their outstanding significance. It differs from general Corporate Culture via its concept of core hazards and the potentially large effects associated with the release of radioactivity. One can talk about positive and negative Safety Cultures. A positive Safety Culture assumes that the whole is more than the sum of the parts. The different parts interact to increase the overall effectiveness. In a negative Safety Culture the opposite is the case, with the action of some individuals restricted by the cynicism of others. Some examples of issues that contribute to a negative safety culture are: non-adherence to the established instructions and procedures, unclear definition of responsibilities, disinterest and inattentiveness, overestimation of own capabilities and arrogance, unclear rules, and mistrust between involved organisations. In addition to differentiation and importance of Safety Culture, necessary commitment levels, safety management framework, the paper discusses integration of Safety Culture in transient analyses of nuclear power plants. In this course the commitment to Safety Culture is defined as: a good Safety Culture depends on the continuous commitment and fulfilment of all involved organizations, persons and processes without any exception. (author)

Achieving reasonable conservatism in nuclear safety analyses

International Nuclear Information System (INIS)

Jamali, Kamiar

2015-01-01

In the absence of methods that explicitly account for uncertainties, seeking reasonable conservatism in nuclear safety analyses can quickly lead to extreme conservatism. The rate of divergence to extreme conservatism is often beyond the expert analysts’ intuitive feeling, but can be demonstrated mathematically. Too much conservatism in addressing the safety of nuclear facilities is not beneficial to society. Using certain properties of lognormal distributions for representation of input parameter uncertainties, example calculations for the risk and consequence of a fictitious facility accident scenario are presented. Results show that there are large differences between the calculated 95th percentiles and the extreme bounding values derived from using all input variables at their upper-bound estimates. Showing the relationship of the mean values to the key parameters of the output distributions, the paper concludes that the mean is the ideal candidate for representation of the value of an uncertain parameter. The mean value is proposed as the metric that is consistent with the concept of reasonable conservatism in nuclear safety analysis, because its value increases towards higher percentiles of the underlying positively skewed distribution with increasing levels of uncertainty. Insensitivity of the results to the actual underlying distributions is briefly demonstrated. - Highlights: • Multiple conservative assumptions can quickly diverge into extreme conservatism. • Mathematics and attractive properties provide basis for wide use of lognormal distribution. • Mean values are ideal candidates for representation of parameter uncertainties. • Mean values are proposed as reasonably conservative estimates of parameter uncertainties

Safety and deterministic failure analyses in high-beta D-D tokamak reactors

International Nuclear Information System (INIS)

Selcow, E.C.

1984-01-01

Safety and deterministic failure analyses were performed to compare major component failure characteristics for different high-beta D-D tokamak reactors. The primary focus was on evaluating damage to the reactor facility. The analyses also considered potential hazards to the general public and operational personnel. Parametric designs of high-beta D-D tokamak reactors were developed, using WILDCAT as the reference. The size, and toroidal field strength were reduced, and the fusion power increased in an independent manner. These changes were expected to improve the economics of D-D tokamaks. Issues examined using these designs were radiation induced failurs, radiation safety, first wall failure from plasma disruptions, and toroidal field magnet coil failure

Passive safety injection experiments and analyses (PAHKO)

International Nuclear Information System (INIS)

Tuunanen, J.

1998-01-01

PAHKO project involved experiments on the PACTEL facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines (Pressure Balancing Line, PBL, and Injection Line, IL). The examined PSIS worked efficiently in SBLOCAs although the flow through the PSIS stopped temporarily if the break was very small and the hot water filled the CMT. The experiments demonstrated the importance of the flow distributor in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable to simulate the overall behaviour of the transients. The detailed analyses of the results showed some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the PSIS phenomena. (orig.)

Safety design analyses of Korea Advanced Liquid Metal Reactor

International Nuclear Information System (INIS)

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

2000-01-01

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

Probabilistic safety analysis and interpretation thereof

International Nuclear Information System (INIS)

Steininger, U.; Sacher, H.

1999-01-01

Increasing use of the instrumentation of PSA is being made in Germany for quantitative technical safety assessment, for example with regard to incidents which must be reported and forwarding of information, especially in the case of modification of nuclear plants. The Commission for Nuclear Reactor Safety recommends regular execution of PSA on a cycle period of ten years. According to the PSA guidance instructions, probabilistic analyses serve for assessing the degree of safety of the entire plant, expressed as the expectation value for the frequency of endangering conditions. The authors describe the method, action sequence and evaluation of the probabilistic safety analyses. The limits of probabilistic safety analyses arise in the practical implementation. Normally the guidance instructions for PSA are confined to the safety systems, so that in practice they are at best suitable for operational optimisation only to a limited extent. The present restriction of the analyses has a similar effect on power output operation of the plant. This seriously degrades the utilitarian value of these analyses for the plant operators. In order to further develop PSA as a supervisory and operational optimisation instrument, both authors consider it to be appropriate to bring together the specific know-how of analysts, manufacturers, plant operators and experts. (orig.) [de

Process hazards analysis (PrHA) program, bridging accident analyses and operational safety

International Nuclear Information System (INIS)

Richardson, J.A.; McKernan, S.A.; Vigil, M.J.

2003-01-01

Recently the Final Safety Analysis Report (FSAR) for the Plutonium Facility at Los Alamos National Laboratory, Technical Area 55 (TA-55) was revised and submitted to the US. Department of Energy (DOE). As a part of this effort, over seventy Process Hazards Analyses (PrHAs) were written and/or revised over the six years prior to the FSAR revision. TA-55 is a research, development, and production nuclear facility that primarily supports US. defense and space programs. Nuclear fuels and material research; material recovery, refining and analyses; and the casting, machining and fabrication of plutonium components are some of the activities conducted at TA-35. These operations involve a wide variety of industrial, chemical and nuclear hazards. Operational personnel along with safety analysts work as a team to prepare the PrHA. PrHAs describe the process; identi fy the hazards; and analyze hazards including determining hazard scenarios, their likelihood, and consequences. In addition, the interaction of the process to facility systems, structures and operational specific protective features are part of the PrHA. This information is rolled-up to determine bounding accidents and mitigating systems and structures. Further detailed accident analysis is performed for the bounding accidents and included in the FSAR. The FSAR is part of the Documented Safety Analysis (DSA) that defines the safety envelope for all facility operations in order to protect the worker, the public, and the environment. The DSA is in compliance with the US. Code of Federal Regulations, 10 CFR 830, Nuclear Safety Management and is approved by DOE. The DSA sets forth the bounding conditions necessary for the safe operation for the facility and is essentially a 'license to operate.' Safely of day-to-day operations is based on Hazard Control Plans (HCPs). Hazards are initially identified in the PrI-IA for the specific operation and act as input to the HCP. Specific protective features important to worker

PRA and the implementation of quantitative safety goals

International Nuclear Information System (INIS)

Okrent, D.

1983-01-01

With the adoption by the U.S. Nuclear Regulatory Commission (NRC) in January, 1983, of a Policy Statement on Safety Goals for the Operation of Nuclear Power Plants, probabilitstic risk assessment (PRA) has taken on increased importance in nuclear reactor safety. Although the Reactor Safety Study, WASH-1400, was a major pioneering effort that revolutionized thinking about reactor safety, PRA was used only on occasion by the NRC regulatory staff prior to the accident at Three Mile Island. Since then, PRA has been used more and more as an important factor in decision making, usually for specific issues. The nuclear industry has also employed PRA, sometimes to make its case on specific issues, sometimes to present a position on overall risk. The advent of the Zion and Indian Point PRAs, with their treatment of risks from fire, wind, and earthquakes, and their examination of the course of core melt accidents, has added a new dimension to the overall picture. Although the NRC has stated that during the next two year evolution period, its quantitative design objectives and PRA are not to enter directly into the licensing process, many important issues will be influenced significantly by the results of risk and reliability studies. In fact, PRA may be coming into a position of great importance before the methodology, data, and process are sufficiently mature for the task. Large gaps still exist in our understanding of phenomena and in input information; and much of the final result depends on subjective input; large differences of opinion can and should be expected to persist. Accepted standards for quality assurance, and adequacy and depth of independent, peer review remain to be formulated and achieved. This paper will summarize the recently adopted NRC safety policy and the two-year evaluation plan, and will provide, by example, some words of caution concerning a few of the difficulties which may arise. (orig.)

Analysing context-dependent deviations in interacting with safety-critical systems

International Nuclear Information System (INIS)

Paterno, Fabio; Santoro, Carmen

2006-01-01

Mobile technology is penetrating many areas of human life. This implies that the context of use can vary in many respects. We present a method that aims to support designers in managing the complex design space when considering applications with varying contexts and help them to identify solutions that support users in performing their activities while preserving usability and safety. The method is a novel combination of an analysis of both potential deviations in task performance and most suitable information representations based on distributed cognition. The originality of the contribution is in providing a conceptual tool for better understanding the impact of context of use on user interaction in safety-critical domains. In order to present our approach we provide an example in which the implications of introducing new support through mobile devices in a safety-critical system are identified and analysed in terms of potential hazards

75 FR 29537 - Draft Transportation Conformity Guidance for Quantitative Hot-spot Analyses in PM2.5

Science.gov (United States)

2010-05-26

... Quantitative Hot- spot Analyses in PM 2.5 and PM 10 Nonattainment and Maintenance Areas AGENCY: Environmental... finalized, this guidance would help state and local agencies complete quantitative PM 2.5 and PM 10 hot-spot... projects. A hot-spot analysis includes an estimation of project-level emissions, air quality modeling, and...

Building patient safety in intensive care nursing : Patient safety culture, team performance and simulation-based training

OpenAIRE

Ballangrud, Randi

2013-01-01

Aim: The overall aim of the thesis was to investigate patient safety culture, team performance and the use of simulation-based team training for building patient safety in intensive care nursing. Methods: Quantitative and qualitative methods were used. In Study I, 220 RNs from ten ICUs responded to a patient safety culture questionnaire analysed with statistics. Studies II-IV were based on an evaluation of a simulation-based team training programme. Studies II-III included 53 RNs from seven I...

Operation safety of control systems. Principles and methods

International Nuclear Information System (INIS)

Aubry, J.F.; Chatelet, E.

2008-01-01

This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)

NPP Krsko periodic safety review. Safety assessment and analyses

International Nuclear Information System (INIS)

Basic, I.; Spiler, J.; Thaulez, F.

2002-01-01

Definition of a PSR (Periodic Safety Review) project is a comprehensive safety review of a plant after ten years of operation. The objective is a verification by means of a comprehensive review using current methods that the plant remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. The overall goals of the NEK PSR Program are defined in compliance with the basic role of a PSR and the current practice typical for most of the countries in EU. This practice is described in the related guides and good practice documents issued by international organizations. The overall goals of the NEK PSR are formulated as follows: to demonstrate that the plant is as safe as originally intended; to evaluate the actual plant status with respect to aging and wear-out identifying any structures, systems or components that could limit the life of the plant in the foreseeable future, and to identify appropriate corrective actions, where needed; to compare current level of safety in the light of modern standards and knowledge, and to identify where improvements would be beneficial for minimizing deviations at justifiable costs. The Krsko PSR will address the following safety factors: Operational Experience, Safety Assessment, EQ and Aging Management, Safety Culture, Emergency Planning, Environmental Impact and Radioactive Waste.(author)

The impact of safety analyses on the design of the Hanford Waste Vitrification Plant

International Nuclear Information System (INIS)

Koppenaal, T.J.; Yee, A.K.; Reisdorf, J.B.; Hall, B.W.

1993-04-01

Accident analyses are being performed to evaluate and document the safety of the Hanford Waste Vitrification Plant (HWVP). The safety of the HWVP is assessed by evaluating worst-case accident scenarios and determining the dose to offsite and onsite receptors. Air dispersion modeling is done with the GENII computer code. Three accidents are summarized in this paper, and their effects on the safety and the design of the HWVP are demonstrated

Use of probabilistic safety analyses in severe accident management

International Nuclear Information System (INIS)

Neogy, P.; Lehner, J.

1991-01-01

An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe accident management are explored using ex-vessel accident management at a PWR ice-condenser plant as an example. 2 refs., 1 fig., 3 tabs

A quantitative approach for risk-informed safety significance categorization in option-2

International Nuclear Information System (INIS)

Ha, Jun Su; Seong, Poong Hyun

2004-01-01

OPTION-2 recommends that Structures, Systems, or Components (SSCs) of Nuclear Power Plants (NPPs) should be categorized into four groups according to their safety significance as well as whether they are safety-related or not. With changes to the scope of SSCs covered by 10 CFR 50, safety-related components which categorized into low safety significant SSC (RISC-3 SSC) can be exempted from the existing conservative burden (or requirements). As OPTION-2 paradigm is applied, a lot of SSCs may be categorized into RISC-3 SSCs. Changes in treatment of the RISC-3 SSCs will be recommended and then finally the recommended changes shall be evaluated. Consequently, before recommending the changes in treatment, probable candidate SSCs for the changes in treatment need to be identified for efficient risk-informed regulation and application (RIRA). Hence, in this work, a validation focused on the RISC-3 SSCs is proposed to identify probable candidate SSCs. Burden to Importance Ratio (BIR) is utilized as a quantitative measure for the validation. BIR is a measure representing the extent of resources or requirements imposed on a SSC with respect to the value of the importance measure of the SSC. Therefore SSCs having high BIR can be considered as probable candidate SSCs for the changes in treatment. In addition, the final decision whether RISC-3 SSCs can be considered as probable candidate SSCs or not should be made by an expert panel. For the effective decision making, a structured mathematical decision-making process is constructed based on Belief Networks (BBN) to overcome demerits of conventional group meeting based on unstructured discussion for decision-making. To demonstrate the usefulness of the proposed approach, the approach is applied to 22 components selected from 512 In-Service Test (IST) components of Ulchin unit 3. The results of the application show that the proposed approach can identify probable candidate SSCs for changes in treatment. The identification of the

Safety analyses for NHR-200

Energy Technology Data Exchange (ETDEWEB)

Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

1997-09-01

The NHR-200 is a commercial 200-MW District Heating Reactor developed in China. It is designed on the basis of design, construction and four-year operating experience of the 5MW Experimental Heating Reactor (NHR-5). It has special safety features which are briefly described in this paper. Accident classification and safety criteria are also explained. Some typical and serious accidents are studied theoretically, and their results are detailed in this paper. They demonstrate the excellent safety characteristics of HR-200. (author). 4 refs, 9 figs, 1 tab.

Seismic and tsunami safety margin assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Seismic and tsunami safety margin assessment

International Nuclear Information System (INIS)

2013-01-01

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

Use of the deterministic safety analyses in support to the NPP Krsko modification

International Nuclear Information System (INIS)

Feretic, D.; Cavlina, N.; Debrecin, N.; Grgic, D.; Bajs, T.; Spalj, S.

2004-01-01

The ultimate goal of the safety analysis is to verify that Nuclear Power Plant (NPP) meets safety and operational requirements. To this aim it is necessary to demonstrate that plant safety has not been deteriorated in the case of the modifications to the plant Systems, Structures and Components (SSC) or changes to the plant procedures. In addition, safety analyses are needed in the case of reassessment of an existing plant. The reasons for reassessment may be different, e.g. due to the changes in the methodology and assumptions used in the original design, if the original design basis or acceptance criteria may no longer be adequate, if the safety analysis tools used may have been superseded by more sophisticated methods or if the original design basis may no longer be met. The operation of the NPP Krsko has experienced numerous changes from the original design for the majority of the reasons that have been mentioned before. On the other side, the application of the large best-estimate thermalhydraulic codes has evolved to the wide spread support in the operation of the NPP: compliance with the regulatory goals, support to the PSA studies, analysis of the operational transients, plant modifications studies, equipment qualification, training of the operators, preparation of the operating procedures, etc. This trend has been followed at the Faculty of Electrical Engineering Zagreb (FER) and applied to the on-going needs due to the modifications and changes at NPP Krsko. In this paper, an overview of the deterministic safety analyses performed at FER in the support to the NPP Krsko modifications and changes is presented.(author)

The role of safety analyses in site selection. Some personal observations based on the experience from the Swiss site selection process

Energy Technology Data Exchange (ETDEWEB)

Zuidema, Piet [Nagra, Wettingen (Switzerland)

2015-07-01

geological barrier (host rock and confining units); long-term stability (erosion, differential movements, etc.); reliability of geological information (explorability; predictability); technical feasibility (sufficient space for allocating the disposal rooms; depth of repository; rock strength, etc.). For some of these issues, rather detailed quantitative analyses are made (e.g. for erosion). Besides long-term safety, also operational safety is considered. This is done to ensure that suitable sites are chosen for the surface infrastructure (waste acceptance facilities, entrance to access to underground). The main emphasis is on external events (e.g. very severe flooding) that need to be avoided. The involvement of society in the site selection process is also very important. This requires that the scientific information needed (and wanted) by society is delivered in a format understandable to them. This helps society develop an understanding of the question ''why here and not there'' in the siting decision; something that is considered essential to get the necessary support for the siting decision.

The role of safety analyses in site selection. Some personal observations based on the experience from the Swiss site selection process

International Nuclear Information System (INIS)

Zuidema, Piet

2015-01-01

geological barrier (host rock and confining units); long-term stability (erosion, differential movements, etc.); reliability of geological information (explorability; predictability); technical feasibility (sufficient space for allocating the disposal rooms; depth of repository; rock strength, etc.). For some of these issues, rather detailed quantitative analyses are made (e.g. for erosion). Besides long-term safety, also operational safety is considered. This is done to ensure that suitable sites are chosen for the surface infrastructure (waste acceptance facilities, entrance to access to underground). The main emphasis is on external events (e.g. very severe flooding) that need to be avoided. The involvement of society in the site selection process is also very important. This requires that the scientific information needed (and wanted) by society is delivered in a format understandable to them. This helps society develop an understanding of the question ''why here and not there'' in the siting decision; something that is considered essential to get the necessary support for the siting decision.

Cost/benefit analyses of reactor safety systems

International Nuclear Information System (INIS)

1988-01-01

The study presents a methodology for quantitative assessment of the benefit yielded by the various engineered safety systems of a nuclear reactor containment from the standpoint of their capacity to protect the environment compared to their construction costs. The benefit is derived from an estimate of the possible damage from which the environment is protected, taking account of the probabilities of occurrence of malfunctions and accidents. For demonstration purposes, the methodology was applied to a 1 300-MWe PWR nuclear power station. The accident sequence considered was that of a major loss-of-coolant accident as investigated in detail in the German risk study. After determination of the benefits and cost/benefit ratio for the power plant and the containment systems as designed, the performance characteristics of three subsystems, the leakoff system, annulus exhaust air handling system and spray system, were varied. For this purpose, the parameters which describe these systems in the activity release programme were altered. The costs were simultaneously altered in order to take account of the performance divergences. By varying the performance of the individual sub-systems an optimization in design of these systems can be arrived at

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Matsuoka, Takeshi; Yang Ming

2014-01-01

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR. For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. (author)

A study on the quantitative evaluation of the reliability for safety critical software using Bayesian belief nets

International Nuclear Information System (INIS)

Eom, H. S.; Jang, S. C.; Ha, J. J.

2003-01-01

Despite the efforts to avoid undesirable risks, or at least to bring them under control in the world, new risks that are highly difficult to manage continue to emerge from the use of new technologies, such as the use of digital instrumentation and control (I and C) components in nuclear power plant. Whenever new risk issues came out by now, we have endeavored to find the most effective ways to reduce risks, or to allocate limited resources to do this. One of the major challenges is the reliability analysis of safety-critical software associated with digital safety systems. Though many activities such as testing, verification and validation (V and V) techniques have been carried out in the design stage of software, however, the process of quantitatively evaluating the reliability of safety-critical software has not yet been developed because of the irrelevance of the conventional software reliability techniques to apply for the digital safety systems. This paper focuses on the applicability of Bayesian Belief Net (BBN) techniques to quantitatively estimate the reliability of safety-critical software adopted in digital safety system. In this paper, a typical BBN model was constructed using the dedication process of the Commercial-Off-The-Shelf (COTS) installed by KAERI. In conclusion, the adoption of BBN technique can facilitate the process of evaluating the safety-critical software reliability in nuclear power plant, as well as provide very useful information (e.g., 'what if' analysis) associated with software reliability in the viewpoint of practicality

Flightdeck Automation Problems (FLAP) Model for Safety Technology Portfolio Assessment

Science.gov (United States)

Ancel, Ersin; Shih, Ann T.

2014-01-01

NASA's Aviation Safety Program (AvSP) develops and advances methodologies and technologies to improve air transportation safety. The Safety Analysis and Integration Team (SAIT) conducts a safety technology portfolio assessment (PA) to analyze the program content, to examine the benefits and risks of products with respect to program goals, and to support programmatic decision making. The PA process includes systematic identification of current and future safety risks as well as tracking several quantitative and qualitative metrics to ensure the program goals are addressing prominent safety risks accurately and effectively. One of the metrics within the PA process involves using quantitative aviation safety models to gauge the impact of the safety products. This paper demonstrates the role of aviation safety modeling by providing model outputs and evaluating a sample of portfolio elements using the Flightdeck Automation Problems (FLAP) model. The model enables not only ranking of the quantitative relative risk reduction impact of all portfolio elements, but also highlighting the areas with high potential impact via sensitivity and gap analyses in support of the program office. Although the model outputs are preliminary and products are notional, the process shown in this paper is essential to a comprehensive PA of NASA's safety products in the current program and future programs/projects.

Towards an Industrial Application of Statistical Uncertainty Analysis Methods to Multi-physical Modelling and Safety Analyses

International Nuclear Information System (INIS)

Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

2013-01-01

Since 1980's, Tractebel Engineering (TE) has being developed and applied a multi-physical modelling and safety analyses capability, based on a code package consisting of the best estimate 3D neutronic (PANTHER), system thermal hydraulic (RELAP5), core sub-channel thermal hydraulic (COBRA-3C), and fuel thermal mechanic (FRAPCON/FRAPTRAN) codes. A series of methodologies have been developed to perform and to license the reactor safety analysis and core reload design, based on the deterministic bounding approach. Following the recent trends in research and development as well as in industrial applications, TE has been working since 2010 towards the application of the statistical sensitivity and uncertainty analysis methods to the multi-physical modelling and licensing safety analyses. In this paper, the TE multi-physical modelling and safety analyses capability is first described, followed by the proposed TE best estimate plus statistical uncertainty analysis method (BESUAM). The chosen statistical sensitivity and uncertainty analysis methods (non-parametric order statistic method or bootstrap) and tool (DAKOTA) are then presented, followed by some preliminary results of their applications to FRAPCON/FRAPTRAN simulation of OECD RIA fuel rod codes benchmark and RELAP5/MOD3.3 simulation of THTF tests. (authors)

Deep Borehole Disposal Safety Analysis.

Energy Technology Data Exchange (ETDEWEB)

Freeze, Geoffrey A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); MacKinnon, Robert J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Tillman, Jack Bruce [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2016-10-01

This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-08-01

There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (²³⁸Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (²³⁷Np) dioxide (NpO₂)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs

Assessing the validity of road safety evaluation studies by analysing causal chains.

Science.gov (United States)

Elvik, Rune

2003-09-01

This paper discusses how the validity of road safety evaluation studies can be assessed by analysing causal chains. A causal chain denotes the path through which a road safety measure influences the number of accidents. Two cases are examined. One involves chemical de-icing of roads (salting). The intended causal chain of this measure is: spread of salt --> removal of snow and ice from the road surface --> improved friction --> shorter stopping distance --> fewer accidents. A Norwegian study that evaluated the effects of salting on accident rate provides information that describes this causal chain. This information indicates that the study overestimated the effect of salting on accident rate, and suggests that this estimate is influenced by confounding variables the study did not control for. The other case involves a traffic club for children. The intended causal chain in this study was: join the club --> improve knowledge --> improve behaviour --> reduce accident rate. In this case, results are rather messy, which suggests that the observed difference in accident rate between members and non-members of the traffic club is not primarily attributable to membership in the club. The two cases show that by analysing causal chains, one may uncover confounding factors that were not adequately controlled in a study. Lack of control for confounding factors remains the most serious threat to the validity of road safety evaluation studies.

Quantitative methods for analysing cumulative effects on fish migration success: a review.

Science.gov (United States)

Johnson, J E; Patterson, D A; Martins, E G; Cooke, S J; Hinch, S G

2012-07-01

It is often recognized, but seldom addressed, that a quantitative assessment of the cumulative effects, both additive and non-additive, of multiple stressors on fish survival would provide a more realistic representation of the factors that influence fish migration. This review presents a compilation of analytical methods applied to a well-studied fish migration, a more general review of quantitative multivariable methods, and a synthesis on how to apply new analytical techniques in fish migration studies. A compilation of adult migration papers from Fraser River sockeye salmon Oncorhynchus nerka revealed a limited number of multivariable methods being applied and the sub-optimal reliance on univariable methods for multivariable problems. The literature review of fisheries science, general biology and medicine identified a large number of alternative methods for dealing with cumulative effects, with a limited number of techniques being used in fish migration studies. An evaluation of the different methods revealed that certain classes of multivariable analyses will probably prove useful in future assessments of cumulative effects on fish migration. This overview and evaluation of quantitative methods gathered from the disparate fields should serve as a primer for anyone seeking to quantify cumulative effects on fish migration survival. © 2012 The Authors. Journal of Fish Biology © 2012 The Fisheries Society of the British Isles.

Development of a quantitative safety assessment method for nuclear I and C systems including human operators

International Nuclear Information System (INIS)

Kim, Man Cheol

2004-02-01

Conventional PSA (probabilistic safety analysis) is performed in the framework of event tree analysis and fault tree analysis. In conventional PSA, I and C systems and human operators are assumed to be independent for simplicity. But, the dependency of human operators on I and C systems and the dependency of I and C systems on human operators are gradually recognized to be significant. I believe that it is time to consider the interdependency between I and C systems and human operators in the framework of PSA. But, unfortunately it seems that we do not have appropriate methods for incorporating the interdependency between I and C systems and human operators in the framework of Pasa. Conventional human reliability analysis (HRA) methods are not developed to consider the interdependecy, and the modeling of the interdependency using conventional event tree analysis and fault tree analysis seem to be, event though is does not seem to be impossible, quite complex. To incorporate the interdependency between I and C systems and human operators, we need a new method for HRA and a new method for modeling the I and C systems, man-machine interface (MMI), and human operators for quantitative safety assessment. As a new method for modeling the I and C systems, MMI and human operators, I develop a new system reliability analysis method, reliability graph with general gates (RGGG), which can substitute conventional fault tree analysis. RGGG is an intuitive and easy-to-use method for system reliability analysis, while as powerful as conventional fault tree analysis. To demonstrate the usefulness of the RGGG method, it is applied to the reliability analysis of Digital Plant Protection System (DPPS), which is the actual plant protection system of Ulchin 5 and 6 nuclear power plants located in Republic of Korea. The latest version of the fault tree for DPPS, which is developed by the Integrated Safety Assessment team in Korea Atomic Energy Research Institute (KAERI), consists of 64

Safety culture and learning from incidents: the role of incident reporting and causal analyses

International Nuclear Information System (INIS)

Wilpert, B.

1994-01-01

Nuclear industry more than any other industrial branch has developed and used predictive risk analysis as a method of feedforward control of safety and reliability. Systematic evaluation of operating experience, statistical documentation of component failures, systematic documentation and analysis of incidents are important complementary elements of feedback control: we are dealing here with adjustment and learning from experience, in particular from past incidents. Using preliminary findings from ongoing research at the Research Center Systems Safety at the Berlin University of Technology the contribution discusses preconditions for an effective use of lessons to be learnt from closely matched incident reporting and in depth analyses of causal chains leading to incidents. Such conditions are especially standardized documentation, reporting and analyzing methods of incidents; structured information flows and feedback loops; abstaining from culpability search; mutual trust of employees and management; willingness of all concerned to continually evaluate and optimize the established learning system. Thus, incident related reporting and causal analyses contribute to safety culture, which is seen to emerge from tightly coupled organizational measures and respective change in attitudes and behaviour. (author) 2 figs., 7 refs

Quantitative Model for Economic Analyses of Information Security Investment in an Enterprise Information System

Directory of Open Access Journals (Sweden)

Bojanc Rok

2012-11-01

Full Text Available The paper presents a mathematical model for the optimal security-technology investment evaluation and decision-making processes based on the quantitative analysis of security risks and digital asset assessments in an enterprise. The model makes use of the quantitative analysis of different security measures that counteract individual risks by identifying the information system processes in an enterprise and the potential threats. The model comprises the target security levels for all identified business processes and the probability of a security accident together with the possible loss the enterprise may suffer. The selection of security technology is based on the efficiency of selected security measures. Economic metrics are applied for the efficiency assessment and comparative analysis of different protection technologies. Unlike the existing models for evaluation of the security investment, the proposed model allows direct comparison and quantitative assessment of different security measures. The model allows deep analyses and computations providing quantitative assessments of different options for investments, which translate into recommendations facilitating the selection of the best solution and the decision-making thereof. The model was tested using empirical examples with data from real business environment.

Safety analyses for an in-pile SCWR fuel qualification test loop

Energy Technology Data Exchange (ETDEWEB)

Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

2014-07-01

As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

Qualitative and quantitative reliability analysis of safety systems

International Nuclear Information System (INIS)

Karimi, R.; Rasmussen, N.; Wolf, L.

1980-05-01

A code has been developed for the comprehensive analysis of a fault tree. The code designated UNRAC (UNReliability Analysis Code) calculates the following characteristics of an input fault tree: (1) minimal cut sets; (2) top event unavailability as point estimate and/or in time dependent form; (3) quantitative importance of each component involved; and, (4) error bound on the top event unavailability. UNRAC can analyze fault trees, with any kind of gates (EOR, NAND, NOR, AND, OR), up to a maximum of 250 components and/or gates. The code is benchmarked against WAMCUT, MODCUT, KITT, BIT-FRANTIC, and PL-MODT. The results showed that UNRAC produces results more consistent with the KITT results than either BIT-FRANTIC or PL-MODT. Overall it is demonstrated that UNRAC is an efficient easy-to-use code and has the advantage of being able to do a complete fault tree analysis with this single code. Applications of fault tree analysis to safety studies of nuclear reactors are considered

Quantitative safety assessment of air traffic control systems through system control capacity

Science.gov (United States)

Guo, Jingjing

Quantitative Safety Assessments (QSA) are essential to safety benefit verification and regulations of developmental changes in safety critical systems like the Air Traffic Control (ATC) systems. Effectiveness of the assessments is particularly desirable today in the safe implementations of revolutionary ATC overhauls like NextGen and SESAR. QSA of ATC systems are however challenged by system complexity and lack of accident data. Extending from the idea "safety is a control problem" in the literature, this research proposes to assess system safety from the control perspective, through quantifying a system's "control capacity". A system's safety performance correlates to this "control capacity" in the control of "safety critical processes". To examine this idea in QSA of the ATC systems, a Control-capacity Based Safety Assessment Framework (CBSAF) is developed which includes two control capacity metrics and a procedural method. The two metrics are Probabilistic System Control-capacity (PSC) and Temporal System Control-capacity (TSC); each addresses an aspect of a system's control capacity. And the procedural method consists three general stages: I) identification of safety critical processes, II) development of system control models and III) evaluation of system control capacity. The CBSAF was tested in two case studies. The first one assesses an en-route collision avoidance scenario and compares three hypothetical configurations. The CBSAF was able to capture the uncoordinated behavior between two means of control, as was observed in a historic midair collision accident. The second case study compares CBSAF with an existing risk based QSA method in assessing the safety benefits of introducing a runway incursion alert system. Similar conclusions are reached between the two methods, while the CBSAF has the advantage of simplicity and provides a new control-based perspective and interpretation to the assessments. The case studies are intended to investigate the

ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

WILLIAMS, J.C.

2003-11-15

This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

International Nuclear Information System (INIS)

Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

2005-01-01

It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted in a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident postulated to occur during transportation, for the purpose of gaining acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and thus, accident conditions leading to mechanical damages and thermal failure were determined to characterize the scenarios. As a result, the worst-case conditions of run-off-the-road accidents were set up to define the impact against a concrete or asphalt surface. For fire accident scenarios to be set up, collisions were assumed to occur with an oil tanker carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside a tunnel without ventilation. Then the cask models were determined for these safety demonstration analyses to represent those commonly used for fresh nuclear fuel transported throughout Japan. Following the postulated accident scenarios, the mechanical damages were analyzed by using the general-purpose finite element code LS-DYNA with three-dimensional elements. It was found that leak tightness of the package be maintained even in the severe impact scenario. Then the thermal safety was analyzed by using the general-purpose finite element code ABAOUS with three-dimensional elements to describe cask geometry. As a result of the thermal analyses, the integrity of the containment

Expermental Studies of quantitative evaluation using HPLC and safety of Sweet Bee Venom

OpenAIRE

Ki Rok Kwon; Ching Seng Chu; Hee Soo Park; Min Ki Kim; Bae Chun Cha; Eun Lee

2007-01-01

Objectives : This study was conducted to carry out quantitative evaluation and safety of Sweet Bee Venom. Methods : Content analysis was done using HPLC, measurement of LD50 was conducted intravenous, subcutaneous, and intra-muscular injection to the ICR mice. Results : 1. According to HPLC analysis, removal of the enzymes containing phospholipase A2 was successfully rendered on Sweet Bee Venom. And analyzing melittin content, Sweet Bee Venom contained 12% more melittin than Bee Venom. ...

Systematic review of economic analyses in patient safety: a protocol designed to measure development in the scope and quality of evidence.

Science.gov (United States)

Carter, Alexander W; Mandavia, Rishi; Mayer, Erik; Marti, Joachim; Mossialos, Elias; Darzi, Ara

2017-08-18

Recent avoidable failures in patient care highlight the ongoing need for evidence to support improvements in patient safety. According to the most recent reviews, there is a dearth of economic evidence related to patient safety. These reviews characterise an evidence gap in terms of the scope and quality of evidence available to support resource allocation decisions. This protocol is designed to update and improve on the reviews previously conducted to determine the extent of methodological progress in economic analyses in patient safety. A broad search strategy with two core themes for original research (excluding opinion pieces and systematic reviews) in 'patient safety' and 'economic analyses' has been developed. Medline, Econlit and National Health Service Economic Evaluation Database bibliographic databases will be searched from January 2007 using a combination of medical subject headings terms and research-derived search terms (see table 1). The method is informed by previous reviews on this topic, published in 2012. Screening, risk of bias assessment (using the Cochrane collaboration tool) and economic evaluation quality assessment (using the Drummond checklist) will be conducted by two independent reviewers, with arbitration by a third reviewer as needed. Studies with a low risk of bias will be assessed using the Drummond checklist. High-quality economic evaluations are those that score >20/35. A qualitative synthesis of evidence will be performed using a data collection tool to capture the study design(s) employed, population(s), setting(s), disease area(s), intervention(s) and outcome(s) studied. Methodological quality scores will be compared with previous reviews where possible. Effect size(s) and estimate uncertainty will be captured and used in a quantitative synthesis of high-quality evidence, where possible. Formal ethical approval is not required as primary data will not be collected. The results will be disseminated through a peer

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

International Nuclear Information System (INIS)

Hedin, A.

2008-01-01

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

The long-term safety and performance analyses of the surface disposal facility for the Belgian category a waste at Dessel

Energy Technology Data Exchange (ETDEWEB)

Cool, Wim; Vermarien, Elise; Wacquier, William [ONDRAF/NIRAS Avenue des Arts 14, BE-1210 Bruxelles (Belgium); Perko, Janez [SCK-CEN Boeretang 200, BE-2400 Mol (Belgium)

2013-07-01

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium. This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology. An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal. This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified. (authors)

Japanese standard method for safety evaluation using best estimate code based on uncertainty and scaling analyses with statistical approach

International Nuclear Information System (INIS)

Mizokami, Shinya; Hotta, Akitoshi; Kudo, Yoshiro; Yonehara, Tadashi; Watada, Masayuki; Sakaba, Hiroshi

2009-01-01

Current licensing practice in Japan consists of using conservative boundary and initial conditions(BIC), assumptions and analytical codes. The safety analyses for licensing purpose are inherently deterministic. Therefore, conservative BIC and assumptions, such as single failure, must be employed for the analyses. However, using conservative analytical codes are not considered essential. The standard committee of Atomic Energy Society of Japan(AESJ) has drawn up the standard for using best estimate codes for safety analyses in 2008 after three-years of discussions reflecting domestic and international recent findings. (author)

Determination of correction coefficients for quantitative analysis by mass spectrometry. Application to uranium impurities analysis; Recherche des coefficients de correction permettant l'analyse quantitative par spectrometrie de masse. Application a l'analyse d'impuretes dans l'uranium

Energy Technology Data Exchange (ETDEWEB)

Billon, J P [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes

1970-07-01

Some of basic principles in spark source mass spectrometry are recalled. It is shown how this method can lead to quantitative analysis when attention is paid to some theoretical aspects. A time constant relation being assumed between the analysed solid sample and the ionic beam it gives we determined experimental relative sensitivity factors for impurities in uranium matrix. Results being in fairly good agreement with: an unelaborate theory on ionization yield in spark-source use of theoretically obtained relative sensitivity factors in uranium matrix has been developed. (author) [French] Apres avoir rappele quelques principes fondamentaux regissant la spectrometrie de masse a etincelles, nous avons montre que moyennant un certain nombre de precautions, il etait possible d'utiliser cette methode en analyse quantitative. Ayant admis qu'il existait une relation constante dans le temps entre l'echantillon solide analyse et le faisceau ionique qui en est issu, nous avons d'abord entrepris de determiner des coefficients de correction experimentaux pour des matrices d'uranium. Les premiers resultats pratiques semblant en accord avec une theorie simple relative au rendement d'ionisation dans la source a etincelles, nous avons etudie la possibilite d'appliquer directement les coefficients theoriques ainsi definis, l'application etant toujours faite sur des matrices d'uranium. (auteur)

Dry critical experiments and analyses performed in support of the Topaz-2 Safety Program

International Nuclear Information System (INIS)

Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Loynstev, V.A.

1994-01-01

In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations

Dry critical experiments and analyses performed in support of the TOPAZ-2 safety program

International Nuclear Information System (INIS)

Pelowitz, D.B.; Sapir, J.; Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Kompanietz, G.B.; Krutov, A.M.; Polyakov, D.N.; Lobynstev, V.A.

1995-01-01

In December 1991, the Strategic Defense Initiative Organization decided to investigate the possibility of launching a Russian Topaz-2 space nuclear power system. Functional safety requirements developed for the Topaz mission mandated that the reactor remain subcritical when flooded and immersed in water. Initial experiments and analyses performed in Russia and the United States indicated that the reactor could potentially become supercritical in several water- or sand-immersion scenarios. Consequently, a series of critical experiments was performed on the Narciss M-II facility at the Kurchatov Institute to measure the reactivity effects of water and sand immersion, to quantify the effectiveness of reactor modifications proposed to preclude criticality, and to benchmark the calculational methods and nuclear data used in the Topaz-2 safety analyses. In this paper we describe the Narciss M-II experimental configurations along with the associated calculational models and methods. We also present and compare the measured and calculated results for the dry experimental configurations. copyright 1995 American Institute of Physics

Risk analyses in nuclear engineerig, their value in terms of information, and their limits in terms of applicability

International Nuclear Information System (INIS)

Heuser, F.W.

1983-01-01

This contribution first briefly explains the main pillars of the deterministic safety concept as developed in nuclear engineering, and some basic ideas on risk analyses in general. This is followed by an outline of the methodology and main purposes of risk analyses. The German Risk Study is taken as an example to discuss selected aspects with regard to information value and limits of risk analyses. The main conclusions state that risk analyses are a valuable instrument for quantitative safety evaluation, leading to a better understanding of safety problems and their prevention, and allowing a comparative assessment of various safety measures. They furthermore allow a refined evaluation of a variety of accident parameters and other impacts determining the risk emanating from accidents. The current state of the art in this sector still leaves numerous uncertainties so that risk analyses yield information for assessments rather than for definite predictions. However, the urge for quantifying the lack of knowledge leads to a better and more precise determination of the gaps still to be filled up by researchers and engineers. Thus risk analyses are a useful help in defining suitable approaches and setting up standards, showing the tasks to be fulfilled in safety research in general. (orig./HSCH) [de

Formal Safety versus Real Safety: Quantitative and Qualitative Approaches to Safety Culture – Evidence from Estonia

Directory of Open Access Journals (Sweden)

Järvis Marina

2016-10-01

Full Text Available This paper examines differences between formal safety and real safety in Estonian small and medium-sized enterprises. The results reveal key issues in safety culture assessment. Statistical analysis of safety culture questionnaires showed many organisations with an outstanding safety culture and positive safety attitudes. However, qualitative data indicated some important safety weaknesses and aspects that should be included in the process of evaluation of safety culture in organisations.

Using a quantitative risk register to promote learning from a patient safety reporting system.

Science.gov (United States)

Mansfield, James G; Caplan, Robert A; Campos, John S; Dreis, David F; Furman, Cathie

2015-02-01

Patient safety reporting systems are now used in most health care delivery organizations. These systems, such as the one in use at Virginia Mason (Seattle) since 2002, can provide valuable reports of risk and harm from the front lines of patient care. In response to the challenge of how to quantify and prioritize safety opportunities, a risk register system was developed and implemented. Basic risk register concepts were refined to provide a systematic way to understand risks reported by staff. The risk register uses a comprehensive taxonomy of patient risk and algorithmically assigns each patient safety report to 1 of 27 risk categories in three major domains (Evaluation, Treatment, and Critical Interactions). For each category, a composite score was calculated on the basis of event rate, harm, and cost. The composite scores were used to identify the "top five" risk categories, and patient safety reports in these categories were analyzed in greater depth to find recurrent patterns of risk and associated opportunities for improvement. The top five categories of risk were easy to identify and had distinctive "profiles" of rate, harm, and cost. The ability to categorize and rank risks across multiple dimensions yielded insights not previously available. These results were shared with leadership and served as input for planning quality and safety initiatives. This approach provided actionable input for the strategic planning process, while at the same time strengthening the Virginia Mason culture of safety. The quantitative patient safety risk register serves as one solution to the challenge of extracting valuable safety lessons from large numbers of incident reports and could profitably be adopted by other organizations.

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.

A simplified method for quantitative assessment of the relative health and safety risk of environmental management activities

International Nuclear Information System (INIS)

Eide, S.A.; Smith, T.H.; Peatross, R.G.; Stepan, I.E.

1996-09-01

This report presents a simplified method to assess the health and safety risk of Environmental Management activities of the US Department of Energy (DOE). The method applies to all types of Environmental Management activities including waste management, environmental restoration, and decontamination and decommissioning. The method is particularly useful for planning or tradeoff studies involving multiple conceptual options because it combines rapid evaluation with a quantitative approach. The method is also potentially applicable to risk assessments of activities other than DOE Environmental Management activities if rapid quantitative results are desired

Scoping analyses for the safety injection system configuration for Korean next generation reactor

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Song, Jin Ho; Park, Jong Kyoon

1996-01-01

Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are performed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSl pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SlT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA

Reactivity initiated accident analyses for the safety assessment of upgraded JRR-3

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1984-08-01

JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. This report describes the analytical results of reactivity initiated accidents for the safety assessment of upgraded JRR-3. The following five cases have been selected for the assessment; (1) uncontrolled control rod withdrawal from zero power, (2) uncontrolled control rod withdrawal from full power, (3) removal of irradiation samples, (4) increase of primary coolant flow, (5) failure of heavy water tank. Parameter studies have been made for each of the above cases to cover possible uncertainties. All analyses have been made by a computer code EUREKA-2. The results show that the safety criteria for upgraded JRR-3 are all met and the adequacy of the design is confirmed. (author)

Methodology development for statistical evaluation of reactor safety analyses

International Nuclear Information System (INIS)

Mazumdar, M.; Marshall, J.A.; Chay, S.C.; Gay, R.

1976-07-01

In February 1975, Westinghouse Electric Corporation, under contract to Electric Power Research Institute, started a one-year program to develop methodology for statistical evaluation of nuclear-safety-related engineering analyses. The objectives of the program were to develop an understanding of the relative efficiencies of various computational methods which can be used to compute probability distributions of output variables due to input parameter uncertainties in analyses of design basis events for nuclear reactors and to develop methods for obtaining reasonably accurate estimates of these probability distributions at an economically feasible level. A series of tasks was set up to accomplish these objectives. Two of the tasks were to investigate the relative efficiencies and accuracies of various Monte Carlo and analytical techniques for obtaining such estimates for a simple thermal-hydraulic problem whose output variable of interest is given in a closed-form relationship of the input variables and to repeat the above study on a thermal-hydraulic problem in which the relationship between the predicted variable and the inputs is described by a short-running computer program. The purpose of the report presented is to document the results of the investigations completed under these tasks, giving the rationale for choices of techniques and problems, and to present interim conclusions

A new look on the safety case for geologic disposal

International Nuclear Information System (INIS)

Pescatore, Claudio; Riotte, Hans; Voinis, Sylvie

2005-01-01

It has become evident that the development of a geologic repository will involve a number of stages punctuated by interdependent decisions on whether and how to move to the next stage. These decisions require a clear and traceable presentation of technical and scientific arguments that will help in giving confidence in the feasibility and safety of a proposed concept. A detailed safety assessment is typically required at major decision points in repository planning and implementation, including decisions that require the granting of licenses. In recent years the scope of the safety assessment has broadened to include the collation of a broad range of evidence and arguments that complement and support the reliability of the results of quantitative analyses, and the broader term 'post-closure safety case' or simply 'safety case' is used to refer to these studies. This paper reflects the historical development from integrated safety assessment to modern safety cases and outlines the main elements of a safety case for geologic disposal. The presentation of the safety strategy, multiple barrier concept and strategies to deal with uncertainties are analysed and the importance of an explicit statement of confidence is emphasized. (author)

Safety and sensitivity analyses of a generic geologic disposal system for high-level radioactive waste

International Nuclear Information System (INIS)

Kimura, Hideo; Takahashi, Tomoyuki; Shima, Shigeki; Matsuzuru, Hideo

1994-11-01

This report describes safety and sensitivity analyses of a generic geologic disposal system for HLW, using a GSRW code and an automated sensitivity analysis methodology based on the Differential Algebra. An exposure scenario considered here is based on a normal evolution scenario which excludes events attributable to probabilistic alterations in the environment. The results of sensitivity analyses indicate that parameters related to a homogeneous rock surrounding a disposal facility have higher sensitivities to the output analyzed here than those of a fractured zone and engineered barriers. The sensitivity analysis methodology provides technical information which might be bases for the optimization of design of the disposal facility. Safety analyses were performed on the reference disposal system which involve HLW in amounts corresponding to 16,000 MTU of spent fuels. The individual dose equivalent due to the exposure pathway ingesting drinking water was calculated using both the conservative and realistic values of geochemical parameters. In both cases, the committed dose equivalent evaluated here is the order of 10 -7 Sv, and thus geologic disposal of HLW may be feasible if the disposal conditions assumed here remain unchanged throughout the periods assessed here. (author)

Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2011-01-01

Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

Recognising safety critical events: can automatic video processing improve naturalistic data analyses?

Science.gov (United States)

Dozza, Marco; González, Nieves Pañeda

2013-11-01

New trends in research on traffic accidents include Naturalistic Driving Studies (NDS). NDS are based on large scale data collection of driver, vehicle, and environment information in real world. NDS data sets have proven to be extremely valuable for the analysis of safety critical events such as crashes and near crashes. However, finding safety critical events in NDS data is often difficult and time consuming. Safety critical events are currently identified using kinematic triggers, for instance searching for deceleration below a certain threshold signifying harsh braking. Due to the low sensitivity and specificity of this filtering procedure, manual review of video data is currently necessary to decide whether the events identified by the triggers are actually safety critical. Such reviewing procedure is based on subjective decisions, is expensive and time consuming, and often tedious for the analysts. Furthermore, since NDS data is exponentially growing over time, this reviewing procedure may not be viable anymore in the very near future. This study tested the hypothesis that automatic processing of driver video information could increase the correct classification of safety critical events from kinematic triggers in naturalistic driving data. Review of about 400 video sequences recorded from the events, collected by 100 Volvo cars in the euroFOT project, suggested that drivers' individual reaction may be the key to recognize safety critical events. In fact, whether an event is safety critical or not often depends on the individual driver. A few algorithms, able to automatically classify driver reaction from video data, have been compared. The results presented in this paper show that the state of the art subjective review procedures to identify safety critical events from NDS can benefit from automated objective video processing. In addition, this paper discusses the major challenges in making such video analysis viable for future NDS and new potential

Plasma-safety assessment model and safety analyses of ITER

International Nuclear Information System (INIS)

Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

2001-01-01

A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

Multi-person and multi-attribute design evaluations using evidential reasoning based on subjective safety and cost analyses

International Nuclear Information System (INIS)

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

1996-01-01

This paper presents an approach for ranking proposed design options based on subjective safety and cost analyses. Hierarchical system safety analysis is carried out using fuzzy sets and evidential reasoning. This involves safety modelling by fuzzy sets at the bottom level of a hierarchy and safety synthesis by evidential reasoning at higher levels. Fuzzy sets are also used to model the cost incurred for each design option. An evidential reasoning approach is then employed to synthesise the estimates of safety and cost, which are made by multiple designers. The developed approach is capable of dealing with problems of multiple designers, multiple attributes and multiple design options to select the best design. Finally, a practical engineering example is presented to demonstrate the proposed multi-person and multi-attribute design selection approach

Process and plant safety

CERN Document Server

Hauptmanns, Ulrich

2015-01-01

Accidents in technical installations are random events. Hence they cannot be totally avoided. Only the probability of their occurrence may be reduced and their consequences be mitigated. The book proceeds from hazards caused by materials and process conditions to indicating technical and organizational measures for achieving the objectives of reduction and mitigation. Qualitative methods for identifying weaknesses of design and increasing safety as well as models for assessing accident consequences are presented. The quantitative assessment of the effectiveness of safety measures is explained. The treatment of uncertainties plays a role there. They stem from the random character of the accident and from lacks of knowledge on some of the phenomena to be addressed. The reader is acquainted with the simulation of accidents, safety and risk analyses and learns how to judge the potential and limitations of mathematical modelling. Risk analysis is applied amongst others to “functional safety” and the determinat...

Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

International Nuclear Information System (INIS)

Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

2004-01-01

It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

International Nuclear Information System (INIS)

Wetzel, N.; Scharfe, A.

1998-01-01

The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

International Nuclear Information System (INIS)

Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

2015-01-01

The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

Preliminary standard review guide for Environmental Restoration/Decontamination and Decommissioning safety analyses

International Nuclear Information System (INIS)

Ellingson, D.R.

1993-06-01

The review guide is based on the shared experiences, approaches, and philosophies of the Environmental Restoration/Decontamination and Decommissioning (ER/D ampersand D) subgroup members. It is presented in the form of a review guide to maximize the benefit to both the safety analyses practitioner and reviewer. The guide focuses on those challenges that tend to be unique to ER/D ampersand D cleanup activities. Some of these experiences, approaches, and philosophies may find application or be beneficial to a broader spectrum of activities such as terminal cleanout or even new operations. Challenges unique to ER/D ampersand D activities include (1) consent agreements requiring activity startup on designated dates; (2) the increased uncertainty of specific hazards; and (3) the highly variable activities covered under the broad category of ER/D ampersand D. These unique challenges are in addition to the challenges encountered in all activities; e.g., new and changing requirements and multiple interpretations. The experiences in approaches, methods, and solutions to the challenges are documented from the practitioner and reviewer's perspective, thereby providing the viewpoints on why a direction was taken and the concerns expressed. Site cleanup consent agreements with predetermined dates for restoration activity startup add the dimension of imposed punitive actions for failure to meet the date. Approval of the safety analysis is a prerequisite to startup. Actions that increase expediency are (1) assuring activity safety; (2) documenting that assurance; and (3) acquiring the necessary approvals. These actions increase the timeliness of startup and decrease the potential for punitive action. Improvement in expediency has been achieved by using safety analysis techniques to provide input to the line management decision process rather than as a review of line management decisions. Expediency is also improved by sharing the safety input and resultant decisions with

A Unique Digital Electrocardiographic Repository for the Development of Quantitative Electrocardiography and Cardiac Safety: The Telemetric and Holter ECG Warehouse (THEW)

Science.gov (United States)

Couderc, Jean-Philippe

2010-01-01

The sharing of scientific data reinforces open scientific inquiry; it encourages diversity of analysis and opinion while promoting new research and facilitating the education of next generations of scientists. In this article, we present an initiative for the development of a repository containing continuous electrocardiographic information and their associated clinical information. This information is shared with the worldwide scientific community in order to improve quantitative electrocardiology and cardiac safety. First, we present the objectives of the initiative and its mission. Then, we describe the resources available in this initiative following three components: data, expertise and tools. The Data available in the Telemetric and Holter ECG Warehouse (THEW) includes continuous ECG signals and associated clinical information. The initiative attracted various academic and private partners whom expertise covers a large list of research arenas related to quantitative electrocardiography; their contribution to the THEW promotes cross-fertilization of scientific knowledge, resources, and ideas that will advance the field of quantitative electrocardiography. Finally, the tools of the THEW include software and servers to access and review the data available in the repository. To conclude, the THEW is an initiative developed to benefit the scientific community and to advance the field of quantitative electrocardiography and cardiac safety. It is a new repository designed to complement the existing ones such as Physionet, the AHA-BIH Arrhythmia Database, and the CSE database. The THEW hosts unique datasets from clinical trials and drug safety studies that, so far, were not available to the worldwide scientific community. PMID:20863512

A Methodology for Evaluating Quantitative Nuclear Safety Culture Impact

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

Through several accidents of NPPs including the Fukushima Daiichi in 2011 and Chernobyl accidents in 1986, nuclear safety culture has been emphasized in reactor safety world-widely. In Korea, KHNP evaluates the safety culture of NPP itself. KHNP developed the principles of the safety culture in consideration of the international standards. A questionnaire and interview questions are also developed based on these principles and it is used for evaluating the safety culture. However, existing methodology to evaluate the safety culture has some disadvantages. First, it is difficult to maintain the consistency of the assessment. Second, the period of safety culture assessment is too long (every two years) so it has limitations in preventing accidents occurred by a lack of safety culture. Third, it is not possible to measure the change in the risk of NPPs by weak safety culture since it is not clearly explains the effect of safety culture on the safety of NPPs. In this study, Safety Culture Impact Assessment Model (SCIAM) is developed overcoming these disadvantages. In this study, SCIAM which overcoming disadvantages of exiting safety culture assessment method is developed. SCIAM uses SCII to monitor the statues of the safety culture periodically and also uses RCDF to quantify the safety culture impact on NPP's safety. It is significant that SCIAM represents the standard of the healthy nuclear safety culture, while the exiting safety culture assessment presented only vulnerability of the safety culture of organization. SCIAM might contribute to monitoring the level of safety culture periodically and, to improving the safety of NPP.

A Methodology for Evaluating Quantitative Nuclear Safety Culture Impact

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2015-01-01

Through several accidents of NPPs including the Fukushima Daiichi in 2011 and Chernobyl accidents in 1986, nuclear safety culture has been emphasized in reactor safety world-widely. In Korea, KHNP evaluates the safety culture of NPP itself. KHNP developed the principles of the safety culture in consideration of the international standards. A questionnaire and interview questions are also developed based on these principles and it is used for evaluating the safety culture. However, existing methodology to evaluate the safety culture has some disadvantages. First, it is difficult to maintain the consistency of the assessment. Second, the period of safety culture assessment is too long (every two years) so it has limitations in preventing accidents occurred by a lack of safety culture. Third, it is not possible to measure the change in the risk of NPPs by weak safety culture since it is not clearly explains the effect of safety culture on the safety of NPPs. In this study, Safety Culture Impact Assessment Model (SCIAM) is developed overcoming these disadvantages. In this study, SCIAM which overcoming disadvantages of exiting safety culture assessment method is developed. SCIAM uses SCII to monitor the statues of the safety culture periodically and also uses RCDF to quantify the safety culture impact on NPP's safety. It is significant that SCIAM represents the standard of the healthy nuclear safety culture, while the exiting safety culture assessment presented only vulnerability of the safety culture of organization. SCIAM might contribute to monitoring the level of safety culture periodically and, to improving the safety of NPP

Operation safety of control systems. Principles and methods; Surete de fonctionnement des systemes de commande. Principes et methodes

Energy Technology Data Exchange (ETDEWEB)

Aubry, J.F. [Institut National Polytechnique, 54 - Nancy (France); Chatelet, E. [Universite de Technologie de Troyes, 10 (France)

2008-09-15

This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)

Qualitative and quantitative analyses of flavonoids in Spirodela polyrrhiza by high-performance liquid chromatography coupled with mass spectrometry.

Science.gov (United States)

Qiao, Xue; He, Wen-ni; Xiang, Cheng; Han, Jian; Wu, Li-jun; Guo, De-an; Ye, Min

2011-01-01

Spirodela polyrrhiza (L.) Schleid. is a traditional Chinese herbal medicine for the treatment of influenza. Despite its wide use in Chinese medicine, no report on quality control of this herb is available so far. To establish qualitative and quantitative analytical methods by high-performance liquid chromatography (HPLC) coupled with mass spectrometry (MS) for the quality control of S. polyrrhiza. The methanol extract of S. polyrrhiza was analysed by HPLC/ESI-MS(n). Flavonoids were identified by comparing with reference standards or according to their MS(n) (n = 2-4) fragmentation behaviours. Based on LC/MS data, a standardised HPLC fingerprint was established by analysing 15 batches of commercial herbal samples. Furthermore, quantitative analysis was conducted by determining five major flavonoids, namely luteolin 8-C-glucoside, apigenin 8-C-glucoside, luteolin 7-O-glucoside, apigenin 7-O-glucoside and luteolin. A total of 18 flavonoids were identified by LC/MS, and 14 of them were reported from this herb for the first time. The HPLC fingerprints contained 10 common peaks, and could differentiate good quality batches from counterfeits. The total contents of five major flavonoids in S. polyrrhiza varied significantly from 4.28 to 19.87 mg/g. Qualitative LC/MS and quantitative HPLC analytical methods were established for the comprehensive quality control of S. polyrrhiza. Copyright © 2011 John Wiley & Sons, Ltd.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Reentry safety for the Topaz II Space Reactor: Issues and analyses

International Nuclear Information System (INIS)

Connell, L.W.; Trost, L.C.

1994-03-01

This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents

Quantitative analysis of pedestrian safety at uncontrolled multi-lane mid-block crosswalks in China.

Science.gov (United States)

Zhang, Cunbao; Zhou, Bin; Chen, Guojun; Chen, Feng

2017-11-01

A lot of pedestrian-vehicle crashes at mid-block crosswalks severely threaten pedestrian's safety around the world. The situations are even worse in China due to low yielding rate of vehicles at crosswalks. In order to quantitatively analyze pedestrian's safety at multi-lane mid-block crosswalks, the number of pedestrian-vehicle conflicts was utilized to evaluate pedestrian's accident risk. Five mid-block crosswalks (Wuhan, China) were videoed to collect data of traffic situation and pedestrian-vehicle conflicts, and the quantity and spatial distribution of pedestrian-vehicle conflicts at multi-lane mid-block crosswalk were analyzed according to lane-based post-encroachment time(LPET). Statistical results indicate that conflicts are mainly concentrated in lane3 and lane6. Percentage of conflict of each lane numbered from 1 to 6 respectively are 4.1%, 13.1%, 19.8%, 8.4%, 19.0%, 28.1%. Conflict rate under different crossing strategies are also counted. Moreover, an order probit (OP) model of pedestrian-vehicle conflict analysis (PVCA) was built to find out the contributions corresponding to those factors (such as traffic volume, vehicle speed, pedestrian crossing behavior, pedestrian refuge, etc.) to pedestrian-vehicle conflicts. The results show that: pedestrian refuge have positive effects on pedestrian safety; on the other hand, high vehicle speed, high traffic volume, rolling gap crossing pattern, and larger pedestrian platoon have negative effects on pedestrian safety. Based on our field observation and PVCA model, the number of conflicts will rise by 2% while the traffic volume increases 200 pcu/h; similarly, if the vehicle speed increases 5km/h, the number of conflicts will rise by 12% accordingly. The research results could be used to evaluate pedestrian safety at multi-lane mid-block crosswalks, and useful to improve pedestrian safety by means of pedestrian safety education, pedestrian refuge setting, vehicle speed limiting, and so on. Copyright © 2017

Development of Quantitative Framework for Event Significance Evaluation

International Nuclear Information System (INIS)

Lee, Durk Hun; Kim, Min Chull; Kim, Inn Seock

2010-01-01

There is an increasing trend in quantitative evaluation of the safety significance of operational events using Probabilistic Safety Assessment (PSA) technique. An integrated framework for evaluation of event significance has been developed by Korea Institute of Nuclear Safety (KINS), which consists of an assessment hierarchy and a number of matrices. The safety significance of various events, e.g., internal or external initiating events that occurred during at-power or shutdown conditions, can be quantitatively analyzed using this framework, and then, the events rated according to their significance. This paper briefly describes the basic concept of the integrated quantitative framework for evaluation of event significance, focusing on the assessment hierarchy

BBN based Quantitative Assessment of Software Design Specification

International Nuclear Information System (INIS)

Eom, Heung-Seop; Park, Gee-Yong; Kang, Hyun-Gook; Kwon, Kee-Choon; Chang, Seung-Cheol

2007-01-01

Probabilistic Safety Assessment (PSA), which is one of the important methods in assessing the overall safety of a nuclear power plant (NPP), requires quantitative reliability information of safety-critical software, but the conventional reliability assessment methods can not provide enough information for PSA of a NPP. Therefore current PSA which includes safety-critical software does not usually consider the reliability of the software or uses arbitrary values for it. In order to solve this situation this paper proposes a method that can produce quantitative reliability information of safety-critical software for PSA by making use of Bayesian Belief Networks (BBN). BBN has generally been used to model an uncertain system in many research fields including the safety assessment of software. The proposed method was constructed by utilizing BBN which can combine the qualitative and the quantitative evidence relevant to the reliability of safety critical software. The constructed BBN model can infer a conclusion in a formal and a quantitative way. A case study was carried out with the proposed method to assess the quality of software design specification (SDS) of safety-critical software that will be embedded in a reactor protection system. The intermediate V and V results of the software design specification were used as inputs to the BBN model

Quantitative Analyse und Visualisierung der Herzfunktionen

Science.gov (United States)

Sauer, Anne; Schwarz, Tobias; Engel, Nicole; Seitel, Mathias; Kenngott, Hannes; Mohrhardt, Carsten; Loßnitzer, Dirk; Giannitsis, Evangelos; Katus, Hugo A.; Meinzer, Hans-Peter

Die computergestützte bildbasierte Analyse der Herzfunktionen ist mittlerweile Standard in der Kardiologie. Die verfügbaren Produkte erfordern meist ein hohes Maß an Benutzerinteraktion und somit einen erhöhten Zeitaufwand. In dieser Arbeit wird ein Ansatz vorgestellt, der dem Kardiologen eine größtenteils automatische Analyse der Herzfunktionen mittels MRT-Bilddaten ermöglicht und damit Zeitersparnis schafft. Hierbei werden alle relevanten herzphysiologsichen Parameter berechnet und mithilfe von Diagrammen und Graphen visualisiert. Diese Berechnungen werden evaluiert, indem die ermittelten Werte mit manuell vermessenen verglichen werden. Der hierbei berechnete mittlere Fehler liegt mit 2,85 mm für die Wanddicke und 1,61 mm für die Wanddickenzunahme immer noch im Bereich einer Pixelgrösse der verwendeten Bilder.

Quantitative analysis of fission products by {gamma} spectrography; Analyse quantitative des produits de fission par spectrographie {gamma}

Energy Technology Data Exchange (ETDEWEB)

Malet, G

1962-07-01

The activity of the fission products present in treated solutions of irradiated fuels is given as a function of the time of cooling and of the irradiation time. The variation of the ratio ({sup 144}Ce + {sup 144}Pr activity/{sup 137}Cs activity) as a function of these same parameters is also given. From these results a method is deduced giving the 'age' of the solution analyzed. By {gamma}-scintillation spectrography it was possible to estimate the following elements individually: {sup 141}Ce, {sup 144}Ce + {sup 144}Pr, {sup 103}Ru, {sup 106}Ru + {sup 106}Rh, {sup 137}Cs, {sup 95}Zr + {sup 95}Nb. Yield curves are given for the case of a single emitter. Of the various existing methods, that of the least squares was used for the quantitative analysis of the afore-mentioned fission products. The accuracy attained varies from 3 to 10%. (author) [French] L'activite des produits de fission presents dans les solutions de traitement de combustibles irradies est donnee en fonction du temps de refroidissement et du temps d'irradiation. On etudie de plus la variation du rapport Activite du {sup 144}Ce + {sup 144}Pr /Activite du {sup 137}Cs en fonction de ces memes parametres. De ces resultats, on deduit une methode donnant l'age de la solution analysee. La spectrographie {gamma} a scintillation a permis le dosage individuel des produits suivants: {sup 141}Ce, {sup 144}Ce + {sup 144}Pr, {sup 103}Ru, {sup 106}Ru + {sup 106}Rh, {sup 137}Cs, {sup 95}Zr + {sup 95}Nb. Des courbes de rendement sont donnees dans le cas d'un emetteur unique. Des differentes methodes existantes, la methode des moindres carres a ete employee pour l'analyse quantitative des produits de fission precites. La precision obtenue varie entre 3 et 10 pour cent. (auteur)

A study on a quantitative V and V for safety-critical software

International Nuclear Information System (INIS)

Eom, Heung Seop; Son, Han Seong; Kang, Hyun Gook; Chang, Seung Cheol

2004-01-01

Verification and Validation (V and V) plays important role in assessing the safety-critical software embedded in the digital systems for a Nuclear Power Plant. A conventional V and V usually adopts a checklist method and its answers are mostly qualitative. There are some limitations to this conventional V and V method. First, the difficulties in using the checklist method are: Even for an acceptable software, some checklist questions will have negative answers. The checklist itself does not help to explain the reasons for drawing an overall positive conclusion in the presence of a few negative answers. The checklist does not help decide when enough issues have been examined to achieve a reasonable confidence in the software. The checklist method does not support a consideration of different kinds of information, such as software engineering measures. Second, a difficulty comes from the qualitative form of the answers in the checklist method, which is: It is usually hard to know when sufficient evidence has been collected. Finally a difficulty comes from a human expert's way of combining a great number of diverse evidence and inferring the conclusion, which is: Some of this evidence is qualitative and others are quantitative. Both are necessary to evaluate the quality of the software correctly. But, in general, the experts' way of combining the diverse evidence and performing an inference is usually informal and qualitative, which is hard to discuss and will eventually lead to a debate about the conclusion. Our overall goal is to develop a systematic method that can obtain quantitative information of the software quality from the works of V and V. To achieve this goal and to solve the above-mentioned problems in the current V and V method, we studied a method that can combine qualitative and quantitative evidence, and can infer a conclusion in a formal and a quantitative way by using the benefits of BBN

C4P cross-section libraries for safety analyses with SIMMER and related studies

International Nuclear Information System (INIS)

Rineiski, A.; Sinitsa, V.; Gabrielli, F.; Maschek, W.

2011-01-01

A code and data system, C 4 P, is under development at KIT. It includes fine-group master libraries and tools for generating problem-oriented cross-section libraries, primarily for safety studies with the SIMMER code and related analyses. In the paper, the 560-group master library and problem oriented 40-group and 72-group cross-section libraries, for thermal and fast systems, respectively, are described and their performances are investigated. (author)

Application of Safety Maturity Model and 4P-4C Model in Safety Culture Assessment

International Nuclear Information System (INIS)

Choi, K. S.; Lee, Y. E.; Ha, J. T.; Chang, H. S.; Kam, S. C.

2010-01-01

Korean government and utility have made efforts to enhance the nuclear safety culture and the development of quantitative index of safety culture was promoted for past several years. Quantitative index of safety culture and the past efforts to understand safety culture need insight into the concept of culture. This paper aims to apply new method of measuring nuclear safety culture through the review of approaches of evaluating safety culture in non-nuclear industries. Scoring table has been developed based on new models and example of result of interviews evaluating the nuclear safety culture is also shown

Quantitative analyses of the 3D nuclear landscape recorded with super-resolved fluorescence microscopy.

Science.gov (United States)

Schmid, Volker J; Cremer, Marion; Cremer, Thomas

2017-07-01

Recent advancements of super-resolved fluorescence microscopy have revolutionized microscopic studies of cells, including the exceedingly complex structural organization of cell nuclei in space and time. In this paper we describe and discuss tools for (semi-) automated, quantitative 3D analyses of the spatial nuclear organization. These tools allow the quantitative assessment of highly resolved different chromatin compaction levels in individual cell nuclei, which reflect functionally different regions or sub-compartments of the 3D nuclear landscape, and measurements of absolute distances between sites of different chromatin compaction. In addition, these tools allow 3D mapping of specific DNA/RNA sequences and nuclear proteins relative to the 3D chromatin compaction maps and comparisons of multiple cell nuclei. The tools are available in the free and open source R packages nucim and bioimagetools. We discuss the use of masks for the segmentation of nuclei and the use of DNA stains, such as DAPI, as a proxy for local differences in chromatin compaction. We further discuss the limitations of 3D maps of the nuclear landscape as well as problems of the biological interpretation of such data. Copyright © 2017 Elsevier Inc. All rights reserved.

Quantifying system safety: A comparison of the SBOAT & Safety Barrier Manager tools

OpenAIRE

Hansen, Zaza Nadja Lee; Duijm, Nijs Jan; Markert, Frank; Herbert, Luke Thomas

2015-01-01

This paper presents two software tools for analyzing safety risks, SBOAT (Stochastic BPMN Optimisation and Analysis Tool) and SBM (SafetyBarrierManagerr). SBOAT employs principles from stochastic model checking to allow for the quantitative verification of workflows. SBM supports the creation of valid safety-barrier diagrams and allows the quantitative analysis of the probability of all possible end states of the barrier diagram, i.e. the outcomes if one or several of the barriers fail to per...

Design provisions for safety

International Nuclear Information System (INIS)

Birkhofer, A.

1983-01-01

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

Current regulatory developments concerning the implementation of probabilistic safety analyses for external hazards in Germany

International Nuclear Information System (INIS)

Krauss, Matias; Berg, Heinz-Peter

2014-01-01

The Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) initiated in September 2003 a comprehensive program for the revision of the national nuclear safety regulations which has been successfully completed in November 2012. These nuclear regulations take into account the current recommendations of the International Atomic Energy Agency (IAEA) and Western European Nuclear Regulators Association (WENRA). In this context, the recommendations and guidelines of the Nuclear Safety Standards Commission (KTA) and the technical documents elaborated by the respective expert group on Probabilistic Safety Analysis for Nuclear Power Plants (FAK PSA) are being updated or in the final process of completion. A main topic of the revision was the issue external hazards. As part of this process and in the light of the accident at Fukushima and the findings of the related actions resulting in safety reviews of nuclear power plants at national level in Germany and on European level, a revision of all relevant standards and documents has been made, especially the recommendations of KTA and FAK PSA. In that context, not only design issues with respect to events such as earthquakes and floods have been discussed, but also methodological issues regarding the implementation of improved probabilistic safety analyses on this topic. As a result of the revision of the KTA 2201 series 'Design of Nuclear Power Plants against Seismic Events' with their parts 1 to 6, part 1 'Principles' was published as the first standard in November 2011, followed by the revised versions of KTA 2201.2 (soil) and 2201.4 (systems and components) in 2012. The modified the standard KTA 2201.3 (structures) is expected to be issued before the end of 2013. In case of part 5 (seismic instrumentation) and part 6 (post>seismic actions) draft amendments are expected in 2013. The expert group 'Probabilistic Safety Assessments for Nuclear Power Plants' (FAK PSA) is an advisory body of the Federal

On the road to quantitative genetic/genomic analyses of root growth and development components underlying root architecture

International Nuclear Information System (INIS)

Draye, X.; Dorlodot, S. de; Lavigne, T.

2006-01-01

The quantitative genetic and functional genomic analyses of root development, growth and plasticity will be instrumental in revealing the major regulatory pathways of root architecture. Such knowledge, combined with in-depth consideration of root physiology (e.g. uptake, exsudation), form (space-time dynamics of soil exploration) and ecology (including root environment), will settle the bases for designing root ideotypes for specific environments, for low-input agriculture or for successful agricultural production with minimal impact on the environment. This report summarizes root research initiated in our lab between 2000 and 2004 in the following areas: quantitative analysis of root branching in bananas, high throughput characterisation of root morphology, image analysis, QTL mapping of detailed features of root architecture in rice, and attempts to settle a Crop Root Research Consortium. (author)

Quantitative evaluation of the fault tolerance of systems important to the safety of atomic power plants

International Nuclear Information System (INIS)

Malkin, S.D.; Sivokon, V.P.; Shmatkova, L.V.

1989-01-01

Fault tolerance is the property of a system to preserve its performance upon failures of its components. Thus, in nuclear-reactor technology one has only a qualitative evaluation of fault tolerance - the single-failure criterion, which does not enable one to compare and perform goal-directed design of fault-tolerant systems, and in the field of computer technology there are no generally accepted evaluations of fault tolerance that could be applied effectively to reactor systems. This paper considers alternative evaluations of fault tolerance and a method of comprehensive automated calculation of the reliability and fault tolerance of complex systems. The authors presented quantitative estimates of fault tolerance that develop the single-failure criterion. They have limiting processes that allow simple and graphical standardization. They worked out a method and a program for comprehensive calculation of the reliability and fault tolerance of systems of complex structure that are important to the safety of atomic power plants. The quantitative evaluation of the fault tolerance of these systems exhibits a degree of insensitivity to failures and shows to what extent their reliability is determined by a rigorously defined structure, and to what extent by the probabilistic reliability characteristics of the components. To increase safety, one must increase the fault tolerance of the most important systems of atomic power plants

Microsegregation in multicomponent alloy analysed by quantitative phase-field model

International Nuclear Information System (INIS)

Ohno, M; Takaki, T; Shibuta, Y

2015-01-01

Microsegregation behaviour in a ternary alloy system has been analysed by means of quantitative phase-field (Q-PF) simulations with a particular attention directed at an influence of tie-line shift stemming from different liquid diffusivities of the solute elements. The Q-PF model developed for non-isothermal solidification in multicomponent alloys with non-zero solid diffusivities was applied to analysis of microsegregation in a ternary alloy consisting of fast and slow diffusing solute elements. The accuracy of the Q-PF simulation was first verified by performing the convergence test of segregation ratio with respect to the interface thickness. From one-dimensional analysis, it was found that the microsegregation of slow diffusing element is reduced due to the tie-line shift. In two-dimensional simulations, the refinement of microstructure, viz., the decrease of secondary arms spacing occurs at low cooling rates due to the formation of diffusion layer of slow diffusing element. It yields the reductions of degrees of microsegregation for both the fast and slow diffusing elements. Importantly, in a wide range of cooling rates, the degree of microsegregation of the slow diffusing element is always lower than that of the fast diffusing element, which is entirely ascribable to the influence of tie-line shift. (paper)

Comparison of simplified quantitative analyses of FDG uptake

International Nuclear Information System (INIS)

Graham, M.M.; Peterson, L.M.; Hayward, R.M.

2000-01-01

Quantitative analysis of [ 18 F]-fluoro-deoxyglucose (FDG) uptake is important in oncologic positron emission tomography (PET) studies to be able to set an objective threshold in determining if a tissue is malignant or benign, in assessing response to therapy, and in attempting to predict the aggressiveness of an individual tumor. The most common method used today for simple, clinical quantitation is standardized uptake value (SUV). SUV is normalized for body weight. Other potential normalization factors are lean body mass (LBM) or body surface area (BSA). More complex quantitation schemes include simplified kinetic analysis (SKA), Patlak graphical analysis (PGA), and parameter optimization of the complete kinetic model to determine FDG metabolic rate (FDGMR). These various methods were compared in a group of 40 patients with colon cancer metastatic to the liver. The methods were assessed by (1) correlation with FDGMR, (2) ability to predict survival using Kaplan-Meier plots, and (3) area under receiver operating characteristic (ROC) curves for distinguishing between tumor and normal liver. The best normalization scheme appears to be BSA with minor differences depending on the specific formula used to calculate BSA. Overall, PGA is the best predictor of outcome and best discriminator between normal tissue and tumor. SKA is almost as good. In conventional PET imaging it is worthwhile to normalize SUV using BSA. If a single blood sample is available, it is possible to use the SKA method, which is distinctly better. If more than one image is available, along with at least one blood sample, PGA is feasible and should produce the most accurate results

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Swiss regulatory use of databanks for nuclear power plant life management, surveillance and safety analyses

International Nuclear Information System (INIS)

Tipping, Ph.; Beutler, R.; Schoen, G.; Noeggerath, J.

2002-01-01

Full text: As operational time is accumulated, the overall safety and performance of nuclear power plants (NPPs) will tend to be characterised by those areas in which structures, systems and components (SSCs) have not performed as well, or as reliably, as expected. The reasons for non-availability of equipment in NPPs due to SSC material malfunction or unsatisfactory performance, leading to events or even accidents, are varied and they must be analysed in order to obtain the root causes. Once the root causes are identified, corresponding measures can be applied in order to improve reliability and therefore safety. The root cause information obtained, if brought into user-friendly databanks (DBs), can be used to follow NPP performance trends, to check whether a repair or replacement has been effective, to focus regulatory attention and NPP surveillance on known weak-spots and to serve as an advance indicator where potential problems may arise. Using the DBs, similar occurrences of failures or problems in other NPPs can be identified and generic issues recognised early on and preventative action taken. The following describes the Swiss Federal Nuclear Safety Inspectorate's (HSK) DB concepts for keeping track of NPP safety and lifetime management issues. Typical sources of data for the Inspectorate's DBs are, for example, the IAEA/NEA Incident Reporting System (IRS) reports, US-NRC Generic Letters, the Swiss NPP's own reports (monthly, annual and normal outage) and, more importantly, the document that these NPPs must issue to the Inspectorate whenever a reportable event takes place. Specifically, the reporting of events in the NPPs is laid down in the Inspectorate's Guideline (R-15 'Reporting Guideline Concerning The Operation of Nuclear Power Plants'). In this Guideline, reportable events are defined and the criteria for assessing the degree of importance or impact on nuclear safety are given. In this manner, a standard and consistent approach to data collection is

Safety barriers on oil and gas platforms. Means to prevent hydrocarbon releases

Energy Technology Data Exchange (ETDEWEB)

Sklet, Snorre

2005-12-15

The main objective of the PhD project has been to develop concepts and methods that can be used to define, illustrate, analyse, and improve safety barriers in the operational phase of offshore oil and gas production platforms. The main contributions of this thesis are; Clarification of the term safety barrier with respect to definitions, classification, and relevant attributes for analysis of barrier performance Development and discussion of a representative set of hydrocarbon release scenarios Development and testing of a new method, BORA-Release, for qualitative and quantitative risk analysis of hydrocarbon releases Safety barriers are defined as physical and/or non-physical means planned to prevent, control, or mitigate undesired events or accidents. The means may range from a single technical unit or human actions, to a complex socio-technical system. It is useful to distinguish between barrier functions and barrier systems. Barrier functions describe the purpose of safety barriers or what the safety barriers shall do in order to prevent, control, or mitigate undesired events or accidents. Barrier systems describe how a barrier function is realized or executed. If the barrier system is functioning, the barrier function is performed. If a barrier function is performed successfully, it should have a direct and significant effect on the occurrence and/or consequences of an undesired event or accident. It is recommended to address the following attributes to characterize the performance of safety barriers; a) functionality/effectiveness, b) reliability/ availability, c) response time, d) robustness, and e) triggering event or condition. For some types of barriers, not all the attributes are relevant or necessary in order to describe the barrier performance. The presented hydrocarbon release scenarios include initiating events, barrier functions introduced to prevent hydrocarbon releases, and barrier systems realizing the barrier functions. Both technical and human

Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

Energy Technology Data Exchange (ETDEWEB)

2009-11-15

The objective with this report is to: - provide design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel, to form the basis for the development of the reference design of the repository. The design premises are used as input to the documents, called production reports, that present the reference design to be analysed in the long term safety assessment SR-Site. It is the aim that the production reports should verify that the chosen design complies with the design premises given in this report, whereas this report takes the burden of justifying why these design premises are relevant. The more specific aims and objectives with the production reports are provided in these reports. The following approach is used: - The reference design analysed in SR-Can is a starting point for setting safety related design premises for the next design step. - A few design basis cases, in accordance with the definition used in the regulation SSMFS 2008:211 and mainly related to the canister, can be derived from the results of the SR-Can assessment. From these it is possible to formulate some specific design premises for the canister. - The design basis cases involve several assumptions on the state of other barriers. These implied conditions are thus set as design premises for these barriers. - Even if there are few load cases on individual barriers that can be directly derived from the analyses, SR-Can provides substantial feedback on most aspects of the analysed reference design. This feedback is also formulated as design premises. - An important part of SR-Can Main report is the formulation and assessment of safety function indicator criteria. These criteria are a basis for formulating design premises, but they are not the same as the design premises discussed in the present report. Whereas the former should be upheld throughout the assessment period, the latter refer to the initial state and must be defined such that they give a margin for

Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

International Nuclear Information System (INIS)

2009-11-01

The objective with this report is to: - provide design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel, to form the basis for the development of the reference design of the repository. The design premises are used as input to the documents, called production reports, that present the reference design to be analysed in the long term safety assessment SR-Site. It is the aim that the production reports should verify that the chosen design complies with the design premises given in this report, whereas this report takes the burden of justifying why these design premises are relevant. The more specific aims and objectives with the production reports are provided in these reports. The following approach is used: - The reference design analysed in SR-Can is a starting point for setting safety related design premises for the next design step. - A few design basis cases, in accordance with the definition used in the regulation SSMFS 2008:211 and mainly related to the canister, can be derived from the results of the SR-Can assessment. From these it is possible to formulate some specific design premises for the canister. - The design basis cases involve several assumptions on the state of other barriers. These implied conditions are thus set as design premises for these barriers. - Even if there are few load cases on individual barriers that can be directly derived from the analyses, SR-Can provides substantial feedback on most aspects of the analysed reference design. This feedback is also formulated as design premises. - An important part of SR-Can Main report is the formulation and assessment of safety function indicator criteria. These criteria are a basis for formulating design premises, but they are not the same as the design premises discussed in the present report. Whereas the former should be upheld throughout the assessment period, the latter refer to the initial state and must be defined such that they give a margin for

Theory and Practice in Quantitative Genetics

DEFF Research Database (Denmark)

Posthuma, Daniëlle; Beem, A Leo; de Geus, Eco J C

2003-01-01

With the rapid advances in molecular biology, the near completion of the human genome, the development of appropriate statistical genetic methods and the availability of the necessary computing power, the identification of quantitative trait loci has now become a realistic prospect for quantitative...... geneticists. We briefly describe the theoretical biometrical foundations underlying quantitative genetics. These theoretical underpinnings are translated into mathematical equations that allow the assessment of the contribution of observed (using DNA samples) and unobserved (using known genetic relationships......) genetic variation to population variance in quantitative traits. Several statistical models for quantitative genetic analyses are described, such as models for the classical twin design, multivariate and longitudinal genetic analyses, extended twin analyses, and linkage and association analyses. For each...

The Study on the Quantitative Analysis in LPG Tank's Fire and Explosion

Energy Technology Data Exchange (ETDEWEB)

Bae, S.J.; Kim, B.J. [Department of chemical Engineering, Soongsil University, Seoul (Korea)

1999-04-01

Chemical plant's fire and explosion does not only damage to the chemical plants themselves but also damage to people in or near of the accident spot and the neighborhood of chemical plant. For that reason, Chemical process safety management has become important. One of safety management methods is called 'the quantitative analysis', which is used to reduce and prevent the accident. The results of the quantitative analysis could be used to arrange the equipments, evaluate the minimum safety distance, prepare the safety equipments. In this study we make the computer program to make easy to do quantitative analysis of the accident. The output of the computer program is the magnitude of fire(pool fire and fireball) and explosion (UVCE and BLEVE) effects. We used the thermal radiation as a measure of fire magnitude and used the overpressure as a measure of explosion magnitude. In case of BLEVE, the fly distance of fragment can be evaluated. Also probit analysis was done in every case. As the case study, Buchun LPG explosion accident in Korea was analysed by the program developed. The simulation results showed that the permissible distance was 800m and probit analysis showed that 1st degree burn, 2nd degree burn, and death distances are 450, 280, 260m, respectively. the simulation results showed the good agreement with the result from SAFER PROGRAM made by DuPont. 13 refs., 4 figs., 2 tabs.

Precursor analyses - The use of deterministic and PSA based methods in the event investigation process at nuclear power plants

International Nuclear Information System (INIS)

2004-09-01

The efficient feedback of operating experience (OE) is a valuable source of information for improving the safety and reliability of nuclear power plants (NPPs). It is therefore essential to collect information on abnormal events from both internal and external sources. Internal operating experience is analysed to obtain a complete understanding of an event and of its safety implications. Corrective or improvement measures may then be developed, prioritized and implemented in the plant if considered appropriate. Information from external events may also be analysed in order to learn lessons from others' experience and prevent similar occurrences at our own plant. The traditional ways of investigating operational events have been predominantly qualitative. In recent years, a PSA-based method called probabilistic precursor event analysis has been developed, used and applied on a significant scale in many places for a number of plants. The method enables a quantitative estimation of the safety significance of operational events to be incorporated. The purpose of this report is to outline a synergistic process that makes more effective use of operating experience event information by combining the insights and knowledge gained from both approaches, traditional deterministic event investigation and PSA-based event analysis. The PSA-based view on operational events and PSA-based event analysis can support the process of operational event analysis at the following stages of the operational event investigation: (1) Initial screening stage. (It introduces an element of quantitative analysis into the selection process. Quantitative analysis of the safety significance of nuclear plant events can be a very useful measure when it comes to selecting internal and external operating experience information for its relevance.) (2) In-depth analysis. (PSA based event evaluation provides a quantitative measure for judging the significance of operational events, contributors to

Quantitative risk assessment of digitalized safety systems

Energy Technology Data Exchange (ETDEWEB)

Shin, Sung Min; Lee, Sang Hun; Kang, Hym Gook [KAIST, Daejeon (Korea, Republic of); Lee, Seung Jun [UNIST, Ulasn (Korea, Republic of)

2016-05-15

A report published by the U.S. National Research Council indicates that appropriate methods for assessing reliability are key to establishing the acceptability of digital instrumentation and control (I and C) systems in safety-critical plants such as NPPs. Since the release of this issue, the methodology for the probabilistic safety assessment (PSA) of digital I and C systems has been studied. However, there is still no widely accepted method. Kang and Sung found three critical factors for safety assessment of digital systems: detection coverage of fault-tolerant techniques, software reliability quantification, and network communication risk. In reality the various factors composing digitalized I and C systems are not independent of each other but rather closely connected. Thus, from a macro point of view, a method that can integrate risk factors with different characteristics needs to be considered together with the micro approaches to address the challenges facing each factor.

Quantitative evaluation of safety use limit for crevice corrosion in Ni-Cr-Mo alloys

International Nuclear Information System (INIS)

Fukaya, Yuichi; Akashi, Masatsune; Sasaki, Hidetsugu; Tsujikawa, Shigeo

2007-01-01

The most important problem with corrosion-resistant alloys such as stainless steels is localized corrosion. Crevice corrosion, which is a typical localized corrosion, occurs under the mildest environmental conditions. Consequently, whether crevice corrosion occurs or not is an important issue in structural material selection. This study investigated highly corrosion-resistant Ni-Cr-Mo alloys whose resistance for crevice corrosion is difficult to evaluate with the JIS G 0592 standard for common strainless steels. The optimized procedures for determining the critical potential and temperature for crevice corrosion of the alloys were developed based on the JIS method. The limits of safety usage of various Ni-Cr-Mo alloys were evaluated quantitatively in chloride solution environments. (author)

Practicable methods for histological section thickness measurement in quantitative stereological analyses.

Science.gov (United States)

Matenaers, Cyrill; Popper, Bastian; Rieger, Alexandra; Wanke, Rüdiger; Blutke, Andreas

2018-01-01

The accuracy of quantitative stereological analysis tools such as the (physical) disector method substantially depends on the precise determination of the thickness of the analyzed histological sections. One conventional method for measurement of histological section thickness is to re-embed the section of interest vertically to its original section plane. The section thickness is then measured in a subsequently prepared histological section of this orthogonally re-embedded sample. However, the orthogonal re-embedding (ORE) technique is quite work- and time-intensive and may produce inaccurate section thickness measurement values due to unintentional slightly oblique (non-orthogonal) positioning of the re-embedded sample-section. Here, an improved ORE method is presented, allowing for determination of the factual section plane angle of the re-embedded section, and correction of measured section thickness values for oblique (non-orthogonal) sectioning. For this, the analyzed section is mounted flat on a foil of known thickness (calibration foil) and both the section and the calibration foil are then vertically (re-)embedded. The section angle of the re-embedded section is then calculated from the deviation of the measured section thickness of the calibration foil and its factual thickness, using basic geometry. To find a practicable, fast, and accurate alternative to ORE, the suitability of spectral reflectance (SR) measurement for determination of plastic section thicknesses was evaluated. Using a commercially available optical reflectometer (F20, Filmetrics®, USA), the thicknesses of 0.5 μm thick semi-thin Epon (glycid ether)-sections and of 1-3 μm thick plastic sections (glycolmethacrylate/ methylmethacrylate, GMA/MMA), as regularly used in physical disector analyses, could precisely be measured within few seconds. Compared to the measured section thicknesses determined by ORE, SR measures displayed less than 1% deviation. Our results prove the applicability

Investigation of nuclear power safety objects

International Nuclear Information System (INIS)

2003-09-01

It is a report of ground and concept of nuclear safety objects and future issues in Japan, which has investigated by the Committee of Experts on Investigation of Nuclear Safety Objects in the Nuclear Safety Research Association. The report consisted of member of committee, main conclusions and five chapters. The first chapter contains construction of safety objects and range of object, the second chapter qualitative safety objects, the third chapter quantitative safety objects, the forth subsiding objects and the fifth other items under consideration. The qualitative safety objects on individual and society, the quantitative one on effects on health and social cost, aspect of safety objects, relation between radiation protection and safety objects, practical objective values and earthquake are stated. (S.Y.)

Safety analysis procedures for PHWR

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, Hyoung Tae; Yoo, Kun Joong

2004-03-01

The methodology of safety analyses for CANDU reactors in Canada, a vendor country, uses a combination of best-estimate physical models and conservative input parameters so as to minimize the uncertainty of the plant behavior predictions. As using the conservative input parameters, the results of the safety analyses are assured the regulatory requirements such as the public dose, the integrity of fuel and fuel channel, the integrity of containment and reactor structures, etc. However, there is not the comprehensive and systematic procedures for safety analyses for CANDU reactors in Korea. In this regard, the development of the safety analyses procedures for CANDU reactors is being conducted not only to establish the safety analyses system, but also to enhance the quality assurance of the safety assessment. In the first phase of this study, the general procedures of the deterministic safety analyses are developed. The general safety procedures are covered the specification of the initial event, selection of the methodology and accident sequences, computer codes, safety analysis procedures, verification of errors and uncertainties, etc. Finally, These general procedures of the safety analyses are applied to the Large Break Loss Of Coolant Accident (LBLOCA) in Final Safety Analysis Report (FSAR) for Wolsong units 2, 3, 4

Scanning electron microscopic analyses of Ferrocyanide tank wastes for the Ferrocyanide safety program

International Nuclear Information System (INIS)

Callaway, W.S.

1995-09-01

This is Fiscal Year 1995 Annual Report on the progress of activities relating to the application of scanning electron microscopy in addressing the Ferrocyanide Safety Issue associated with Hanford Site high-level radioactive waste tanks. The status of the FY 1995 activities directed towards establishing facilities capable of providing SEM based micro-characterization of ferrocyanide tank wastes is described. A summary of key events in the SEM task over FY 1995 and target activities in FY 1996 are presented. A brief overview of the potential applications of computer controlled SEM analytical data in light of analyses of ferrocyanide simulants performed by an independent contractor is also presented

Fault tree synthesis for software design analysis of PLC based safety-critical systems

International Nuclear Information System (INIS)

Koo, S. R.; Cho, C. H.; Seong, P. H.

2006-01-01

As a software verification and validation should be performed for the development of PLC based safety-critical systems, a software safety analysis is also considered in line with entire software life cycle. In this paper, we propose a technique of software safety analysis in the design phase. Among various software hazard analysis techniques, fault tree analysis is most widely used for the safety analysis of nuclear power plant systems. Fault tree analysis also has the most intuitive notation and makes both qualitative and quantitative analyses possible. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Consequently, we can analyze the safety of software on the basis of fault tree synthesis. (authors)

Engineering approach to relative quantitative assessment of safety culture and related social issues in NPP operation

International Nuclear Information System (INIS)

Sivokon, V.; Gladyshev, M.; Malkin, S.

2005-01-01

The report is devoted to presentation of engineering approach and software tool developed for Safety Culture (SC) assessment as well as to the results of their implementation at Smolensk NPP. The engineering approach is logic evolution of the IAEA ASSET method broadly used at European NPPs in 90-s. It was implemented at Russian and other plants including Olkiluoto NPP in Finland. The approach allows relative quantitative assessing and trending the aspects of SC by the analysis of evens features and causes, calculation and trending corresponding indicators. At the same time plant's operational performances and related social issues, including efficiency of plant operation and personnel reliability, can be monitored. With the help of developed tool the joint team combined from personnel of Smolensk NPP and RRC 'Kurchatov Institute' ('KI') issued the SC self-assessment report, which identifies: families of recurrent events, main safety and operational problems ; their trends and importance to SC and plant efficiency; recommendations to enhance SC and operational performance

Quantitative Information Flow as Safety and Liveness Hyperproperties

Directory of Open Access Journals (Sweden)

Hirotoshi Yasuoka

2012-07-01

Full Text Available We employ Clarkson and Schneider's "hyperproperties" to classify various verification problems of quantitative information flow. The results of this paper unify and extend the previous results on the hardness of checking and inferring quantitative information flow. In particular, we identify a subclass of liveness hyperproperties, which we call "k-observable hyperproperties", that can be checked relative to a reachability oracle via self composition.

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

ATHENA/INTRA analyses for ITER, NSSR-2

International Nuclear Information System (INIS)

Shen, Kecheng; Eriksson, John; Sjoeberg, A.

1999-02-01

The present report is a summary report including thermal-hydraulic analyses made at Studsvik Eco and Safety AB for the ITER NSSR-2 safety documentation. The objective of the analyses was to reveal the safety characteristics of various heat transfer systems at specified operating conditions and to indicate the conditions for which there were obvious risks of jeopardising the structural integrity of the coolant systems. In the latter case also some analyses were made to indicate conceivable mitigating measures for maintaining the integrity.The analyses were primarily concerned with the First Wall and Divertor heat transfer systems. Several enveloping transients were analysed with associated specific flow and heat load boundary conditions. The analyses were performed with the ATHENA and INTRA codes

ATHENA/INTRA analyses for ITER, NSSR-2

Energy Technology Data Exchange (ETDEWEB)

Shen, Kecheng; Eriksson, John; Sjoeberg, A

1999-02-01

The present report is a summary report including thermal-hydraulic analyses made at Studsvik Eco and Safety AB for the ITER NSSR-2 safety documentation. The objective of the analyses was to reveal the safety characteristics of various heat transfer systems at specified operating conditions and to indicate the conditions for which there were obvious risks of jeopardising the structural integrity of the coolant systems. In the latter case also some analyses were made to indicate conceivable mitigating measures for maintaining the integrity.The analyses were primarily concerned with the First Wall and Divertor heat transfer systems. Several enveloping transients were analysed with associated specific flow and heat load boundary conditions. The analyses were performed with the ATHENA and INTRA codes 8 refs, 14 figs, 15 tabs

Quantitative Safety and Security Analysis from a Communication Perspective

Directory of Open Access Journals (Sweden)

Boris Malinowsky

2015-12-01

Full Text Available This paper introduces and exemplifies a trade-off analysis of safety and security properties in distributed systems. The aim is to support analysis for real-time communication and authentication building blocks in a wireless communication scenario. By embedding an authentication scheme into a real-time communication protocol for safety-critical scenarios, we can rely on the protocol’s individual safety and security properties. The resulting communication protocol satisfies selected safety and security properties for deployment in safety-critical use-case scenarios with security requirements. We look at handover situations in a IEEE 802.11 wireless setup between mobile nodes and access points. The trade-offs involve application-layer data goodput, probability of completed handovers, and effect on usable protocol slots, to quantify the impact of security from a lower-layer communication perspective on the communication protocols. The results are obtained using the network simulator ns-3.

Studies of CTNNBL1 and FDFT1 variants and measures of obesity: analyses of quantitative traits and case-control studies in 18,014 Danes

DEFF Research Database (Denmark)

Andreasen, Camilla Helene; Mogensen, Mette Sloth; Borch-Johnsen, Knut

2009-01-01

of obesity-related quantitative traits, and case-control studies in large study samples of Danes. METHODS: The FDFT1 rs7001819, CTNNBL1 rs6013029 and rs6020846 were genotyped, using TaqMan allelic discrimination, in a combined study sample comprising 18,014 participants ascertained from; the population...... and a previous study. FDFT1 rs7001819 showed no association with obesity, neither when analysing quantitative traits nor when performing case-control studies of obesity.......). The most significantly associating variants within CTNNBL1 including rs6013029 and rs6020846 were additionally confirmed to associate with morbid obesity in a French Caucasian case-control sample. The aim of this study was to investigate the impact of these three variants on obesity, through analyses...

HERMES docking/berthing system pilot study. Quantitative assessment

International Nuclear Information System (INIS)

Munoz Blasco, J.; Goicoechea Sanchez, F.J.

1993-01-01

This study falls within the framework of the incorporation of quantitative risk assessment to the activities planned for the ESA-HERMES project (ESA/ CNES). The main objective behind the study was the analysis and evaluation of the potential contribution of so-called probabilistic or quantitative safety analysis to the optimization of the safety development process for the systems carrying out the safety functions required by the new and complex HERMES Space Vehicle. For this purpose, a pilot study was considered a good start in quantitative safety assessments (QSA), as this approach has been frequently used in the past to establish a solid base in large-scale QSA application programs while avoiding considerable economic risks. It was finally decided to select the HERMES docking/berthing system with Man Tender Free Flyer as the case-study. This report describes the different steps followed in the study, along with the main insights obtained and the general conclusions drawn from the study results. (author)

Utilisation of best estimate system codes and best estimate methods in safety analyses of VVER reactors in the Czech Republic

International Nuclear Information System (INIS)

Macek, Jiri; Kral, Pavel

2010-01-01

The content of the presentation was as follows: Conservative versus best estimate approach, Brief description and selection of methodology, Description of uncertainty methods, Examples of the BE methodology. It is concluded that where BE computer codes are used, uncertainty and sensitivity analyses should be included; if best estimate codes + uncertainty are used, the safety margins increase; and BE + BSA is the next step in licensing analyses. (P.A.)

Safety strategy and safety analysis of nuclear power plants

International Nuclear Information System (INIS)

Franzen, L.F.

1976-01-01

The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the finding derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant, it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essentail for accident analyses, and the determination of the loads occurring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig.) [de

Preliminary Study on the Development of Quantitative Safety Culture Index

International Nuclear Information System (INIS)

Lee, Young Eal; Kim, Hun Sil; Ahn, Nam Sung

2005-01-01

Safety culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance. Because it needs to be recognized as the most significant consciousness to achieve the nuclear safety performance, Korean government and nuclear power generation company have tried to develop the practical method to improve the safety culture from the long term point view. In this study, based on the site interviews to define the potential issues on organizational behavior for the safe operation and the survey on the level of safety culture of occupied workers are conducted. Survey results are quantified as a few indicators of nuclear safety by the statistical method and it can be simulated by the dynamic modeling as time goes on. Currently index and dynamic modeling are still being developed, however, results can be used to suggest the long term strategy which safety is clearly integrated into all activities in the nuclear organization

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

International Nuclear Information System (INIS)

Blomquist, C.A.; Pierce, R.D.; Pedersen, D.R.; Ariman, T.

1977-01-01

The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. Thermal analyses were performed with the Argonne-modified version fo the general heat transfer code THTB, based on the instantaneous addition of 3200 0 K molten fuel with a decay heat of 9 W/gm and 1920 0 K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The stress analysis showed that the Inconel vessel would not fail from the pressure loading, it was also shown that brittle fracture of the tungsten liner from thermal gradients is unlikely. Therefore, the melt-down cup meets the structural design requirements. (Auth.)

Quantitative Prediction of Coalbed Gas Content Based on Seismic Multiple-Attribute Analyses

Directory of Open Access Journals (Sweden)

Renfang Pan

2015-09-01

Full Text Available Accurate prediction of gas planar distribution is crucial to selection and development of new CBM exploration areas. Based on seismic attributes, well logging and testing data we found that seismic absorption attenuation, after eliminating the effects of burial depth, shows an evident correlation with CBM gas content; (positive structure curvature has a negative correlation with gas content; and density has a negative correlation with gas content. It is feasible to use the hydrocarbon index (P*G and pseudo-Poisson ratio attributes for detection of gas enrichment zones. Based on seismic multiple-attribute analyses, a multiple linear regression equation was established between the seismic attributes and gas content at the drilling wells. Application of this equation to the seismic attributes at locations other than the drilling wells yielded a quantitative prediction of planar gas distribution. Prediction calculations were performed for two different models, one using pre-stack inversion and the other one disregarding pre-stack inversion. A comparison of the results indicates that both models predicted a similar trend for gas content distribution, except that the model using pre-stack inversion yielded a prediction result with considerably higher precision than the other model.

Analysing supercritical water reactor's (SCWR's) special safety systems using probabilistic tools

International Nuclear Information System (INIS)

Ituen, I.; Novog, D.R.

2011-01-01

The next generation of reactors, termed Generation IV, has very attractive features -- its superior safety characteristics, high thermal efficiency, and fuel cycle sustainability. A key element of the Generation IV designs is the improvement in safety, which in turn requires improvements in safety system performance and reliability, as well as a reduction in initiating event frequencies. This study compares the response of the systems important to safety in the CANDU-Supercritical Water Reactor to those of the generic CANDU under a main steamline break accident and loss of forced circulation events -- to quantify the improvements in safety for the pre-conceptual CANDU SCWR design. Probabilistic safety analysis is the tool used in this study to test the behavior of the pre- conceptual design during these events. (author)

Labor unions and safety climate: perceived union safety values and retail employee safety outcomes.

Science.gov (United States)

Sinclair, Robert R; Martin, James E; Sears, Lindsay E

2010-09-01

Although trade unions have long been recognized as a critical advocate for employee safety and health, safety climate research has not paid much attention to the role unions play in workplace safety. We proposed a multiple constituency model of workplace safety which focused on three central safety stakeholders: top management, ones' immediate supervisor, and the labor union. Safety climate research focuses on management and supervisors as key stakeholders, but has not considered whether employee perceptions about the priority their union places on safety contributes contribute to safety outcomes. We addressed this gap in the literature by investigating unionized retail employee (N=535) perceptions about the extent to which their top management, immediate supervisors, and union valued safety. Confirmatory factor analyses demonstrated that perceived union safety values could be distinguished from measures of safety training, workplace hazards, top management safety values, and supervisor values. Structural equation analyses indicated that union safety values influenced safety outcomes through its association with higher safety motivation, showing a similar effect as that of supervisor safety values. These findings highlight the need for further attention to union-focused measures related to workplace safety as well as further study of retail employees in general. We discuss the practical implications of our findings and identify several directions for future safety research. 2009 Elsevier Ltd. All rights reserved.

Experimental Studies of quantitative evaluation using HPLC and safety of Bee Venom Acupuncture

Directory of Open Access Journals (Sweden)

Seong Bong Jang

2006-02-01

Full Text Available Objectives : This study was conducted to carry out quantitative evaluation and safety of Bee Venom Acupuncture. Methods : Content analysis was done using HPLC, measurement of , and histological observations were made on the skin and muscles. Results : 1. According to HPLC analysis, each BVA-1 contained approximately , and BVA-2 contained approximately . But the volume of coating was so minute, slight difference exists between each needle. 2. LD50 of mouse with BVA-1 was 16 counts and this is equivalent to 640 needles/kg, making Bee Venom Acupuncture safe treatment apparatus. 3. Regardless of the number of needles, there was no sign of blood stasis or inflammation detected on the skin and muscle tissues. Conclusion : Above results indicate that the Bee Venom Acupuncture can complement shortcomings of syringe usage as a part of Oriental medicine treatment, but extensive researches should be done for further verification.

Human factors in safety assessment. Safety culture assessment

International Nuclear Information System (INIS)

Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

1996-01-01

This paper analyses the present conditions and problems in enterprises safety assessment, and introduces the characteristics and effects of safety culture. The authors think that safety culture must be used as a 'soul' to form the pattern of modern safety management. Furthermore, they propose that the human safety and synthetic safety management assessment in a system should be changed into safety culture assessment. Finally, the assessment indicators are discussed

Software reliability for safety-critical applications

International Nuclear Information System (INIS)

Everett, B.; Musa, J.

1994-01-01

In this talk, the authors address the question open-quotes Can Software Reliability Engineering measurement and modeling techniques be applied to safety-critical applications?close quotes Quantitative techniques have long been applied in engineering hardware components of safety-critical applications. The authors have seen a growing acceptance and use of quantitative techniques in engineering software systems but a continuing reluctance in using such techniques in safety-critical applications. The general case posed against using quantitative techniques for software components runs along the following lines: safety-critical applications should be engineered such that catastrophic failures occur less frequently than one in a billion hours of operation; current software measurement/modeling techniques rely on using failure history data collected during testing; one would have to accumulate over a billion operational hours to verify failure rate objectives of about one per billion hours

Overcooling transient selection and thermal hydraulic analyses of the Loviisa PTS assessments

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, H [IVO Power Engineering Ltd, Vantaa (Finland)

1997-09-01

This paper describes transients selection and thermal hydraulic analyses of various PTS assessment studies performed for the pressure vessels of the Loviisa WWER-reactors. Deterministic analyses have been performed in various stages of the PTS studies and they have always made the formal basis for design and licensing of the reactor pressure vessel. The integrated, probabilistic PTS study was carried out to give an overview of the severity of all different PTS sequences, and give a quantitative estimate of the importance of the PTS issues in relation to the overall safety of the plant. Later, the sequences including external flooding of the pressure vessels were added to the PTS assessment. Thermal recovery annealing of the Loviisa 1 reactor pressure vessel took place during refuelling outage in 1996. (author). 10 refs, 4 figs, 3 tabs.

The roles of emotional intelligence, interpersonal skill, and transformational leadership on improving construction safety performance

Directory of Open Access Journals (Sweden)

Riza Yosia Sunindijo

2013-09-01

Full Text Available Due to the characteristics of the constructionindustry, human skills are essential for working with and through others inmanaging safety. Research has shown that emotional intelligence, interpersonalskill, and transformational leadership are human skill components that generatesuperior performance in today’s workplace. The aim of this research is toinvestigate the influence of project management personnel’s human skills on theimplementation of safety management tasks and development of safety climate inconstruction projects. The structural equation modelling (SEM method wasapplied to analyse the quantitative data collected and establishinterrelationship among the research variables. The results indicate thatemotional intelligence is a key factor for developing interpersonal skill andtransformational leadership, and for implementing safety management tasks whichleads to the development of safety climate. This research also found thatinterpersonal skill is needed for becoming transformational leaders whocontribute to the development of safety climate.

Preliminary safety analysis of the HTTR-IS nuclear hydrogen production system

International Nuclear Information System (INIS)

Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Tachibana, Yukio; Sakaba, Nariaki

2010-06-01

Japan Atomic Energy Agency is planning to demonstrate hydrogen production by thermochemical water-splitting IS process utilizing heat from the high-temperature gas-cooled reactor HTTR (HTTR-IS system). The previous study identified that the HTTR modification due to the coupling of hydrogen production plant requires an additional safety review since the scenario and quantitative values of the evaluation items would be altered from the original HTTR safety review. Hence, preliminary safety analyses are conducted by using the system analysis code. Calculation results showed that evaluation items such as a coolant pressure, temperatures of heat transfer tubes at the pressure boundary, etc., did not exceed allowable values. Also, the peak fuel temperature did not exceed allowable value and therefore the reactor core was not damaged and cooled sufficiently. This report compiles calculation conditions, event scenarios and the calculation results of the preliminary safety analysis. (author)

Safety analyses of potential exposure in medical irradiation plants by Fuzzy Fault Tree

International Nuclear Information System (INIS)

Casamirra, Maddalena; Castiglia, Francesco; Giardina, Mariarosa; Tomarchio, Elio

2008-01-01

The results of Fuzzy Fault Tree (FFT) analyses of various accidental scenarios, which involve the operators in potential exposures inside an High Dose Rate (HDR) remote after-loading systems for use in brachytherapy, are reported. To carry out fault tree analyses by means of fuzzy probabilities, the TREEZZY2 computer code is used. Moreover, the HEART (Human Error Assessment and Reduction Technique) model, properly modified on the basis of the fuzzy approach, has been employed to assess the impact of performances haping and error-promoting factors in the context of the accidental events. The assessment of potential dose values for some identified accidental scenarios allows to consider, for a particular event, a fuzzy uncertainty range in potential dose estimate. The availability of lower and upper limits allows evaluating the possibility of optimization of the installation from the point of view of radiation protection. The adequacy of the training and information program for staff and patients (and their family members) and the effectiveness of behavioural rules and safety procedures were tested. Some recommendations on procedures and equipment to reduce the risk of radiological exposure are also provided. (author)

Quantifying system safety: A comparison of the SBOAT & Safety Barrier Manager tools

DEFF Research Database (Denmark)

Hansen, Zaza Nadja Lee; Duijm, Nijs Jan; Markert, Frank

2015-01-01

This paper presents two software tools for analyzing safety risks, SBOAT (Stochastic BPMN Optimisation and Analysis Tool) and SBM (SafetyBarrierManagerr). SBOAT employs principles from stochastic model checking to allow for the quantitative verification of workflows. SBM supports the creation...

Comparison of multipoint linkage analyses for quantitative traits in the CEPH data: parametric LOD scores, variance components LOD scores, and Bayes factors.

Science.gov (United States)

Sung, Yun Ju; Di, Yanming; Fu, Audrey Q; Rothstein, Joseph H; Sieh, Weiva; Tong, Liping; Thompson, Elizabeth A; Wijsman, Ellen M

2007-01-01

We performed multipoint linkage analyses with multiple programs and models for several gene expression traits in the Centre d'Etude du Polymorphisme Humain families. All analyses provided consistent results for both peak location and shape. Variance-components (VC) analysis gave wider peaks and Bayes factors gave fewer peaks. Among programs from the MORGAN package, lm_multiple performed better than lm_markers, resulting in less Markov-chain Monte Carlo (MCMC) variability between runs, and the program lm_twoqtl provided higher LOD scores by also including either a polygenic component or an additional quantitative trait locus.

Evaluation of geological documents available for provisional safety analyses of potential sites for nuclear waste repositories - Are additional geological investigations needed?

International Nuclear Information System (INIS)

2010-10-01

The procedure for selecting repository sites for all categories of radioactive waste in Switzerland is defined in the conceptual part of the Sectoral Plan for Deep Geological Repositories, which foresees a selection of sites in three stages. In Stage I, Nagra proposed geological siting regions based on criteria relating to safety and engineering feasibility. The Swiss Government (the Federal Council) is expected to decide on the siting proposals in 2011. The objective of Stage 2 is to prepare proposals for the location of the surface facilities within the planning perimeters defined by the Federal Council in its decision on Stage 1 and to identify potential sites. Nagra also has to carry out a provisional safety analysis for each site and a safety-based comparison of the sites. Based on this, and taking into account the results of the socio-economic-ecological impact studies, Nagra then has to propose at least two sites for each repository type to be carried through to Stage 3. The proposed sites will then be investigated in more detail in Stage 3 to ensure that the selection of the sites for the General Licence Applications is well founded. In order to realise the objectives of the upcoming Stage 2, the state of knowledge of the geological conditions at the sites has to be sufficient to perform the provisional safety analyses. Therefore, in preparation for Stage 2, the conceptual part of the Sectoral Plan requires Nagra to clarify the need for additional investigations aimed at providing input for the provisional safety analyses. The purpose of the present report is to document Nagra's technical-scientific assessment of this need. The focus is on evaluating the geological information based on processes and parameters that are relevant for safety and engineering feasibility. In evaluating the state of knowledge the key question is whether additional information could lead to a different decision regarding the selection of the sites to be carried through to Stage 3

Determining quantitative road safety targets by applying statistical prediction techniques and a multi-stage adjustment procedure.

Science.gov (United States)

Wittenberg, P; Sever, K; Knoth, S; Sahin, N; Bondarenko, J

2013-01-01

Due to substantial progress made in road safety in the last ten years, the European Union (EU) renewed the ambitious agreement of halving the number of persons killed on the roads within the next decade. In this paper we develop a method that aims at finding an optimal target for each nation, in terms of being as achievable as possible, and with the cumulative EU target being reached. Targets as an important component in road safety policy are given as reduction rate or as absolute number of road traffic deaths. Determination of these quantitative road safety targets (QRST) is done by a top-down approach, formalized in a multi-stage adjustment procedure. Different QRST are derived under consideration of recent research. The paper presents a method to break the national target further down to regional targets in case of the German Federal States. Generalized linear models are fitted to data in the period 1991-2010. Our model selection procedure chooses various models for the EU and solely log-linear models for the German Federal States. If the proposed targets for the EU Member States are attained, the sum of fatalities should not exceed the total value of 15,465 per year by 2020. Both, the mean level and the range of mortality rates within the EU could be lowered from 28-113 in 2010 to 17-41 per million inhabitants in 2020. This study provides an alternative to the determination of safety targets by political commitments only, taking the history of road fatalities trends and population into consideration. Copyright © 2012 Elsevier Ltd. All rights reserved.

Stability Test and Quantitative and Qualitative Analyses of the Amino Acids in Pharmacopuncture Extracted from Scolopendra subspinipes mutilans

Science.gov (United States)

Cho, GyeYoon; Han, KyuChul; Yoon, JinYoung

2015-01-01

Objectives: Scolopendra subspinipes mutilans (S. subspinipes mutilans) is known as a traditional medicine and includes various amino acids, peptides and proteins. The amino acids in the pharmacopuncture extracted from S. subspinipes mutilans by using derivatization methods were analyzed quantitatively and qualitatively by using high performance liquid chromatography (HPLC) over a 12 month period to confirm its stability. Methods: Amino acids of pharmacopuncture extracted from S. subspinipes mutilans were derived by using O-phthaldialdehyde (OPA) & 9-fluorenyl methoxy carbonyl chloride (FMOC) reagent and were analyzed using HPLC. The amino acids were detected by using a diode array detector (DAD) and a fluorescence detector (FLD) to compare a mixed amino acid standard (STD) to the pharmacopuncture from centipedes. The stability tests on the pharmacopuncture from centipedes were done using HPLC for three conditions: a room temperature test chamber, an acceleration test chamber, and a cold test chamber. Results: The pharmacopuncture from centipedes was prepared by using the method of the Korean Pharmacopuncture Institute (KPI) and through quantitative analyses was shown to contain 9 amino acids of the 16 amino acids in the mixed amino acid STD. The amounts of the amino acids in the pharmacopuncture from centipedes were 34.37 ppm of aspartate, 123.72 ppm of arginine, 170.63 ppm of alanine, 59.55 ppm of leucine and 57 ppm of lysine. The relative standard deviation (RSD %) results for the pharmacopuncture from centipedes had a maximum value of 14.95% and minimum value of 1.795% on the room temperature test chamber, the acceleration test chamber and the cold test chamber stability tests. Conclusion: Stability tests on and quantitative and qualitative analyses of the amino acids in the pharmacopuncture extracted from centipedes by using derivatization methods were performed by using HPLC. Through research, we hope to determine the relationship between time and the

Cloning, characterisation, and comparative quantitative expression analyses of receptor for advanced glycation end products (RAGE) transcript forms.

Science.gov (United States)

Sterenczak, Katharina A; Willenbrock, Saskia; Barann, Matthias; Klemke, Markus; Soller, Jan T; Eberle, Nina; Nolte, Ingo; Bullerdiek, Jörn; Murua Escobar, Hugo

2009-04-01

RAGE is a member of the immunoglobulin superfamily of cell surface molecules playing key roles in pathophysiological processes, e.g. immune/inflammatory disorders, Alzheimer's disease, diabetic arteriosclerosis and tumourigenesis. In humans 19 naturally occurring RAGE splicing variants resulting in either N-terminally or C-terminally truncated proteins were identified and are lately discussed as mechanisms for receptor regulation. Accordingly, deregulation of sRAGE levels has been associated with several diseases e.g. Alzheimer's disease, Type 1 diabetes, and rheumatoid arthritis. Administration of recombinant sRAGE to animal models of cancer blocked tumour growth successfully. In spite of its obvious relationship to cancer and metastasis data focusing sRAGE deregulation and tumours is rare. In this study we screened a set of tumours, healthy tissues and various cancer cell lines for RAGE splicing variants and analysed their structure. Additionally, we analysed the ratio of the mainly found transcript variants using quantitative Real-Time PCR. In total we characterised 24 previously not described canine and 4 human RAGE splicing variants, analysed their structure, classified their characteristics, and derived their respective protein forms. Interestingly, the healthy and the neoplastic tissue samples showed in majority RAGE transcripts coding for the complete receptor and transcripts showing insertions of intron 1.

Total and species-specific quantitative analyses of trace elements in sediment by isotope dilution inductively coupled plasma mass spectrometry

International Nuclear Information System (INIS)

Inagaki, Kazumi; Takatsu, Akiko; Yarita, Takashi; Okamoto, Kensaku; Chiba, Koichi

2009-01-01

Isotope dilution inductively coupled plasma mass spectrometry (ID-ICP-MS) is one of the reliable methods for total and species-specific quantitative analysis of trace elements. However, several technical problems (e.g. spectral interference caused from sample constituents) should be overcome to obtain reliable analytical results when environmental samples are analyzed by ID-ICP-MS. In our laboratory, various methods based on ID-ICP-MS have been investigated for reliable quantitative analyses of trace elements in environmental samples. In this paper, coprecipitate separation/ID-ICP-MS for the determination of trace elements in sediment, cation exchange disk filtration/ID-ICP-MS for the determination of selenium in sediment, species-specific ID-ICP-MS using 118 Sn/labeled organotin compounds for the determination of butyltins and phenyltins, and the application of the ID-ICP-MS methods to the certification of sediment reference materials are described. (author)

Systems reliability analyses and risk analyses for the licencing procedure under atomic law

International Nuclear Information System (INIS)

Berning, A.; Spindler, H.

1983-01-01

For the licencing procedure under atomic law in accordance with Article 7 AtG, the nuclear power plant as a whole needs to be assessed, plus the reliability of systems and plant components that are essential to safety are to be determined with probabilistic methods. This requirement is the consequence of safety criteria for nuclear power plants issued by the Home Department (BMI). Systems reliability studies and risk analyses used in licencing procedures under atomic law are identified. The stress is on licencing decisions, mainly for PWR-type reactors. Reactor Safety Commission (RSK) guidelines, examples of reasoning in legal proceedings and arguments put forth by objectors are also dealt with. Correlations are shown between reliability analyses made by experts and licencing decisions by means of examples. (orig./HP) [de

Safety margins in deterministic safety analysis

International Nuclear Information System (INIS)

Viktorov, A.

2011-01-01

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

Laser Beam Focus Analyser

DEFF Research Database (Denmark)

Nielsen, Peter Carøe; Hansen, Hans Nørgaard; Olsen, Flemming Ove

2007-01-01

the obtainable features in direct laser machining as well as heat affected zones in welding processes. This paper describes the development of a measuring unit capable of analysing beam shape and diameter of lasers to be used in manufacturing processes. The analyser is based on the principle of a rotating......The quantitative and qualitative description of laser beam characteristics is important for process implementation and optimisation. In particular, a need for quantitative characterisation of beam diameter was identified when using fibre lasers for micro manufacturing. Here the beam diameter limits...... mechanical wire being swept through the laser beam at varying Z-heights. The reflected signal is analysed and the resulting beam profile determined. The development comprised the design of a flexible fixture capable of providing both rotation and Z-axis movement, control software including data capture...

A review of significant events analysed in general practice: implications for the quality and safety of patient care

Directory of Open Access Journals (Sweden)

Bradley Nick

2009-09-01

Full Text Available Abstract Background Significant event analysis (SEA is promoted as a team-based approach to enhancing patient safety through reflective learning. Evidence of SEA participation is required for appraisal and contractual purposes in UK general practice. A voluntary educational model in the west of Scotland enables general practitioners (GPs and doctors-in-training to submit SEA reports for feedback from trained peers. We reviewed reports to identify the range of safety issues analysed, learning needs raised and actions taken by GP teams. Method Content analysis of SEA reports submitted in an 18 month period between 2005 and 2007. Results 191 SEA reports were reviewed. 48 described patient harm (25.1%. A further 109 reports (57.1% outlined circumstances that had the potential to cause patient harm. Individual 'error' was cited as the most common reason for event occurrence (32.5%. Learning opportunities were identified in 182 reports (95.3% but were often non-specific professional issues not shared with the wider practice team. 154 SEA reports (80.1% described actions taken to improve practice systems or professional behaviour. However, non-medical staff were less likely to be involved in the changes resulting from event analyses describing patient harm (p Conclusion The study provides some evidence of the potential of SEA to improve healthcare quality and safety. If applied rigorously, GP teams and doctors in training can use the technique to investigate and learn from a wide variety of quality issues including those resulting in patient harm. This leads to reported change but it is unclear if such improvement is sustained.

Quantifying Safety Margin Using the Risk-Informed Safety Margin Characterization (RISMC)

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Nakayama, Marvin

2015-04-26

The Risk-Informed Safety Margin Characterization (RISMC), developed by Idaho National Laboratory as part of the Light-Water Reactor Sustainability Project, utilizes a probabilistic safety margin comparison between a load and capacity distribution, rather than a deterministic comparison between two values, as is usually done in best-estimate plus uncertainty analyses. The goal is to determine the failure probability, or in other words, the probability of the system load equaling or exceeding the system capacity. While this method has been used in pilot studies, there has been little work conducted investigating the statistical significance of the resulting failure probability. In particular, it is difficult to determine how many simulations are necessary to properly characterize the failure probability. This work uses classical (frequentist) statistics and confidence intervals to examine the impact in statistical accuracy when the number of simulations is varied. Two methods are proposed to establish confidence intervals related to the failure probability established using a RISMC analysis. The confidence interval provides information about the statistical accuracy of the method utilized to explore the uncertainty space, and offers a quantitative method to gauge the increase in statistical accuracy due to performing additional simulations.

Preliminary safety evaluation for CSR1000 with passive safety system

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

2014-01-01

Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

A Quantitative Microbiological Risk Assessment for Salmonella in Pigs for the European Union

DEFF Research Database (Denmark)

Snary, Emma L.; Swart, Arno N.; Simons, Robin R. L.

2016-01-01

,000 and 1 in 10 million servings given consumption of one of the three product types considered (pork cuts, minced meat, and fermented ready‐to‐eat sausages). Further analyses of the farm‐to‐consumption QMRA suggest that the vast majority of human risk derives from infected pigs with a high concentration......A farm‐to‐consumption quantitative microbiological risk assessment (QMRA) for Salmonella in pigs in the European Union has been developed for the European Food Safety Authority. The primary aim of the QMRA was to assess the impact of hypothetical reductions of slaughter‐pig prevalence...

Development of SAGE, A computer code for safety assessment analyses for Korean Low-Level Radioactive Waste Disposal

International Nuclear Information System (INIS)

Zhou, W.; Kozak, Matthew W.; Park, Joowan; Kim, Changlak; Kang, Chulhyung

2002-01-01

This paper describes a computer code, called SAGE (Safety Assessment Groundwater Evaluation) to be used for evaluation of the concept for low-level waste disposal in the Republic of Korea (ROK). The conceptual model in the code is focused on releases from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. Doses can be calculated for several biosphere systems including drinking contaminated groundwater, and subsequent contamination of foods, rivers, lakes, or the ocean by that groundwater. The flexibility of the code will permit both generic analyses in support of design and site development activities, and straightforward modification to permit site-specific and design-specific safety assessments of a real facility as progress is made toward implementation of a disposal site. In addition, the code has been written to easily interface with more detailed codes for specific parts of the safety assessment. In this way, the code's capabilities can be significantly expanded as needed. The code has the capability to treat input parameters either deterministic ally or probabilistic ally. Parameter input is achieved through a user-friendly Graphical User Interface.

Gas-cooled fast reactor safety - and overview and status of the U.S. program

International Nuclear Information System (INIS)

Torri, A.; Buttemer, D.R.

1981-01-01

In the revised GCFR Safety Program Plan a quantitative risk limit line has been adopted to establish requirements for the safety related functions and systems. The risk limit line is derived from an interpretation of NRC established licensing requirements, including those for LMFBR's. Multiple barriers to the progression of accident sequences are defined in the form of six Lines of Protection (LOPs). LOPs-1 to 3 are dedicated to accident prevention and represent the normal operating systems, the dedicated safety systems and the inherent design features, respectively. LOPs-4 to 6 are dedicated to the mitigation of core melt accident consequences and include in-vessel accident containment, secondary containment integrity and radiological attenuation, respectively. Cumulative frequency limits and consequence limits are established for each LOP. Design features associated with each LOP are described and the results of supporting safety analyses are summarized. (author)

Efficacy and Safety Extrapolation Analyses for Atomoxetine in Young Children with Attention-Deficit/Hyperactivity Disorder.

Science.gov (United States)

Upadhyaya, Himanshu; Kratochvil, Christopher; Ghuman, Jaswinder; Camporeale, Angelo; Lipsius, Sarah; D'Souza, Deborah; Tanaka, Yoko

2015-12-01

This extrapolation analysis qualitatively compared the efficacy and safety profile of atomoxetine from Lilly clinical trial data in 6-7-year-old patients with attention-deficit/hyperactivity disorder (ADHD) with that of published literature in 4-5-year-old patients with ADHD (two open-label [4-5-year-old patients] and one placebo-controlled study [5-year-old patients]). The main efficacy analyses included placebo-controlled Lilly data and the placebo-controlled external study (5-year-old patients) data. The primary efficacy variables used in these studies were the ADHD Rating Scale-IV Parent Version, Investigator Administered (ADHD-RS-IV-Parent:Inv) total score, or the Swanson, Nolan and Pelham (SNAP-IV) scale score. Safety analyses included treatment-emergent adverse events (TEAEs) and vital signs. Descriptive statistics (means, percentages) are presented. Acute atomoxetine treatment improved core ADHD symptoms in both 6-7-year-old patients (n=565) and 5-year-old patients (n=37) (treatment effect: -10.16 and -7.42). In an analysis of placebo-controlled groups, the mean duration of exposure to atomoxetine was ∼ 7 weeks for 6-7-year-old patients and 9 weeks for 5-year-old patients. Decreased appetite was the most common TEAE in atomoxetine-treated patients. The TEAEs observed at a higher rate in 5-year-old versus 6-7-year-old patients were irritability (36.8% vs. 3.6%) and other mood-related events (6.9% each vs. atomoxetine may improve ADHD symptoms, but possibly to a lesser extent than in older children, with some adverse events occurring at a higher rate in 5-year-old patients.

Perceived safety management practices in the logistics sector.

Science.gov (United States)

Auyong, Hui-Nee; Zailani, Suhaiza; Surienty, Lilis

2016-03-09

Malaysia's progress on logistics has been slowed to keep pace with its growth in trade. The Government has been pressing companies to improve the safety of their activities in order to reduce society's loss due to occupational accidents and illnesses. Occupational safety and health is a crucial part of a workplace because every worker has to take care of his/her own safety and health. The main occupational safety and health (OSH) national policy in Malaysia is the enactment of the Occupational Safety and Health Act (OSHA) 1994. Only those companies which have excellent health and safety care have good quality and productive employees. This study investigated safety management practices in the logistics sector. The present study is concerned with the human factors to safety in the logistics industry. The authors examined the perceived safety management practices of workers in the logistics sector. The purpose was to identify the perception of safety management practices of Malaysian logistics personnel. Survey questionnaires were distributed to assess logistics personnel about management commitment. The quantitative method using the availability sampling method was applied. The data gathered from the survey were analysed using SPSS software. The responses to the survey were rated according to the Likert scale type, with '1' indicating strongly disagree and '5' indicating strongly agree. One hundred and three employees of logistics functions completed the survey. The highest mean scores were found for fire apparatus, prioritisation of safety, and safety policy. The results from this study also emphasise the importance of the management's commitment in enhancing workplace safety. Specifically, companies should maintain good relations between the employer and the employee to help reduce workplace injuries.

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

Bokhari, I.H.; Israr, M.; Pervez, S.

1999-01-01

Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

Safety systems I/C reliability analysis of the Kozloduy NPP units 5 and 6; Analiz nadezhnosti KIP i sistem bezopasnosti pyatogo i shestogo bloka AEhS `Kozloduy`

Energy Technology Data Exchange (ETDEWEB)

Marinova, B [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

The purpose of the analysis is to assess the safety systems I/C equipment reliability of the Kozloduy-5 and the Kozloduy-6 reactors. The assessment of quantitative and qualitative effect of control systems unavailability on the safety systems unavailability is performed. The analysis is limited to the following systems: sprinkler management, low pressure emergency spray, emergency injection of boric acid, hydro accumulators, pressure compensator and compressed air. The code for probabilistic safety assessment PSAPACK has been used in analysis. Fault trees for all analysed safety systems have been constructed. Results indicates a high reliability of the safety systems management.

Quantification of human reliability in probabilistic safety assessment

International Nuclear Information System (INIS)

Hirschberg, S.; Dankg, Vinh N.

1996-01-01

Human performance may substantially influence the reliability and safety of complex technical systems. For this reason, Human Reliability Analysis (HRA) constitutes an important part of Probabilistic Safety Assessment (PSAs) or Quantitative Risk Analyses (QRAs). The results of these studies as well as analyses of past accidents and incidents clearly demonstrate the importance of human interactions. The contribution of human errors to the core damage frequency (CDF), as estimated in the Swedish nuclear PSAs, are between 15 and 88%. A survey of the FRAs in the Swiss PSAs shows that also for the Swiss nuclear power plants the estimated HE contributions are substantial (49% of the CDF due to internal events in the case of Beznau and 70% in the case of Muehleberg; for the total CDF, including external events, 25% respectively 20%). Similar results can be extracted from the PSAs carried out for French, German, and US plants. In PSAs or QRAs, the adequate treatment of the human interactions with the system is a key to the understanding of accident sequences and their relative importance to overall risk. The main objectives of HRA are: first, to ensure that the key human interactions are systematically identified and incorporated into the safety analysis in a traceable manner, and second, to quantify the probabilities of their success and failure. Adopting a structured and systematic approach to the assessment of human performance makes it possible to provide greater confidence that the safety and availability of human-machine systems is not unduly jeopardized by human performance problems. Section 2 discusses the different types of human interactions analysed in PSAs. More generally, the section presents how HRA fits in the overall safety analysis, that is, how the human interactions to be quantified are identified. Section 3 addresses the methods for quantification. Section 4 concludes the paper by presenting some recommendations and pointing out the limitations of the

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

NARCIS (Netherlands)

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

2011-01-01

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

The influence of the design matrix on treatment effect estimates in the quantitative analyses of single-subject experimental design research.

Science.gov (United States)

Moeyaert, Mariola; Ugille, Maaike; Ferron, John M; Beretvas, S Natasha; Van den Noortgate, Wim

2014-09-01

The quantitative methods for analyzing single-subject experimental data have expanded during the last decade, including the use of regression models to statistically analyze the data, but still a lot of questions remain. One question is how to specify predictors in a regression model to account for the specifics of the design and estimate the effect size of interest. These quantitative effect sizes are used in retrospective analyses and allow synthesis of single-subject experimental study results which is informative for evidence-based decision making, research and theory building, and policy discussions. We discuss different design matrices that can be used for the most common single-subject experimental designs (SSEDs), namely, the multiple-baseline designs, reversal designs, and alternating treatment designs, and provide empirical illustrations. The purpose of this article is to guide single-subject experimental data analysts interested in analyzing and meta-analyzing SSED data. © The Author(s) 2014.

Reactor safety research and safety technology. Pt. 2

International Nuclear Information System (INIS)

Theenhaus, R.; Wolters, J.

1987-01-01

The state of HTR safety research work reached permits a comprehensive and reliable answer to be given to questions which have been raised by the reactor accident at Chernobyl, regarding HTR safety. Together with the probability safety analyses, the way to a safety concept suitable for an HTR is cleared; instructions are given for design optimisation with regard to safety technique and economy. The consequences of a graphite fire, the neutron physics design and the consequenes of the lack of a safety containment are briefly described. (DG) [de

Food Safety and Quality Control: Hints from Proteomics

Directory of Open Access Journals (Sweden)

Angelo D'Alessandro

2012-01-01

Full Text Available Over the last decade, proteomics has been successfully applied to the study of quality control in production processes of food (including meat, wine and beer, transgenic plants and milk and food safety (screening for food-derived pathogens. Indeed, food quality and safety and their influence on the health of end consumers have growingly become a founding principle in the international agenda of health organizations. The application of proteomics in food science was at first characterized by exploratory analyses of food of various origin (bovine, swine, chicken or lamb meat, but also transgenic food such as genetically modified maize, for example and beverages (beer, wine, in parallel to the genomic and transcriptomic approaches seeking determination of quantitative trait loci. In the last few years, technical improvements such as microbial biotyping strategies have growingly allowed proteomicists to address the safety issue as well. The newly introduced technical improvements (instrumentation characterized by higher sensitivity such as mass spectrometers have paved the way for the individuation of food-contaminating pathogens in a fast and efficient workflow which is mandatory in industrial food production chains.

Expermental Studies of quantitative evaluation using HPLC and safety of Sweet Bee Venom

Directory of Open Access Journals (Sweden)

Ki Rok Kwon

2007-06-01

Full Text Available Objectives : This study was conducted to carry out quantitative evaluation and safety of Sweet Bee Venom. Methods : Content analysis was done using HPLC, measurement of LD50 was conducted intravenous, subcutaneous, and intra-muscular injection to the ICR mice. Results : 1. According to HPLC analysis, removal of the enzymes containing phospholipase A2 was successfully rendered on Sweet Bee Venom. And analyzing melittin content, Sweet Bee Venom contained 12% more melittin than Bee Venom. 2. LD50 of ICR mice with Sweet Bee Venom was more than 20mg/kg in subcutaneous injection and intravenous injection, between 15mg/kg and 20mg/kg in muscular injection. 3. LD50 of ICR mice with Bee Venom was between 6 and 9mg/kg in subcutaneous injection and intravenous injection, and more than 9mg/kg in muscular injection. Conclusion : Above results indicate that Sweet Bee Venom was more safe than Bee Venom and the process of removing enzymes was well rendered in Sweet Bee Venom.

Prioritization of generic safety issues

International Nuclear Information System (INIS)

Emrit, R.; Minners, W.; VanderMolen, H.

1983-12-01

This report presents the priority rankings for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The report focuses on the prioritization of generic safety issues. Issues primarily concerned with the licensing process or environmental protection and not directly related to safety have been excluded from prioritization. The prioritized issues include: TMI Action Plan items under development; previously proposed issues covered by Task Action Plans, except issues designated at Unresolved Safety Issues (USIs) which had already been assigned high priority; and newly-proposed issues. Future supplements to this report will include the prioritization of additional issues. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolutions of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative

IT-CARES: an interactive tool for case-crossover analyses of electronic medical records for patient safety.

Science.gov (United States)

Caron, Alexandre; Chazard, Emmanuel; Muller, Joris; Perichon, Renaud; Ferret, Laurie; Koutkias, Vassilis; Beuscart, Régis; Beuscart, Jean-Baptiste; Ficheur, Grégoire

2017-03-01

The significant risk of adverse events following medical procedures supports a clinical epidemiological approach based on the analyses of collections of electronic medical records. Data analytical tools might help clinical epidemiologists develop more appropriate case-crossover designs for monitoring patient safety. To develop and assess the methodological quality of an interactive tool for use by clinical epidemiologists to systematically design case-crossover analyses of large electronic medical records databases. We developed IT-CARES, an analytical tool implementing case-crossover design, to explore the association between exposures and outcomes. The exposures and outcomes are defined by clinical epidemiologists via lists of codes entered via a user interface screen. We tested IT-CARES on data from the French national inpatient stay database, which documents diagnoses and medical procedures for 170 million inpatient stays between 2007 and 2013. We compared the results of our analysis with reference data from the literature on thromboembolic risk after delivery and bleeding risk after total hip replacement. IT-CARES provides a user interface with 3 columns: (i) the outcome criteria in the left-hand column, (ii) the exposure criteria in the right-hand column, and (iii) the estimated risk (odds ratios, presented in both graphical and tabular formats) in the middle column. The estimated odds ratios were consistent with the reference literature data. IT-CARES may enhance patient safety by facilitating clinical epidemiological studies of adverse events following medical procedures. The tool's usability must be evaluated and improved in further research. © The Author 2016. Published by Oxford University Press on behalf of the American Medical Informatics Association.

Probabilistic evaluation of scenarios in long-term safety analyses. Results of the project ISIBEL; Probabilistische Bewertung von Szenarien in Langzeitsicherheitsanalysen. Ergebnisse des Vorhabens ISIBEL

Energy Technology Data Exchange (ETDEWEB)

Buhmann, Dieter; Becker, Dirk-Alexander; Laggiard, Eduardo; Ruebel, Andre; Spiessl, Sabine; Wolf, Jens

2016-07-15

In the frame of the project ISIBEL deterministic analyses on the radiological consequences of several possible developments of the final repository were performed (VSG: preliminary safety analysis of the site Gorleben). The report describes the probabilistic evaluation of the VSG scenarios using uncertainty and sensitivity analyses. It was shown that probabilistic analyses are important to evaluate the influence of uncertainties. The transfer of the selected scenarios in computational cases and the used modeling parameters are discussed.

A fuzzy-logic-based approach to qualitative safety modelling for marine systems

International Nuclear Information System (INIS)

Sii, H.S.; Ruxton, Tom; Wang Jin

2001-01-01

Safety assessment based on conventional tools (e.g. probability risk assessment (PRA)) may not be well suited for dealing with systems having a high level of uncertainty, particularly in the feasibility and concept design stages of a maritime or offshore system. By contrast, a safety model using fuzzy logic approach employing fuzzy IF-THEN rules can model the qualitative aspects of human knowledge and reasoning processes without employing precise quantitative analyses. A fuzzy-logic-based approach may be more appropriately used to carry out risk analysis in the initial design stages. This provides a tool for working directly with the linguistic terms commonly used in carrying out safety assessment. This research focuses on the development and representation of linguistic variables to model risk levels subjectively. These variables are then quantified using fuzzy sets. In this paper, the development of a safety model using fuzzy logic approach for modelling various design variables for maritime and offshore safety based decision making in the concept design stage is presented. An example is used to illustrate the proposed approach

Quantitative research.

Science.gov (United States)

Watson, Roger

2015-04-01

This article describes the basic tenets of quantitative research. The concepts of dependent and independent variables are addressed and the concept of measurement and its associated issues, such as error, reliability and validity, are explored. Experiments and surveys – the principal research designs in quantitative research – are described and key features explained. The importance of the double-blind randomised controlled trial is emphasised, alongside the importance of longitudinal surveys, as opposed to cross-sectional surveys. Essential features of data storage are covered, with an emphasis on safe, anonymous storage. Finally, the article explores the analysis of quantitative data, considering what may be analysed and the main uses of statistics in analysis.

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Safety balance: Analysis of safety systems

International Nuclear Information System (INIS)

Delage, M.; Giroux, C.

1990-12-01

Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses

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

Science.gov (United States)

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

2010-07-31

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

Coupling of channel thermalhydraulics and fuel behaviour in ACR-1000 safety analyses

International Nuclear Information System (INIS)

Huang, F.L.; Lei, Q.M.; Zhu, W.; Bilanovic, Z.

2008-01-01

Channel thermalhydraulics and fuel thermal-mechanical behaviour are interlinked. This paper describes a channel thermalhydraulics and fuel behaviour coupling methodology that has been used in ACR-1000 safety analyses. The coupling is done for all 12 fuel bundles in a fuel channel using the channel thermalhydraulics code CATHENA MOD-3.5d/Rev 2 and the transient fuel behaviour code ELOCA 2.2. The coupling approach can be used for every fuel element or every group of fuel elements in the channel. Test cases are presented where a total of 108 fuel element models are set up to allow a full coupling between channel thermalhydraulics and detailed fuel analysis for a channel containing a string of 12 fuel bundles. An additional advantage of this coupling approach is that there is no need for a separate detailed fuel analysis because the coupling analysis, once done, provides detailed calculations for the fuel channel (fuel bundles, pressure tube, and calandria tube) as well as all the fuel elements (or element groups) in the channel. (author)

Quantitative analysis with energy dispersive X-ray fluorescence analyser

International Nuclear Information System (INIS)

Kataria, S.K.; Kapoor, S.S.; Lal, M.; Rao, B.V.N.

1977-01-01

Quantitative analysis of samples using radioisotope excited energy dispersive x-ray fluorescence system is described. The complete set-up is built around a locally made Si(Li) detector x-ray spectrometer with an energy resolution of 220 eV at 5.94 KeV. The photopeaks observed in the x-ray fluorescence spectra are fitted with a Gaussian function and the intensities of the characteristic x-ray lines are extracted, which in turn are used for calculating the elemental concentrations. The results for a few typical cases are presented. (author)

Analyses in support of risk-informed natural gas vehicle maintenance facility codes and standards :

Energy Technology Data Exchange (ETDEWEB)

Ekoto, Isaac W.; Blaylock, Myra L.; LaFleur, Angela Christine; LaChance, Jeffrey L.; Horne, Douglas B.

2014-03-01

Safety standards development for maintenance facilities of liquid and compressed gas fueled large-scale vehicles is required to ensure proper facility design and operation envelopes. Standard development organizations are utilizing risk-informed concepts to develop natural gas vehicle (NGV) codes and standards so that maintenance facilities meet acceptable risk levels. The present report summarizes Phase I work for existing NGV repair facility code requirements and highlights inconsistencies that need quantitative analysis into their effectiveness. A Hazardous and Operability study was performed to identify key scenarios of interest. Finally, scenario analyses were performed using detailed simulations and modeling to estimate the overpressure hazards from HAZOP defined scenarios. The results from Phase I will be used to identify significant risk contributors at NGV maintenance facilities, and are expected to form the basis for follow-on quantitative risk analysis work to address specific code requirements and identify effective accident prevention and mitigation strategies.

Traceability of Software Safety Requirements in Legacy Safety Critical Systems

Science.gov (United States)

Hill, Janice L.

2007-01-01

How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

Uncertainty in safety : new techniques for the assessment and optimisation of safety in process industry

NARCIS (Netherlands)

Rouvroye, J.L.; Nieuwenhuizen, J.K.; Brombacher, A.C.; Stavrianidis, P.; Spiker, R.Th.E.; Pyatt, D.W.

1995-01-01

At this moment there is no standardised method for the assessment for safety in the process industry. Many companies and institutes use qualitative techniques for safety analysis while other companies and institutes use quantitative techniques. The authors of this paper will compare different

Reliability analysis of software based safety functions

International Nuclear Information System (INIS)

Pulkkinen, U.

1993-05-01

The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

Job safety and awareness analysis of safety implementation among electrical workers in airport service company

Directory of Open Access Journals (Sweden)

Putra Perdana Suteja

2018-01-01

Full Text Available Electrical is a fundamental process in the company that has high risk and responsibility especially in public service company such as an airport. Hence, the company that operates activities in the airport has to identify and control the safety activities of workers. On the safety implementation, the lack of workers’ awareness is fundamental aspects to the safety failure. Therefore, this study aimed to analyse the safety awareness and identify risk in the electrical workplace. Safety awareness questionnaires are distributed to ten workers in order to analyse their awareness. Job safety analysis method used to identify the risk in the electrical workplace. The preliminary study stated that workers were not aware of personal protective equipment usage so that the awareness and behavioural need to be analysed. The result is the hazard was found such as electrical shock and noise for various intensity in the workplace. While electrical workers were aware of safety implementation but less of safety behaviour. Furthermore, the recommendation can be implemented are the implementation of behaviour-based safety (BBS, 5S implementation and accident report list.

Real Patient and its Virtual Twin: Application of Quantitative Systems Toxicology Modelling in the Cardiac Safety Assessment of Citalopram.

Science.gov (United States)

Patel, Nikunjkumar; Wiśniowska, Barbara; Jamei, Masoud; Polak, Sebastian

2017-11-27

A quantitative systems toxicology (QST) model for citalopram was established to simulate, in silico, a 'virtual twin' of a real patient to predict the occurrence of cardiotoxic events previously reported in patients under various clinical conditions. The QST model considers the effects of citalopram and its most notable electrophysiologically active primary (desmethylcitalopram) and secondary (didesmethylcitalopram) metabolites, on cardiac electrophysiology. The in vitro cardiac ion channel current inhibition data was coupled with the biophysically detailed model of human cardiac electrophysiology to investigate the impact of (i) the inhibition of multiple ion currents (I Kr , I Ks , I CaL ); (ii) the inclusion of metabolites in the QST model; and (iii) unbound or total plasma as the operating drug concentration, in predicting clinically observed QT prolongation. The inclusion of multiple ion channel current inhibition and metabolites in the simulation with unbound plasma citalopram concentration provided the lowest prediction error. The predictive performance of the model was verified with three additional therapeutic and supra-therapeutic drug exposure clinical cases. The results indicate that considering only the hERG ion channel inhibition of only the parent drug is potentially misleading, and the inclusion of active metabolite data and the influence of other ion channel currents should be considered to improve the prediction of potential cardiac toxicity. Mechanistic modelling can help bridge the gaps existing in the quantitative translation from preclinical cardiac safety assessment to clinical toxicology. Moreover, this study shows that the QST models, in combination with appropriate drug and systems parameters, can pave the way towards personalised safety assessment.

Analyse quantitative des effluents de pyrolyse en milieu ouvert et fermé Quantitative Analysis of Pyrolysis Effluents in an Open and Closed System

Directory of Open Access Journals (Sweden)

Behar F.

2006-11-01

Full Text Available Dans la première partie de l'article, nous décrivons une technique de pyrolyse en milieu ouvert qui permet de caractériser les matières organiques complexes comme le kérogène, le charbon, les asphaltènes de roche et d'huiles, les substances humiques et fulviques etc. Les effluents de pyrolyse sont récupérés et fractionnés quantitativement puis analysés par des techniques spécifiques comme la chromatographie en phase gazeuse et le couplage chromatographie/spectrométrie de masse. Dans la deuxième partie, est présentée une technique de pyrolyse en milieu fermé pour simuler au laboratoire l'évolution thermique des kérogènes, asphaltènes ou huiles. Nous nous sommes surtout attachés à dresser des bilans massiques et des bilans de l'hydrogène sur l'ensemble des produits de pyrolyse. Pour cela, nous avons distingué cinq classes de poids moléculaire croissant : C1, C2-C5, C6-C13, C14+ et coke. La récupération quantitative et la séparation de chacune des cinq fractions permet une analyse moléculaire détaillée de chacune d'elles. The first part of this article describes an open pyrolysis system in order to characterize complex organic matter such as kerogen, coal, rock and oil asphaltenes and humic substances, etc. Pyrolysis effluents are recovered, fractionated quantitatively by liquid chromatography, and then they are analyzed by specific techniques such as gas chromatography and chromatography/mass-spectrometry coupling. The second part describes a pyrolysis technique in a closed system, used for the laboratory simulation of the thermal evolution of kerogens, asphaltenes or oils. A special effort has been made to give the mass and hydrogen balances for all pyrolysis products. For this, five classes have been distinguised with increasing molecular weight: C1, C2-C5, C6-C13, C14+ and coke. The quantitative recovery and the separation of each of the five fractions is used to make a detailed molecular analysis of each of

Dry-air drying at room temperature - a practical pre-treatment method of tree leaves for quantitative analyses of phenolics?

Science.gov (United States)

Tegelberg, Riitta; Virjamo, Virpi; Julkunen-Tiitto, Riitta

2018-03-09

In ecological experiments, storage of plant material is often needed between harvesting and laboratory analyses when the number of samples is too large for immediate, fresh analyses. Thus, accuracy and comparability of the results call for pre-treatment methods where the chemical composition remains unaltered and large number of samples can be treated efficiently. To study if a fast dry-air drying provides an efficient pre-treatment method for quantitative analyses of phenolics. Dry-air drying of mature leaves was done in a drying room equipped with dehumifier (10% relative humidity, room temperature) and results were compared to freeze-drying or freeze-drying after pre-freezing in liquid nitrogen. The quantities of methanol-soluble phenolics of Betula pendula Roth, Betula pubescens Ehrh., Salix myrsinifolia Salisb., Picea abies L. Karsten and Pinus sylvestris L. were analysed with HPLC and condensed tannins were analysed using the acid-butanol test. In deciduous tree leaves (Betula, Salix), the yield of most of the phenolic compounds was equal or higher in samples dried in dry-air room than the yield from freeze-dried samples. In Picea abies needles, however, dry-air drying caused severe reductions in picein, stilbenes, condensed tannin and (+)-catechin concentrations compared to freeze-drying. In Pinus sylvestris highest yields of neolignans but lowest yields of acetylated flavonoids were obtained from samples freeze-dried after pre-freezing. Results show that dry-air drying provides effective pre-treatment method for quantifying the soluble phenolics for deciduous tree leaves, but when analysing coniferous species, the different responses between structural classes of phenolics should be taken into account. Copyright © 2018 John Wiley & Sons, Ltd.

Safety

International Nuclear Information System (INIS)

2001-01-01

This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

Contrasting safety assessments of a runway incursion scenario: Event sequence analysis versus multi-agent dynamic risk modelling

International Nuclear Information System (INIS)

Stroeve, Sybert H.; Blom, Henk A.P.; Bakker, G.J.

2013-01-01

In the safety literature it has been argued, that in a complex socio-technical system safety cannot be well analysed by event sequence based approaches, but requires to capture the complex interactions and performance variability of the socio-technical system. In order to evaluate the quantitative and practical consequences of these arguments, this study compares two approaches to assess accident risk of an example safety critical sociotechnical system. It contrasts an event sequence based assessment with a multi-agent dynamic risk model (MA-DRM) based assessment, both of which are performed for a particular runway incursion scenario. The event sequence analysis uses the well-known event tree modelling formalism and the MA-DRM based approach combines agent based modelling, hybrid Petri nets and rare event Monte Carlo simulation. The comparison addresses qualitative and quantitative differences in the methods, attained risk levels, and in the prime factors influencing the safety of the operation. The assessments show considerable differences in the accident risk implications of the performance of human operators and technical systems in the runway incursion scenario. In contrast with the event sequence based results, the MA-DRM based results show that the accident risk is not manifest from the performance of and relations between individual human operators and technical systems. Instead, the safety risk emerges from the totality of the performance and interactions in the agent based model of the safety critical operation considered, which coincides very well with the argumentation in the safety literature.

Lift truck safety review

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, L.C.

1997-03-01

This report presents safety information about powered industrial trucks. The basic lift truck, the counterbalanced sit down rider truck, is the primary focus of the report. Lift truck engineering is briefly described, then a hazard analysis is performed on the lift truck. Case histories and accident statistics are also given. Rules and regulations about lift trucks, such as the US Occupational Safety an Health Administration laws and the Underwriter`s Laboratories standards, are discussed. Safety issues with lift trucks are reviewed, and lift truck safety and reliability are discussed. Some quantitative reliability values are given.

Lift truck safety review

International Nuclear Information System (INIS)

Cadwallader, L.C.

1997-03-01

This report presents safety information about powered industrial trucks. The basic lift truck, the counterbalanced sit down rider truck, is the primary focus of the report. Lift truck engineering is briefly described, then a hazard analysis is performed on the lift truck. Case histories and accident statistics are also given. Rules and regulations about lift trucks, such as the US Occupational Safety an Health Administration laws and the Underwriter's Laboratories standards, are discussed. Safety issues with lift trucks are reviewed, and lift truck safety and reliability are discussed. Some quantitative reliability values are given

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

International Nuclear Information System (INIS)

Chung, Dae Wook

2011-01-01

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

Assessment of the factors with significant influence on safety culture

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.

2013-01-01

In this paper, a qualitative and a quantitative evaluation of the factors with significant impact on safety culture were performed. These techniques were established and applied in accordance with IAEA standards. In order to show the applicability and opportunity of the methodology a specific case study was prepared: safety culture evaluation for INR Pitesti. The qualitative evaluation was performed using specific developed questionnaires. Through analysis of the completed questionnaires was established the development stage of safety culture at INR. The quantitative evaluation was performed using a guide to rate the influence factors. For each factor was identified the influence (negative or positive) and ranking score was estimated using scoring criteria. The results have emphasized safety culture stages. The paper demonstrates the fact that using both quantitative and qualitative assessment techniques, a practical value of the safety culture concept is given. (authors)

A Quantitative Analysis of Nursing Students' Perceptions of Patient Safety Competencies

Science.gov (United States)

Steighner, Tammy Rose

2017-01-01

The purpose of the study was to determine nursing students' perceptions of patient safety competencies as it related to Quality and Safety Education for Nurses (QSEN) competencies and the Safety Competencies Framework developed by The Canadian Patient Safety Institute. The study determined if nursing students knew how to provide safe patient care…

Ferrocyanide safety project: Task 3.5 cyanide species analytical methods development

International Nuclear Information System (INIS)

Bryan, S.A.; Pool, K.H.; Burger, L.L.; Carlson, C.D.; Hess, N.J.; Matheson, J.D.; Ryan, J.L.; Scheele, R.D.; Tingey, J.M.

1993-01-01

This report summarizes the results of studies conducted in FY 1992 to develop methods for the identification and quantification of cyanide species in ferrocyanide tank waste. Currently there are 24 high-level waste storage tanks at the Hanford Site that have been placed on a Ferrocyanide Tank Watchlist because they contain an estimated 1,000 g-moles or greater amount of precipitated ferrocyanide. This amount of ferrocyanide is of concern because the consequences of a potential explosion may exceed those reported previously in safety analyses. The threshold concentration of total cyanide within the tank waste matrix that is expected to be a safety concern is estimated at approximately 1 to 3 wt%. Methods for detection and speciation of ferrocyanide complexes in actual waste are needed to definitively measure and quantitate the amount of ferrocyanides present within actual waste tanks to a lower limit of at least 0.1 wt% in order to bound the safety concern

Grundlegende quantitative Analysen medizinischer Prüfungen [Basic quantitative analyses of medical examinations

Directory of Open Access Journals (Sweden)

Möltner, Andreas

2006-08-01

Full Text Available [english] The evaluation steps are described which are necessary for an elementary test-theoretic analysis of an exam and sufficient as a basis of item-revisions, improvements of the composition of tests and feedback to teaching coordinators and curriculum developers. These steps include the evaluation of the results, the analysis of item difficulty and discrimination and - where appropriate - the corresponding evaluation of single answers. To complete the procedure, the internal consistency is determined, which makes an estimate of the reliability and significance of the examination. [german] Es werden die Auswertungsschritte beschrieben, die für eine einfache testtheoretische Analyse einer Prüfung notwendig sowie als Grundlage von Aufgabenrevisionen, Verbesserungen von Prüfungszusammenstellungen und Rückmeldung an Lehrbeauftragte und Curriculumsentwickler ausreichend sind. Diese Schritte umfassen die Ergebnisauswertung, die Analyse der Aufgabenschwierigkeiten und der Trennschärfen, sowie - wo angebracht - die entsprechenden Auswertungen der Einzelantworten. Vervollständigt wird das Vorgehen durch die Bestimmung der internen Konsistenz, durch die die Zuverlässigkeit und Aussagekraft (Reliabilität der Prüfung abgeschätzt wird.

Quantitative risk trends deriving from PSA-based event analyses. Analysis of results from U.S.NRC's accident sequence precursor program

International Nuclear Information System (INIS)

Watanabe, Norio

2004-01-01

The United States Nuclear Regulatory Commission (U.S.NRC) has been carrying out the Accident Sequence Precursor (ASP) Program to identify and categorize precursors to potential severe core damage accident sequences using the probabilistic safety assessment (PSA) technique. The ASP Program has identified a lot of risk significant events as precursors that occurred at U.S. nuclear power plants. Although the results from the ASP Program include valuable information that could be useful for obtaining and characterizing risk significant insights and for monitoring risk trends in nuclear power industry, there are only a few attempts to determine and develop the trends using the ASP results. The present study examines and discusses quantitative risk trends for the industry level, using two indicators, that is, the occurrence frequency of precursors and the annual core damage probability, deriving from the results of the ASP analysis. It is shown that the core damage risk at U.S. nuclear power plants has been lowered and the likelihood of risk significant events has been remarkably decreasing. As well, the present study demonstrates that two risk indicators used here can provide quantitative information useful for examining and monitoring the risk trends and/or risk characteristics in nuclear power industry. (author)

Qualitative and Quantitative Security Analyses for ZigBee Wireless Sensor Networks

DEFF Research Database (Denmark)

Yuksel, Ender

methods and techniques in different areas and brings them together to create an efficient verification system. The overall ambition is to provide a wide range of powerful techniques for analyzing models with quantitative and qualitative security information. We stated a new approach that first verifies...... applications, home automation, and traffic control. The challenges for research in this area are due to the unique features of wireless sensor devices such as low processing power and associated low energy. On top of this, wireless sensor networks need secure communication as they operate in open fields...... low level security protocol s in a qualitative manner and guarantees absolute security, and then takes these verified protocols as actions of scenarios to be verified in a quantitative manner. Working on the emerging ZigBee wireless sensor networks, we used probabilistic verification that can return...

Inherent safety characteristics of innovative reactors

International Nuclear Information System (INIS)

Heil, J.A.

1995-11-01

The added safety value of innovative or third generation reactor designs has been evaluated in order to determine the most suitable candidate for Dutch government funded research and development support. To this end, four innovative reactor concepts, viz. PIUS (Process Inherent Ultimate Safety), PRISM (Power Reactor Innovative Small), HTR-M (High Temperature Reactor Module) and MHTGR (Modular High Temperature Gas-cooled Reactor), have been studied and their passive and inherent safety characteristics have been outlined. Also the outlook for further technological and industrial development has been considered. The results of the study confirm the perspective of the innovative reactors for reduced dependence on active safety provisions and for a further reduced vulnerability to technical failures and human errors. The accident responses to generic accident initiators, viz. reactivity and cooling accidents, and also to reactor specific accidents show that neither active safety systems nor short term operator actions are required for maintaining the reactor core in a controlled and coolable condition. Whether this gives rise to a higher total safety of the innovative reactor designs, compared to evolutionary or advanced reactors, cannot be concluded. Supplementary experimental and analytical analyses of reactor specific accidents are required to be able to assess the safety of these innovative designs in a more quantitative manner. It is believed that the safety case of innovative reactors, which are less dependent on active safety systems, can be communicated with the general public in a more transparent way. Considering the perspective for further technological and industrial development it is not expected that any of the considered innovative reactor concepts will become commercially available within the next one to two decades. However, they could be made available earlier if they would receive sufficient financial backing. Considering the added safety perspectives

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

International Nuclear Information System (INIS)

2014-12-01

In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-12-15

In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the

International validation of safety analyses for nuclear power plants; Mednarodno preverjanje varnostnih analiz za jedrske elektrane

Energy Technology Data Exchange (ETDEWEB)

Gregoric, N; Mavko, B [Institut ' Jozef Stefan' Ljubljana (Yugoslavia)

1988-07-01

Paper describes the participation of 'J.Stefan' Institute in international standard problems for validation of modeling and programs for safety analysis. Listed are main international experimental facilities for collecting data basic for understanding of physical phenomena, code development and validation of modelling and programs. Since the results of international standard problem analyses are published in a joint final report, it is simple to asses the conformance of the results of a particular group with the experiment. Good results from three international exercises done so far, have encouraged the group to currently participate in OECD-ISP-22 which is a model of the Italian three loop PWR. (author)

Development of safety analysis technology for LMR

International Nuclear Information System (INIS)

Lee, Y. B.; Kwon, Y. M.; Suk, S. D.

2005-03-01

The MATRA-LMR-FB has been developed internally for the damage prevention as well as the safety assessment during a channel blockage accident and, as a the result, the quality of the code becomes comparable to that developed in the leading countries. For a code-to-code comparison, KAERI could have access to the SASSYS-1 through a bilateral collaboration between KAERI and ANL. The study could bring into the reliability improvements both on the reactivity models in the SSC-K and on the SSC-K prediction capability. It finally leads to the completion of the SSC-K version 1.3 resulting from the qualitative and quantitative code-to-code comparison. The preliminary analysis for a metal fueled LMR could also become possible with the MELT-III and the VENUS-II, which had originally been developed for the HCDA analysis with an oxidized fuel, by developing the relevant models For the development of the safety evaluation technology, the safety limits have been set up, and the analyses of the internal and external channel blockages in an assembly have also been performed. Besides, the more reliable analysis results on the key design concepts could be obtained by way of the methodology improvement resulting from the qualitative and quantitative comparison study. For an efficient and systematic control of the main project, the integration of the developed technologies and the establishment of their data base have been pursued. It has gone through the development of the process control with taking account of interfaces among the sub-projects, the overall coordination of the developed technologies, the data base for the design products, and so on

Safety-barrier diagrams as a tool for modelling safety of hydrogen applications

DEFF Research Database (Denmark)

Duijm, Nijs Jan; Markert, Frank

2009-01-01

Safety-barrier diagrams have proven to be a useful tool in documenting the safety measures taken to prevent incidents and accidents in process industry. Especially during the introduction of new hydrogen technologies or applications, as e.g. hydrogen refuelling stations, safety-barrier diagrams...... are considered a valuable supplement to other traditional risk analysis tools to support the communication with authorities and other stakeholders during the permitting process. Another advantage of safety-barrier diagrams is that they highlight the importance of functional and reliable safety barriers in any...... system and here is a direct focus on those barriers that need to be subject to safety management in terms of design and installation, operational use, inspection and monitoring, and maintenance. Safety-barrier diagrams support both quantitative and qualitative approaches. The paper will describe...

Incorporating assumption deviation risk in quantitative risk assessments: A semi-quantitative approach

International Nuclear Information System (INIS)

Khorsandi, Jahon; Aven, Terje

2017-01-01

Quantitative risk assessments (QRAs) of complex engineering systems are based on numerous assumptions and expert judgments, as there is limited information available for supporting the analysis. In addition to sensitivity analyses, the concept of assumption deviation risk has been suggested as a means for explicitly considering the risk related to inaccuracies and deviations in the assumptions, which can significantly impact the results of the QRAs. However, challenges remain for its practical implementation, considering the number of assumptions and magnitude of deviations to be considered. This paper presents an approach for integrating an assumption deviation risk analysis as part of QRAs. The approach begins with identifying the safety objectives for which the QRA aims to support, and then identifies critical assumptions with respect to ensuring the objectives are met. Key issues addressed include the deviations required to violate the safety objectives, the uncertainties related to the occurrence of such events, and the strength of knowledge supporting the assessments. Three levels of assumptions are considered, which include assumptions related to the system's structural and operational characteristics, the effectiveness of the established barriers, as well as the consequence analysis process. The approach is illustrated for the case of an offshore installation. - Highlights: • An approach for assessing the risk of deviations in QRA assumptions is presented. • Critical deviations and uncertainties related to their occurrence are addressed. • The analysis promotes critical thinking about the foundation and results of QRAs. • The approach is illustrated for the case of an offshore installation.

A prioritization of generic safety issues

International Nuclear Information System (INIS)

Emrit, R.; Riggs, R.; Milstead, W.; Pittman, J.

1991-07-01

This report presents the priority rankings for generic safety issues and related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The report focuses on the prioritization of generic safety issues. Issues primarily concerned with the licensing process or environmental protection and not directly related to safety have been excluded from prioritization. The prioritized issues include: TMI Action Plan items under development; previously proposed issues covered by Task Action Plans, except issues designated as Un-resolved Safety Issues (USIs) which had already been assigned high priority; and newly-proposed issues. Future supplements to this report will include the prioritization of additional issues. The safety priority rankings are High, Medium, Low, and Drop and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolutions of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative. 1310 refs

Similarity analyses of chromatographic herbal fingerprints: A review

International Nuclear Information System (INIS)

Goodarzi, Mohammad; Russell, Paul J.; Vander Heyden, Yvan

2013-01-01

Graphical abstract: -- Highlights: •Similarity analyses of herbal fingerprints are reviewed. •Different (dis)similarity approaches are discussed. •(Dis)similarity-metrics and exploratory-analysis approaches are illustrated. •Correlation and distance-based measures are overviewed. •Similarity analyses illustrated by several case studies. -- Abstract: Herbal medicines are becoming again more popular in the developed countries because being “natural” and people thus often assume that they are inherently safe. Herbs have also been used worldwide for many centuries in the traditional medicines. The concern of their safety and efficacy has grown since increasing western interest. Herbal materials and their extracts are very complex, often including hundreds of compounds. A thorough understanding of their chemical composition is essential for conducting a safety risk assessment. However, herbal material can show considerable variability. The chemical constituents and their amounts in a herb can be different, due to growing conditions, such as climate and soil, the drying process, the harvest season, etc. Among the analytical methods, chromatographic fingerprinting has been recommended as a potential and reliable methodology for the identification and quality control of herbal medicines. Identification is needed to avoid fraud and adulteration. Currently, analyzing chromatographic herbal fingerprint data sets has become one of the most applied tools in quality assessment of herbal materials. Mostly, the entire chromatographic profiles are used to identify or to evaluate the quality of the herbs investigated. Occasionally only a limited number of compounds are considered. One approach to the safety risk assessment is to determine whether the herbal material is substantially equivalent to that which is either readily consumed in the diet, has a history of application or has earlier been commercialized i.e. to what is considered as reference material. In order

Similarity analyses of chromatographic herbal fingerprints: A review

Energy Technology Data Exchange (ETDEWEB)

Goodarzi, Mohammad [Department of Analytical Chemistry and Pharmaceutical Technology, Center for Pharmaceutical Research, Vrije Universiteit Brussel, Laarbeeklaan 103, B-1090 Brussels (Belgium); Russell, Paul J. [Safety and Environmental Assurance Centre, Unilever, Colworth Science Park, Sharnbrook, Bedfordshire MK44 1LQ (United Kingdom); Vander Heyden, Yvan, E-mail: yvanvdh@vub.ac.be [Department of Analytical Chemistry and Pharmaceutical Technology, Center for Pharmaceutical Research, Vrije Universiteit Brussel, Laarbeeklaan 103, B-1090 Brussels (Belgium)

2013-12-04

Graphical abstract: -- Highlights: •Similarity analyses of herbal fingerprints are reviewed. •Different (dis)similarity approaches are discussed. •(Dis)similarity-metrics and exploratory-analysis approaches are illustrated. •Correlation and distance-based measures are overviewed. •Similarity analyses illustrated by several case studies. -- Abstract: Herbal medicines are becoming again more popular in the developed countries because being “natural” and people thus often assume that they are inherently safe. Herbs have also been used worldwide for many centuries in the traditional medicines. The concern of their safety and efficacy has grown since increasing western interest. Herbal materials and their extracts are very complex, often including hundreds of compounds. A thorough understanding of their chemical composition is essential for conducting a safety risk assessment. However, herbal material can show considerable variability. The chemical constituents and their amounts in a herb can be different, due to growing conditions, such as climate and soil, the drying process, the harvest season, etc. Among the analytical methods, chromatographic fingerprinting has been recommended as a potential and reliable methodology for the identification and quality control of herbal medicines. Identification is needed to avoid fraud and adulteration. Currently, analyzing chromatographic herbal fingerprint data sets has become one of the most applied tools in quality assessment of herbal materials. Mostly, the entire chromatographic profiles are used to identify or to evaluate the quality of the herbs investigated. Occasionally only a limited number of compounds are considered. One approach to the safety risk assessment is to determine whether the herbal material is substantially equivalent to that which is either readily consumed in the diet, has a history of application or has earlier been commercialized i.e. to what is considered as reference material. In order

Quantitative and qualitative analyses of subacromial impingement by kinematic open MRI.

Science.gov (United States)

Tasaki, Atsushi; Nimura, Akimoto; Nozaki, Taiki; Yamakawa, Akira; Niitsu, Mamoru; Morita, Wataru; Hoshikawa, Yoshimitsu; Akita, Keiichi

2015-05-01

Quantitative and qualitative kinematic analyses of subacromial impingement by 1.2T open MRI were performed to determine the location of impingement and the involvement of the acromioclavicular joint. In 20 healthy shoulders, 10 sequential images in the scapular plane were taken in a 10-s pause at equal intervals from 30° to maximum abduction in neutral and internal rotation. The distances between the rotator cuff (RC) and the acromion and the acromioclavicular joint were measured. To comprehend the positional relationships, cadaveric specimens were also observed. Although asymptomatic, the RC came into contact with the acromion and the acromioclavicular joint in six and five cases, respectively. The superior RC acted as a depressor for the humeral head against the acromion as the shoulder elevated. The mean elevation angle and distance at the closest position between the RC and the acromion in neutral rotation were 93.5° and 1.6 mm, respectively, while those between the RC and the acromioclavicular joint were 86.7° and 2.0 mm. When comparing this distance and angle, there was no significant difference between the RC to the acromion and to the acromioclavicular joint. The minimum distance between the RC and the acromion was significantly shorter than that between the greater tuberosity and the acromion. The location of RC closest to the acromion and the acromioclavicular joint differed significantly. Although asymptomatic, contact was found between the RC and the acromion and the acromioclavicular joint. The important role of the RC to prevent impingement was observed, and hence, dysfunction of the RC could lead to impingement that could result in a RC lesion. The RC lesions may differ when they are caused by impingement from either the acromion or the acromioclavicular joint.

Quantitative characterization of colloidal assembly of graphene oxide-silver nanoparticle hybrids using aerosol differential mobility-coupled mass analyses.

Science.gov (United States)

Nguyen, Thai Phuong; Chang, Wei-Chang; Lai, Yen-Chih; Hsiao, Ta-Chih; Tsai, De-Hao

2017-10-01

In this work, we develop an aerosol-based, time-resolved ion mobility-coupled mass characterization method to investigate colloidal assembly of graphene oxide (GO)-silver nanoparticle (AgNP) hybrid nanostructure on a quantitative basis. Transmission electron microscopy (TEM) and zeta potential (ZP) analysis were used to provide visual information and elemental-based particle size distributions, respectively. Results clearly show a successful controlled assembly of GO-AgNP by electrostatic-directed heterogeneous aggregation between GO and bovine serum albumin (BSA)-functionalized AgNP under an acidic environment. Additionally, physical size, mass, and conformation (i.e., number of AgNP per nanohybrid) of GO-AgNP were shown to be proportional to the number concentration ratio of AgNP to GO (R) and the selected electrical mobility diameter. An analysis of colloidal stability of GO-AgNP indicates that the stability increased with its absolute ZP, which was dependent on R and environmental pH. The work presented here provides a proof of concept for systematically synthesizing hybrid colloidal nanomaterials through the tuning of surface chemistry in aqueous phase with the ability in quantitative characterization. Graphical Abstract Colloidal assembly of graphene oxide-silver nanoparticle hybrids characterized by aerosol differential mobility-coupled mass analyses.

NASA System Safety Handbook. Volume 1; System Safety Framework and Concepts for Implementation

Science.gov (United States)

Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Smith, Curtis; Stamatelatos, Michael; Youngblood, Robert

2011-01-01

System safety assessment is defined in NPR 8715.3C, NASA General Safety Program Requirements as a disciplined, systematic approach to the analysis of risks resulting from hazards that can affect humans, the environment, and mission assets. Achievement of the highest practicable degree of system safety is one of NASA's highest priorities. Traditionally, system safety assessment at NASA and elsewhere has focused on the application of a set of safety analysis tools to identify safety risks and formulate effective controls.1 Familiar tools used for this purpose include various forms of hazard analyses, failure modes and effects analyses, and probabilistic safety assessment (commonly also referred to as probabilistic risk assessment (PRA)). In the past, it has been assumed that to show that a system is safe, it is sufficient to provide assurance that the process for identifying the hazards has been as comprehensive as possible and that each identified hazard has one or more associated controls. The NASA Aerospace Safety Advisory Panel (ASAP) has made several statements in its annual reports supporting a more holistic approach. In 2006, it recommended that "... a comprehensive risk assessment, communication and acceptance process be implemented to ensure that overall launch risk is considered in an integrated and consistent manner." In 2009, it advocated for "... a process for using a risk-informed design approach to produce a design that is optimally and sufficiently safe." As a rationale for the latter advocacy, it stated that "... the ASAP applauds switching to a performance-based approach because it emphasizes early risk identification to guide designs, thus enabling creative design approaches that might be more efficient, safer, or both." For purposes of this preface, it is worth mentioning three areas where the handbook emphasizes a more holistic type of thinking. First, the handbook takes the position that it is important to not just focus on risk on an individual

Study on safety subsidiary objective of nuclear power plant in USA

International Nuclear Information System (INIS)

Chen Yan; Zhang Chunming; Fu Zhiwei; Song Wei; Li Chaojun; Wang Zhe; Zuo Jiaxu

2013-01-01

This paper reviewed the development of the quantitative safety objective and subsidiary objective in USA. The expressions of CDF and LERF were obtained according to NUREG-1150. The relationship between the subsidiary objective and the quantitative safety objective was derived. The method was compared with that used in NUREG-1860. The requirements of safety objective for the future nuclear power plant and the development of probabilistic safety analysis (PSA) technology in USA were studied and can be used as reference in China. (authors)

Requirements of safety and reliability

International Nuclear Information System (INIS)

Franzen, L.F.

1977-01-01

The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the findings derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essential for accident analyses, and the determination of the loads occuring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig./HP) [de

A method for quantitative measurement of safety culture based on ISO 26262

NARCIS (Netherlands)

Khabbaz Saberi, A.; Benders, F.; Koch, R.; Lukkien, J.J.; van den Brand, M.G.J.; Parsons, M.; Kelly, T.

2018-01-01

Safety culture is the collective attitude of members of an organization regarding safety issues: awareness, communication, knowledge, etc. In the automotive industry, specifically in its Research and Development (R&D) environments, safety culture is relatively new. Recent incidents related to

Contamination, decontamination and radiochemical safety analyses of the RA reactor (Report 1966)

International Nuclear Information System (INIS)

Maksimovic, Z.

1966-12-01

This contract is concerned with development of methods for detection of fission products i the heavy water and quantitative radiochemical analysis for detecting one fission product which enables reliable verification of heavy water contamination by fission products and estimation of contamination level. Qualitative and quantitative radiometry measurements of fission products in water are shown on page 4. Page 6 shows study of contamination and decontamination of water on the laboratory level. Experiments have shown that the majority of fission products was adsorbed on the uranium oxide and that the iodine isotopes are partly in water (non-adsorbed). Gamma spectrometry analyses showed 131 I moves to distillate with the initial quantities of distilled water. decontamination factors compared to the total activity of fission products in distillator and distillate are not higher than ∼10 3 . Decontamination of water contaminated by uranium oxide and fission products in the distillation device of the RA reactor is shown on page 8. Experiments demanded special preparation due to high activity of uranium (1.7 g of uranium irradiated in the reactor for 10 days at neutron flux 1.10 13 n.cm 2 /s. Prior preparations for transport and dissolution of irradiated metal uranium as well as sampling were needed. Distillation was done under lower pressure and temperature to avoid possible contamination of the environment bu fission products and iodine. Decontamination factors are shown in Table. Contamination and decontamination of stainless steel on the laboratory level are described on page 5. It was found that the deposition of activity on the stainless steel plates is inhomogeneous showing that the uranium oxide and fission products are deposited on the rough metal surfaces. According to literature data and our laboratory studies decontamination was done by nitric acid solution (2MHNO 3 ). Since the heavy water system of the RA reactor was made of stainless teel (except the

Updated safety analysis of ITER

Energy Technology Data Exchange (ETDEWEB)

Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

2011-10-15

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Updated safety analysis of ITER

International Nuclear Information System (INIS)

Taylor, Neill; Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid

2011-01-01

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

ITER safety

International Nuclear Information System (INIS)

Raeder, J.; Piet, S.; Buende, R.

1991-01-01

As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

[Patient safety and errors in medicine: development, prevention and analyses of incidents].

Science.gov (United States)

Rall, M; Manser, T; Guggenberger, H; Gaba, D M; Unertl, K

2001-06-01

"Patient safety" and "errors in medicine" are issues gaining more and more prominence in the eyes of the public. According to newer studies, errors in medicine are among the ten major causes of death in association with the whole area of health care. A new era has begun incorporating attention to a "systems" approach to deal with errors and their causes in the health system. In other high-risk domains with a high demand for safety (such as the nuclear power industry and aviation) many strategies to enhance safety have been established. It is time to study these strategies, to adapt them if necessary and apply them to the field of medicine. These strategies include: to teach people how errors evolve in complex working domains and how types of errors are classified; the introduction of critical incident reporting systems that are free of negative consequences for the reporters; the promotion of continuous medical education; and the development of generic problem-solving skills incorporating the extensive use of realistic simulators wherever possible. Interestingly, the field of anesthesiology--within which realistic simulators were developed--is referred to as a model for the new patient safety movement. Despite this proud track record in recent times though, there is still much to be done even in the field of anesthesiology. Overall though, the most important strategy towards a long-term improvement in patient safety will be a change of "culture" throughout the entire health care system. The "culture of blame" focused on individuals should be replaced by a "safety culture", that sees errors and critical incidents as a problem of the whole organization. The acceptance of human fallability and an open-minded non-punitive analysis of errors in the sense of a "preventive and proactive safety culture" should lead to solutions at the systemic level. This change in culture can only be achieved with a strong commitment from the highest levels of an organization. Patient

LWR safety studies. Analyses and further assessments relating to the German Risk Assessment Study on Nuclear Power Plants. Vol. 3

International Nuclear Information System (INIS)

1983-01-01

Critical review of the analyses of the German Risk Assessment Study on Nuclear Power Plants (DRS) concerning the reliability of the containment under accident conditions and the conditions of fission product release (transport and distribution in the environment). Main point of interest in this context is an explosion in the steam section and its impact on the containment. Critical comments are given on the models used in the DRS for determining the accident consequences. The analyses made deal with the mathematical models and database for propagation calculations, the methods of dose computation and assessment of health hazards, and the modelling of protective and safety measures. Social impacts of reactor accidents are also considered. (RF) [de

Safety study application guide

International Nuclear Information System (INIS)

1993-07-01

Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

Methods and strategies for future reactor safety goals

Science.gov (United States)

Arndt, Steven Andrew

-informed analyses and discussions. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of a comparable non-nuclear electric generation facility, as well as the risks associated with mining and transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric, as opposed to using single plant safety goals. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals, is explored. Finally, the role risk perception should play in establishing safety goals has been explored. To complete this evaluation, a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception, and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals' usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants than

Unique differences in applying safety analyses for a graphite moderated, channel reactor

International Nuclear Information System (INIS)

Moffitt, R.L.

1993-06-01

Unlike its predecessors, the N Reactor at the Hanford Site in Washington State was designed to produce electricity for civilian energy use as well as weapons-grade plutonium. This paper describes the major problems associated with applying safety analysis methodologies developed for commercial light water reactors (LWR) to a unique reactor like the N Reactor. The focus of the discussion is on non-applicable LWR safety standards and computer modeling/analytical variances of standards. The approaches used to resolve these problems to develop safety standards and limits for the N Reactor are described

RECOMMENDED TRITIUM OXIDE DEPOSITION VELOCITY FOR USE IN SAVANNAH RIVER SITE SAFETY ANALYSES

Energy Technology Data Exchange (ETDEWEB)

Lee, P.; Murphy, C.; Viner, B.; Hunter, C.; Jannik, T.

2012-04-03

The Defense Nuclear Facilities Safety Board (DNFSB) has recently questioned the appropriate value for tritium deposition velocity used in the MELCOR Accident Consequence Code System Ver. 2 (Chanin and Young 1998) code when estimating bounding dose (95th percentile) for safety analysis (DNFSB 2011). The purpose of this paper is to provide appropriate, defensible values of the tritium deposition velocity for use in Savannah River Site (SRS) safety analyses. To accomplish this, consideration must be given to the re-emission of tritium after deposition. Approximately 85% of the surface area of the SRS is forested. The majority of the forests are pine plantations, 68%. The remaining forest area is 6% mixed pine and hardwood and 26% swamp hardwood. Most of the path from potential release points to the site boundary is through forested land. A search of published studies indicate daylight, tritiated water (HTO) vapor deposition velocities in forest vegetation can range from 0.07 to 2.8 cm/s. Analysis of the results of studies done on an SRS pine plantation and climatological data from the SRS meteorological network indicate that the average deposition velocity during daylight periods is around 0.42 cm/s. The minimum deposition velocity was determined to be about 0.1 cm/s, which is the recommended bounding value. Deposition velocity and residence time (half-life) of HTO in vegetation are related by the leaf area and leaf water volume in the forest. For the characteristics of the pine plantation at SRS the residence time corresponding to the average, daylight deposition velocity is 0.4 hours. The residence time corresponding to the night-time deposition velocity of 0.1 cm/s is around 2 hours. A simple dispersion model which accounts for deposition and re-emission of HTO vapor was used to evaluate the impact on exposure to the maximally exposed offsite individual (MOI) at the SRS boundary (Viner 2012). Under conditions that produce the bounding, 95th percentile MOI exposure

Safety climate in OHSAS 18001-certified organisations: antecedents and consequences of safety behaviour.

Science.gov (United States)

Fernández-Muñiz, Beatriz; Montes-Peón, José Manuel; Vázquez-Ordás, Camilo José

2012-03-01

The occupational health and safety standard OHSAS 18001 has gained considerable acceptance worldwide, and firms from diverse sectors and of varying sizes have implemented it. Despite this, very few studies have analysed safety management or the safety climate in OHSAS 18001-certified organisations. The current work aims to analyse the safety climate in these organisations, identify its dimensions, and propose and test a structural equation model that will help determine the antecedents and consequences of employees' safety behaviour. For this purpose, the authors carry out an empirical study using a sample of 131 OHSAS 18001-certified organisations located in Spain. The results show that management's commitment, and particularly communication, have an effect on safety behaviour and on safety performance, employee satisfaction, and firm competitiveness. These findings are particularly important for management since they provide evidence about the factors that should be encouraged to reduce risks and improve performance in this type of organisation. Copyright © 2011 Elsevier Ltd. All rights reserved.

Analysis of Paks NPP Personnel Activity during Safety Related Event Sequences

International Nuclear Information System (INIS)

Bareith, A.; Hollo, Elod; Karsa, Z.; Nagy, S.

1998-01-01

Within the AGNES Project (Advanced Generic and New Evaluation of Safety) the Level-1 PSA model of the Paks NPP Unit 3 was developed in form of a detailed event tree/fault tree structure (53 initiating events, 580 event sequences, 6300 basic events are involved). This model gives a good basis for quantitative evaluation of potential consequences of actually occurred safety-related events, i.e. for precursor event studies. To make these studies possible and efficient, the current qualitative event analysis practice should be reviewed and a new additional quantitative analysis procedure and system should be developed and applied. The present paper gives an overview of the method outlined for both qualitative and quantitative analyses of the operator crew activity during off-normal situations. First, the operator performance experienced during past operational events is discussed. Sources of raw information, the qualitative evaluation process, the follow-up actions, as well as the documentation requirements are described. Second, the general concept of the proposed precursor event analysis is described. Types of modeled interactions and the considered performance influences are presented. The quantification of the potential consequences of the identified precursor events is based on the task-oriented, Level-1 PSA model of the plant unit. A precursor analysis system covering the evaluation of operator activities is now under development. Preliminary results gained during a case study evaluation of a past historical event are presented. (authors)

A high seroprevalence of antibodies to pertussis toxin among Japanese adults: Qualitative and quantitative analyses.

Directory of Open Access Journals (Sweden)

Takumi Moriuchi

Full Text Available In 2013, national serosurveillance detected a high seroprevalence of antibodies to pertussis toxin (PT from Bordetella pertussis among Japanese adults. Thus, we aimed to determine the cause(s of this high seroprevalence, and analyzed the titers of antibodies to PT and filamentous hemagglutinin (FHA among adults (35-44 years old, young children (4-7 years old, and older children (10-14 years old. Our quantitative analyses revealed that adults had higher seroprevalences of anti-PT IgG and PT-neutralizing antibodies, and similar titers of anti-FHA IgG, compared to the young and older children. Positive correlations were observed between the titers of PT-neutralizing antibodies and anti-PT IgG in all age groups (rs values of 0.326-0.522, although the correlation tended to decrease with age. The ratio of PT-neutralizing antibodies to anti-PT IgG was significantly different when we compared the serum and purified IgG fractions among adults (p = 0.016, although this result was not observed among young and older children. Thus, it appears that some adults had non-IgG immunoglobulins to PT. Our analyses also revealed that adults had high-avidity anti-PT IgG (avidity index: 63.5%, similar results were observed among the children; however, the adults had lower-avidity anti-FHA IgG (37.9%, p < 0.05. It is possible that low-avidity anti-FHA IgG is related to infection with other respiratory pathogens (e.g., Bordetella parapertussis, Haemophilus influenzae, or Mycoplasma pneumoniae, which produces antibodies to FHA-like proteins. Our observations suggest that these adults had been infected with B. pertussis and other pathogen(s during their adulthood.

Safety KPIs - Monitoring of safety performance

Directory of Open Access Journals (Sweden)

Andrej Lališ

2014-09-01

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

Methods and Effects of Safety Enhancement in Korean PSR

International Nuclear Information System (INIS)

Kim, Young Gab; Park, Jong Woon

2009-01-01

Periodic Safety Review (PSR) is a comprehensive study on a nuclear power plant safety, taking into account aspects such as operational history, ageing, safety analyses and advances in code and standards since the time of construction. In Korea, PSRs have been performed for 20 units and have been effectively used to obtain an overall view of actual plant safety to determine reasonable and practical modifications that should be made in order to obtain a higher level of safety approaching that of modern plants. Among many safety enhancements achieved from Korean PSRs, new safety analyses are the important methods to confirm plant safety by increasing safety margin for specific safety issues. Methods and effects of safety enhancements applied in Korean PSRs are reviewed in this paper in light of new safety analyses to obtain additional safety margins

Safety; Avertissement

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS

Energy Technology Data Exchange (ETDEWEB)

GARVIN, L. J.; JENSEN, M. A.

2004-04-13

This document summarizes safety management programs used within the scope of the ''Project Hanford Management Contract''. The document has been developed to meet the format and content requirements of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses''. This document provides summary descriptions of Fluor Hanford safety management programs, which Fluor Hanford nuclear facilities may reference and incorporate into their safety basis when producing facility- or activity-specific documented safety analyses (DSA). Facility- or activity-specific DSAs will identify any variances to the safety management programs described in this document and any specific attributes of these safety management programs that are important for controlling potentially hazardous conditions. In addition, facility- or activity-specific DSAs may identify unique additions to the safety management programs that are needed to control potentially hazardous conditions.

Establishment of safety goal and its quantification based on risk assessment

International Nuclear Information System (INIS)

Miyano, Hiroshi; Muramatsu, Ken

2017-01-01

We must clarify the safety objectives sought by society in securing the safety of nuclear reactors and nuclear power plants. For that purpose, it is useful to utilize risk assessment. Quantitative methods including probabilistic risk assessment (PRA) are superior in terms of scientific rationality and quantitative performance compared with conventional deterministic methods, and able to indicate an objective numerical value of safety level. Consequently, quantitative methods can enhance the transparency, consistency, compliance, predictability, and explanatory power of regulatory decisions toward business operators and citizens. Business operators can explain the validity of their own safety assurance activities to regulators and citizens. The goal to be secured becomes clear by incorporating the safety goal into the specific performance goal required for the nuclear power plant from the viewpoint of deep safeguard, and it becomes easy to evaluate the effectiveness of the safety measures. It helps us greatly in judging and selecting the appropriateness of safety measures. It should be noted: the fact that the result of implementing the PRA satisfies the safety goal is not a sufficient condition in the sense of guaranteeing complete safety but a necessary condition. The nuclear power field is a region with large uncertainty, and research/efforts for accuracy improvement and evaluation validity will be required continuously. (A.O.)

A prioritization of generic safety issues. Supplement 21, Revision insertion instructions

Energy Technology Data Exchange (ETDEWEB)

None, None

1996-12-31

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative.

A prioritization of generic safety issues. Supplement 21, Revision insertion instructions

International Nuclear Information System (INIS)

1996-01-01

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative.

From extended integrity monitoring to the safety evaluation of satellite-based localisation system

International Nuclear Information System (INIS)

Legrand, Cyril; Beugin, Julie; Marais, Juliette; Conrard, Blaise; El-Koursi, El-Miloudi; Berbineau, Marion

2016-01-01

Global Navigation Satellite Systems (GNSS) such as GPS, already used in aeronautics for safety-related applications, can play a major role in railway safety by allowing a train to locate itself safely. However, in order to implement this positioning solution in any embedded system, its performances must be evaluated according to railway standards. The evaluation of GNSS performances is not based on the same attributes class than RAMS evaluation. Face to these diffculties, we propose to express the integrity attribute, performance of satellite-based localisation. This attribute comes from aeronautical standards and for a hybridised GNSS with inertial system. To achieve this objective, the integrity attribute must be extended to this kind of system and algorithms initially devoted to GNSS integrity monitoring only must be adapted. Thereafter, the formalisation of this integrity attribute permits us to analyse the safety quantitatively through the probabilities of integrity risk and wrong-side failure. In this paper, after an introductory discussion about the use of localisation systems in railway safety context together with integrity issues, a particular integrity monitoring is proposed and described. The detection events of this algorithm permit us to conclude about safety level of satellite-based localisation system.

System safety analysis of an autonomous mobile robot

International Nuclear Information System (INIS)

Bartos, R.J.

1994-01-01

Analysis of the safety of operating and maintaining the Stored Waste Autonomous Mobile Inspector (SWAMI) II in a hazardous environment at the Fernald Environmental Management Project (FEMP) was completed. The SWAMI II is a version of a commercial robot, the HelpMate trademark robot produced by the Transitions Research Corporation, which is being updated to incorporate the systems required for inspecting mixed toxic chemical and radioactive waste drums at the FEMP. It also has modified obstacle detection and collision avoidance subsystems. The robot will autonomously travel down the aisles in storage warehouses to record images of containers and collect other data which are transmitted to an inspector at a remote computer terminal. A previous study showed the SWAMI II has economic feasibility. The SWAMI II will more accurately locate radioactive contamination than human inspectors. This thesis includes a System Safety Hazard Analysis and a quantitative Fault Tree Analysis (FTA). The objectives of the analyses are to prevent potentially serious events and to derive a comprehensive set of safety requirements from which the safety of the SWAMI II and other autonomous mobile robots can be evaluated. The Computer-Aided Fault Tree Analysis (CAFTA copyright) software is utilized for the FTA. The FTA shows that more than 99% of the safety risk occurs during maintenance, and that when the derived safety requirements are implemented the rate of serious events is reduced to below one event per million operating hours. Training and procedures in SWAMI II operation and maintenance provide an added safety margin. This study will promote the safe use of the SWAMI II and other autonomous mobile robots in the emerging technology of mobile robotic inspection

Improving safety culture through the health and safety organization: a case study.

Science.gov (United States)

Nielsen, Kent J

2014-02-01

International research indicates that internal health and safety organizations (HSO) and health and safety committees (HSC) do not have the intended impact on companies' safety performance. The aim of this case study at an industrial plant was to test whether the HSO can improve company safety culture by creating more and better safety-related interactions both within the HSO and between HSO members and the shop-floor. A quasi-experimental single case study design based on action research with both quantitative and qualitative measures was used. Based on baseline mapping of safety culture and the efficiency of the HSO three developmental processes were started aimed at the HSC, the whole HSO, and the safety representatives, respectively. Results at follow-up indicated a marked improvement in HSO performance, interaction patterns concerning safety, safety culture indicators, and a changed trend in injury rates. These improvements are interpreted as cultural change because an organizational double-loop learning process leading to modification of the basic assumptions could be identified. The study provides evidence that the HSO can improve company safety culture by focusing on safety-related interactions. © 2013. Published by Elsevier Ltd and National Safety Council.

Functional Safety Specification of Communication Profile PROFIsafe

Directory of Open Access Journals (Sweden)

Jan Rofar

2006-01-01

Full Text Available Paper maps the trends in area of safety-related communication within PROFIBUS and PROFINET industry networks. There are analyses safety measures and Fail-safe parameters of PROFIsafe profile in version V2 and their localisation in Safety Communication Layer SCL, which guarantees Safety Integrity Level SIL according to standard IEC 61508. The last chapter analyses the reaction in the event of fault during transmission of messages.

A survey on the quantitative incorporation organizational factors into PSA

International Nuclear Information System (INIS)

Park, S. Z.; Jea, M. S.; Ahn, N. S.

2002-01-01

The effects of organizational factors on the human performance and safety in nuclear power plants have been known through the results of research for several years. The organizational factor, which belongs to 11 elements of PSR (Periodic Safety Review), has been an important research area. In this study the state-of-the-art of qualitative and quantitative evaluation methodologies on organizational factors has been surveyed. The results of this study may contribute to developing a quantitative evaluation methodology on organizational factors as well as the basic research for conducting the PSR research, and for incorporating the quality of organization factors into PSA

Advanced handbook for accident analyses of German nuclear power plants; Weiterentwicklung eines Handbuches fuer Stoerfallanalysen deutscher Kernkraftwerke

Energy Technology Data Exchange (ETDEWEB)

Kerner, Alexander; Broecker, Annette; Hartung, Juergen; Mayer, Gerhard; Pallas Moner, Guim

2014-09-15

The advanced handbook of safety analyses (HSA) comprises a comprehensive electronic collection of knowledge for the compilation and conduction of safety analyses in the area of reactor, plant and containment behaviour as well as results of existing safety analyses (performed by GRS in the past) with characteristic specifications and further background information. In addition, know-how from the analysis software development and validation process is presented and relevant rules and regulations with regard to safety demonstration are provided. The HSA comprehensively covers the topic thermo-hydraulic safety analyses (except natural hazards, man-made hazards and malicious acts) for German pressurized and boiling water reactors for power and non-power operational states. In principle, the structure of the HSA-content represents the analytical approach utilized by safety analyses and applying the knowledge from safety analyses to technical support services. On the basis of a multilevel preparation of information to the topics ''compilation of safety analyses'', ''compilation of data bases'', ''assessment of safety analyses'', ''performed safety analyses'', ''rules and regulation'' and ''ATHLET-validation'' the HSA addresses users with different background, allowing them to enter the HSA at different levels. Moreover, the HSA serves as a reference book, which is designed future-oriented, freely configurable related to the content, completely integrated into the GRS internal portal and prepared to be used by a growing user group.

A guide on the elicitation of expert knowledge in constructing BBN for quantitative reliability assessment of safety critical software

International Nuclear Information System (INIS)

Eom, H. S.; Kang, H. G.; Chang, S. C.; Ha, J. J.

2003-08-01

This report describes the methodology which could elicit probabilistic representation from the experts' knowledge or qualitative data. It is necessary to elicit expert's knowledge while we quantitatively assess the reliability of safety critical software using Bayesian Belief Nets(BBNs). Especially in composing the node probability table and in making out the input data for BBN model, experts' qualitative judgment or qualitative data should be converted into probabilistic representation. This conversion process is vulnerable to bias or error. The purpose of the report is to provide the guideline to avoid the occurrence of this kinds of bias/error or to eliminate them which is included in the existing data prepared by experts. The contents of the report are: o The types and the explanation of bias and error The types of bias and error which might be occur in the process of eliciting the expert's knowledge. o The procedure of expert's judgment elicitation. The process and techniques to avoid bias and error in eliciting the expert's judgments. o The examples of expert's knowledge appeared in the BBNs The examples of expert's knowledge (probability values) appeared in the BBNs for assessing the safety of digital system

Patterns of use and impact of standardised MedDRA query analyses on the safety evaluation and review of new drug and biologics license applications.

Directory of Open Access Journals (Sweden)

Lin-Chau Chang

Full Text Available Standardised MedDRA Queries (SMQs have been developed since the early 2000's and used by academia, industry, public health, and government sectors for detecting safety signals in adverse event safety databases. The purpose of the present study is to characterize how SMQs are used and the impact in safety analyses for New Drug Application (NDA and Biologics License Application (BLA submissions to the United States Food and Drug Administration (USFDA.We used the PharmaPendium database to capture SMQ use in Summary Basis of Approvals (SBoAs of drugs and biologics approved by the USFDA. Characteristics of the drugs and the SMQ use were employed to evaluate the role of SMQ safety analyses in regulatory decisions and the veracity of signals they revealed.A comprehensive search of the SBoAs yielded 184 regulatory submissions approved from 2006 to 2015. Search strategies more frequently utilized restrictive searches with "narrow terms" to enhance specificity over strategies using "broad terms" to increase sensitivity, while some involved modification of search terms. A majority (59% of 1290 searches used descriptive statistics, however inferential statistics were utilized in 35% of them. Commentary from reviewers and supervisory staff suggested that a small, yet notable percentage (18% of 1290 searches supported regulatory decisions. The searches with regulatory impact were found in 73 submissions (40% of the submissions investigated. Most searches (75% of 227 searches with regulatory implications described how the searches were confirmed, indicating prudence in the decision-making process.SMQs have an increasing role in the presentation and review of safety analysis for NDAs/BLAs and their regulatory reviews. This study suggests that SMQs are best used for screening process, with descriptive statistics, description of SMQ modifications, and systematic verification of cases which is crucial for drawing regulatory conclusions.

Patterns of use and impact of standardised MedDRA query analyses on the safety evaluation and review of new drug and biologics license applications.

Science.gov (United States)

Chang, Lin-Chau; Mahmood, Riaz; Qureshi, Samina; Breder, Christopher D

2017-01-01

Standardised MedDRA Queries (SMQs) have been developed since the early 2000's and used by academia, industry, public health, and government sectors for detecting safety signals in adverse event safety databases. The purpose of the present study is to characterize how SMQs are used and the impact in safety analyses for New Drug Application (NDA) and Biologics License Application (BLA) submissions to the United States Food and Drug Administration (USFDA). We used the PharmaPendium database to capture SMQ use in Summary Basis of Approvals (SBoAs) of drugs and biologics approved by the USFDA. Characteristics of the drugs and the SMQ use were employed to evaluate the role of SMQ safety analyses in regulatory decisions and the veracity of signals they revealed. A comprehensive search of the SBoAs yielded 184 regulatory submissions approved from 2006 to 2015. Search strategies more frequently utilized restrictive searches with "narrow terms" to enhance specificity over strategies using "broad terms" to increase sensitivity, while some involved modification of search terms. A majority (59%) of 1290 searches used descriptive statistics, however inferential statistics were utilized in 35% of them. Commentary from reviewers and supervisory staff suggested that a small, yet notable percentage (18%) of 1290 searches supported regulatory decisions. The searches with regulatory impact were found in 73 submissions (40% of the submissions investigated). Most searches (75% of 227 searches) with regulatory implications described how the searches were confirmed, indicating prudence in the decision-making process. SMQs have an increasing role in the presentation and review of safety analysis for NDAs/BLAs and their regulatory reviews. This study suggests that SMQs are best used for screening process, with descriptive statistics, description of SMQ modifications, and systematic verification of cases which is crucial for drawing regulatory conclusions.

A Bayesian belief nets based quantitative software reliability assessment for PSA: COTS case study

International Nuclear Information System (INIS)

Eom, H. S.; Sung, T. Y.; Jeong, H. S.; Park, J. H.; Kang, H. G.; Lee, K. Y.; Park, J. K

2002-03-01

Current reliability assessments of safety critical software embedded in the digital systems in nuclear power plants are based on the rule-based qualitative assessment methods. Then recently practical needs require the quantitative features of software reliability for Probabilistic Safety Assessment (PSA) that is one of important methods being used in assessing the whole safety of nuclear power plant. But conventional quantitative software reliability assessment methods are not enough to get the necessary results in assessing the safety critical software used in nuclear power plants. Thus, current reliability assessment methods for these digital systems exclude the software part or use arbitrary values for the software reliability in the assessment. This reports discusses a Bayesian Belief Nets (BBN) based quantification method that models current qualitative software assessment in formal way and produces quantitative results required for PSA. Commercial Off-The-Shelf (COTS) software dedication process that KAERI developed was applied to the discussed BBN based method for evaluating the plausibility of the proposed method in PSA

Investment appraisal using quantitative risk analysis.

Science.gov (United States)

Johansson, Henrik

2002-07-01

Investment appraisal concerned with investments in fire safety systems is discussed. Particular attention is directed at evaluating, in terms of the Bayesian decision theory, the risk reduction that investment in a fire safety system involves. It is shown how the monetary value of the change from a building design without any specific fire protection system to one including such a system can be estimated by use of quantitative risk analysis, the results of which are expressed in terms of a Risk-adjusted net present value. This represents the intrinsic monetary value of investing in the fire safety system. The method suggested is exemplified by a case study performed in an Avesta Sheffield factory.

A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

Energy Technology Data Exchange (ETDEWEB)

Park, Sunghyun; Kim, Namyeong; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2015-10-15

In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant.

A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

International Nuclear Information System (INIS)

Park, Sunghyun; Kim, Namyeong; Jae, Moosung

2015-01-01

In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant

A prioritization of generic safety issues. Supplement 19, Revision insertion instructions

International Nuclear Information System (INIS)

1995-11-01

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative. This document provides revisions and amendments to the report

A prioritization of generic safety issues. Supplement 19, Revision insertion instructions

Energy Technology Data Exchange (ETDEWEB)

None

1995-11-01

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative. This document provides revisions and amendments to the report.

Safety targets for nuclear power plants

International Nuclear Information System (INIS)

Herttrich, P.M.

1985-01-01

By taking as an example the safety targets of the American nuclear energy authority US-NRC, this paper explains what is meant by global, quantitative safety targets for nuclear power plants and what expectations are associated with the selecton of such safety targets. It is shown how probabilistic methods can be an appropriate completion of proven deterministic methods and what are the sectors where their application may become important in future. (orig./HP) [de

Safety and licensing requirements in the Republic of South Africa

International Nuclear Information System (INIS)

Simpson, D.M.; Langford, E.L.

1986-01-01

The principles for licensing of nuclear installations in South Africa are based on the control of mortality risk to the operators of an installation and the population resident in the vicinity of the site. This paper describes the development of this safety philosophy, and the nuclear licensing process used in this country. The structure of the nuclear regulatory function is briefly described, including the respective roles of the Atomic Energy Corporation, Licencing Branch and the Council for Nuclear Safety. The development of risk criteria and quantitative release magnitude-probability criteria for radioactive material is outlined. Tasks that have to be undertaken by a potential waste disposal site licensee before a site licence is issued are described. Once the facility is commissioned periodic monitoring procedures will have to be adopted throughout the lifetime of the facility. The scope of typical monitoring activities is outlined and the ongoing analyses to be performed and the records to be kept are discussed

Quality management and perceptions of teamwork and safety climate in European hospitals.

Science.gov (United States)

Kristensen, Solvejg; Hammer, Antje; Bartels, Paul; Suñol, Rosa; Groene, Oliver; Thompson, Caroline A; Arah, Onyebuchi A; Kutaj-Wasikowska, Halina; Michel, Philippe; Wagner, Cordula

2015-12-01

This study aimed to investigate the associations of quality management systems with teamwork and safety climate, and to describe and compare differences in perceptions of teamwork climate and safety climate among clinical leaders and frontline clinicians. We used a multi-method, cross-sectional approach to collect survey data of quality management systems and perceived teamwork and safety climate. Our data analyses included descriptive and multilevel regression methods. Data on implementation of quality management system from seven European countries were evaluated including patient safety culture surveys from 3622 clinical leaders and 4903 frontline clinicians. Perceived teamwork and safety climate. Teamwork climate was reported as positive by 67% of clinical leaders and 43% of frontline clinicians. Safety climate was perceived as positive by 54% of clinical leaders and 32% of frontline clinicians. We found positive associations between implementation of quality management systems and teamwork and safety climate. Our findings, which should be placed in a broader clinical quality improvement context, point to the importance of quality management systems as a supportive structural feature for promoting teamwork and safety climate. To gain a deeper understanding of this association, further qualitative and quantitative studies using longitudinally collected data are recommended. The study also confirms that more clinical leaders than frontline clinicians have a positive perception of teamwork and safety climate. Such differences should be accounted for in daily clinical practice and when tailoring initiatives to improve teamwork and safety climate. © The Author 2015. Published by Oxford University Press in association with the International Society for Quality in Health Care; all rights reserved.

Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

CERN Document Server

Marshall, T; Topilski, L; Merrill, B

2002-01-01

For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

Safety of nuclear ships

International Nuclear Information System (INIS)

1978-01-01

Interest in the utilization of nuclear steam supply systems for merchant ships and icebreakers has recently increased considerably due to the sharp rise in oil prices and the continuing trend towards larger and faster merchant ships. Canada, for example, is considering construction of an icebreaker in the near future. On the other hand, an accident which could result in serious damage to or the sinking of a nuclear ship is potentially far more dangerous to the general public than a similar accident with a conventional ship. Therefore, it was very important to evaluate in an international forum the safety of nuclear ships in the light of our contemporary safety philosophy, taking into account the results of cumulative operating experience with nuclear ships in operation. The philosophy and safety requirement for land-based nuclear installations were outlined because of many common features for both land-based nuclear installations and nuclear ships. Nevertheless, essential specific safety requirements for nuclear ships must always be considered, and the work on safety problems for nuclear ships sponsored by the NEA was regarded as an important step towards developing an international code of practice by IMCO on the safety of nuclear merchant ships. One session was devoted to the quantitative assessment of nuclear ship safety. The probability technique of an accident risk assessment for nuclear power plants is well known and widely used. Its modification, to make it applicable to nuclear propelled merchant ships, was discussed in some papers. Mathematical models for describing various postulated accidents with nuclear ships were developed and reported by several speakers. Several papers discussed a loss-of-coolant accident (LOCA) with nuclear steam supply systems of nuclear ships and engineering design features to prevent a radioactive effluence after LOCA. Other types of postulated accidents with reactors and systems in static and dynamic conditions were also

Pulmonary nodule characterization: A comparison of conventional with quantitative and visual semi-quantitative analyses using contrast enhancement maps

International Nuclear Information System (INIS)

Petkovska, Iva; Shah, Sumit K.; McNitt-Gray, Michael F.; Goldin, Jonathan G.; Brown, Matthew S.; Kim, Hyun J.; Brown, Kathleen; Aberle, Denise R.

2006-01-01

Purpose: To determine whether conventional nodule densitometry or analysis based on contrast enhancement maps of indeterminate lung nodules imaged with contrast-enhanced CT can distinguish benign from malignant lung nodules. Materials and method: Thin section, contrast-enhanced CT (baseline, and post-contrast series acquired at 45, 90,180, and 360 s) was performed on 29 patients with indeterminate lung nodules (14 benign, 15 malignant). A thoracic radiologist identified the boundary of each nodule using semi-automated contouring to form a 3D region-of-interest (ROI) on each image series. The post-contrast series having the maximum mean enhancement was then volumetrically registered to the baseline series. The two series were subtracted volumetrically and the subtracted voxels were quantized into seven color-coded bins, forming a contrast enhancement map (CEM). Conventional nodule densitometry was performed to obtain the maximum difference in mean enhancement values for each nodule from a circular ROI. Three thoracic radiologists performed visual semi-quantitative analysis of each nodule, scoring each map for: (a) magnitude and (b) heterogeneity of enhancement throughout the entire volume of the nodule on a five-point scale. Receiver operator characteristic (ROC) analysis was conducted on these features to evaluate their diagnostic efficacy. Finally, 14 quantitative texture features were calculated for each map. A statistical analysis was performed to combine the 14 texture features to a single factor. ROC analysis of the derived aggregate factor was done as an indicator of malignancy. All features were analyzed for differences between benign and malignant nodules. Results: Using 15 HU as a threshold, 93% (14/15) of malignant and 79% (11/14) of benign nodules demonstrated enhancement. The ROC curve when higher values of enhancement indicate malignancy was generated and area under the curve (AUC) was 0.76. The visually scored magnitude of enhancement was found to be

Quantitative X-ray Map Analyser (Q-XRMA): A new GIS-based statistical approach to Mineral Image Analysis

Science.gov (United States)

Ortolano, Gaetano; Visalli, Roberto; Godard, Gaston; Cirrincione, Rosolino

2018-06-01

We present a new ArcGIS®-based tool developed in the Python programming language for calibrating EDS/WDS X-ray element maps, with the aim of acquiring quantitative information of petrological interest. The calibration procedure is based on a multiple linear regression technique that takes into account interdependence among elements and is constrained by the stoichiometry of minerals. The procedure requires an appropriate number of spot analyses for use as internal standards and provides several test indexes for a rapid check of calibration accuracy. The code is based on an earlier image-processing tool designed primarily for classifying minerals in X-ray element maps; the original Python code has now been enhanced to yield calibrated maps of mineral end-members or the chemical parameters of each classified mineral. The semi-automated procedure can be used to extract a dataset that is automatically stored within queryable tables. As a case study, the software was applied to an amphibolite-facies garnet-bearing micaschist. The calibrated images obtained for both anhydrous (i.e., garnet and plagioclase) and hydrous (i.e., biotite) phases show a good fit with corresponding electron microprobe analyses. This new GIS-based tool package can thus find useful application in petrology and materials science research. Moreover, the huge quantity of data extracted opens new opportunities for the development of a thin-section microchemical database that, using a GIS platform, can be linked with other major global geoscience databases.

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Ualikhan Zhiyenbayev [KAIST, Daejeon (Korea, Republic of); Chung, Dae Wook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether.

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

International Nuclear Information System (INIS)

Ualikhan Zhiyenbayev; Chung, Dae Wook

2015-01-01

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether

Dimensionality Analyses of the "GRE"® revised General Test Verbal and Quantitative Measures. ETS GRE® Board Research Report. ETS GRE®-16-02. ETS Research Report. RR-16-20

Science.gov (United States)

Robin, Frédéric; Bejar, Isaac; Liang, Longjuan; Rijmen, Frank

2016-01-01

Exploratory and confirmatory factor analyses of domestic data from the" GRE"® revised General Test, introduced in 2011, were conducted separately for the verbal (VBL) and quantitative (QNT) reasoning measures to evaluate the unidimensionality and local independence assumptions required by item response theory (IRT). Results based on data…

Quantifying the effectiveness of ITS in improving safety of VRUs

NARCIS (Netherlands)

Silla, A.; Rämä, P.; Leden, L.; Noort, M. van; Kruijff, J. de; Bell, D.; Morris, A.; Hancox, G.; Scholliers, J.

2017-01-01

This paper presents the results of a safety impact assessment, providing quantitative estimates of the safety impacts of ten intelligent transport systems (ITS) which were designed to improve safety, mobility and comfort of vulnerable road users (VRUs). The evaluation method originally developed to

Deterministic quantitative risk assessment development

Energy Technology Data Exchange (ETDEWEB)

Dawson, Jane; Colquhoun, Iain [PII Pipeline Solutions Business of GE Oil and Gas, Cramlington Northumberland (United Kingdom)

2009-07-01

Current risk assessment practice in pipeline integrity management is to use a semi-quantitative index-based or model based methodology. This approach has been found to be very flexible and provide useful results for identifying high risk areas and for prioritizing physical integrity assessments. However, as pipeline operators progressively adopt an operating strategy of continual risk reduction with a view to minimizing total expenditures within safety, environmental, and reliability constraints, the need for quantitative assessments of risk levels is becoming evident. Whereas reliability based quantitative risk assessments can be and are routinely carried out on a site-specific basis, they require significant amounts of quantitative data for the results to be meaningful. This need for detailed and reliable data tends to make these methods unwieldy for system-wide risk k assessment applications. This paper describes methods for estimating risk quantitatively through the calibration of semi-quantitative estimates to failure rates for peer pipeline systems. The methods involve the analysis of the failure rate distribution, and techniques for mapping the rate to the distribution of likelihoods available from currently available semi-quantitative programs. By applying point value probabilities to the failure rates, deterministic quantitative risk assessment (QRA) provides greater rigor and objectivity than can usually be achieved through the implementation of semi-quantitative risk assessment results. The method permits a fully quantitative approach or a mixture of QRA and semi-QRA to suit the operator's data availability and quality, and analysis needs. For example, consequence analysis can be quantitative or can address qualitative ranges for consequence categories. Likewise, failure likelihoods can be output as classical probabilities or as expected failure frequencies as required. (author)

Quantitative mass spectrometry: an overview

Science.gov (United States)

Urban, Pawel L.

2016-10-01

Mass spectrometry (MS) is a mainstream chemical analysis technique in the twenty-first century. It has contributed to numerous discoveries in chemistry, physics and biochemistry. Hundreds of research laboratories scattered all over the world use MS every day to investigate fundamental phenomena on the molecular level. MS is also widely used by industry-especially in drug discovery, quality control and food safety protocols. In some cases, mass spectrometers are indispensable and irreplaceable by any other metrological tools. The uniqueness of MS is due to the fact that it enables direct identification of molecules based on the mass-to-charge ratios as well as fragmentation patterns. Thus, for several decades now, MS has been used in qualitative chemical analysis. To address the pressing need for quantitative molecular measurements, a number of laboratories focused on technological and methodological improvements that could render MS a fully quantitative metrological platform. In this theme issue, the experts working for some of those laboratories share their knowledge and enthusiasm about quantitative MS. I hope this theme issue will benefit readers, and foster fundamental and applied research based on quantitative MS measurements. This article is part of the themed issue 'Quantitative mass spectrometry'.

Dependency Defence and Dependency Analysis Guidance. Volume 2: Appendix 3-8. How to analyse and protect against dependent failures. Summary report of the Nordic Working Group on Common Cause Failure Analysis

International Nuclear Information System (INIS)

Johanson, Gunnar; Hellstroem, Per; Makamo, Tuomas; Bento, Jean-Pierre; Knochenhauer, Michael; Poern, Kurt

2003-10-01

The safety systems in Nordic nuclear power plants are characterised by substantial redundancy and/or diversification in safety critical functions, as well as by physical separation of critical safety systems, including their support functions. Viewed together with the evident additional fact, that the single failure criterion has been systematically applied in the design of safety systems, this means that the plant risk profile as calculated in existing PSA:s is usually strongly dominated by failures caused by dependencies resulting in the loss of more than one system sub. The overall objective with the working group is to support safety by studying potential and real CCF events, process statistical data and report conclusions and recommendations that can improve the understanding of these events eventually resulting in increased safety. The result is intended for application in NPP operation, maintenance, inspection and risk assessments. The NAFCS project is part of the activities of the Nordic PSA Group (NPSAG), and is financed jointly by the Nordic utilities and authorities. The work is divided into one quantitative and one qualitative part with the following specific objectives: Qualitative objectives-The goal with the qualitative analysis is to compile experience data and generate insights in terms of relevant failure mechanisms and effective CCF protection measures. The results shall be presented as a guide with checklists and recommendations on how to identify current CCF protection standard and improvement possibilities regarding CCF defences decreasing the CCF vulnerability. Quantitative objectives-The goal with the quantitative analysis is to prepare a Nordic C-book where quantitative insights as Impact Vectors and CCF parameters for different redundancy levels are presented. Uncertainties in CCF data shall be reduced as much as possible. The high redundancy systems sensitivity to CCF events demand a well structured quantitative analysis in support of

A volumetric meter chip for point-of-care quantitative detection of bovine catalase for food safety control

International Nuclear Information System (INIS)

Cui, Xingye; Hu, Jie; Choi, Jane Ru; Huang, Yalin; Wang, Xuemin; Lu, Tian Jian; Xu, Feng

2016-01-01

A volumetric meter chip was developed for quantitative point-of-care (POC) analysis of bovine catalase, a bioindicator of bovine mastitis, in milk samples. The meter chip displays multiplexed quantitative results by presenting the distance of ink bar advancement that is detectable by the naked eye. The meter chip comprises a poly(methyl methacrylate) (PMMA) layer, a double-sided adhesive (DSA) layer and a glass slide layer fabricated by the laser-etching method, which is typically simple, rapid (∼3 min per chip), and cost effective (∼$0.2 per chip). Specially designed “U shape” reaction cells are covered by an adhesive tape that serves as an on-off switch, enabling the simple operation of the assay. As a proof of concept, we employed the developed meter chip for the quantification of bovine catalase in raw milk samples to detect catalase concentrations as low as 20 μg/mL. The meter chip has great potential to detect various target analytes for a wide range of POC applications. - Highlights: • The meter chip is a standalone point-of-care diagnostic tool with visible readouts of quantification results. • A fast and low cost fabrication protocol (~3 min and ~$0.2 per chip) of meter chip was proposed. • The chip may hold the potential for rapid scaning of bovine mastitis in cattle farms for food safety control.

A volumetric meter chip for point-of-care quantitative detection of bovine catalase for food safety control

Energy Technology Data Exchange (ETDEWEB)

Cui, Xingye; Hu, Jie; Choi, Jane Ru; Huang, Yalin; Wang, Xuemin [The Key Laboratory of Biomedical Information Engineering of Ministry of Education, School of Life Science and Technology, Xi' an Jiaotong University, Xi' an, 710049 (China); Bioinspired Engineering and Biomechanics Center (BEBC), Xi' an Jiaotong University, Xi' an, 710049 (China); Lu, Tian Jian, E-mail: tjlu@mail.xjtu.edu.cn [Bioinspired Engineering and Biomechanics Center (BEBC), Xi' an Jiaotong University, Xi' an, 710049 (China); Xu, Feng, E-mail: fengxu@mail.xjtu.edu.cn [The Key Laboratory of Biomedical Information Engineering of Ministry of Education, School of Life Science and Technology, Xi' an Jiaotong University, Xi' an, 710049 (China); Bioinspired Engineering and Biomechanics Center (BEBC), Xi' an Jiaotong University, Xi' an, 710049 (China)

2016-09-07

A volumetric meter chip was developed for quantitative point-of-care (POC) analysis of bovine catalase, a bioindicator of bovine mastitis, in milk samples. The meter chip displays multiplexed quantitative results by presenting the distance of ink bar advancement that is detectable by the naked eye. The meter chip comprises a poly(methyl methacrylate) (PMMA) layer, a double-sided adhesive (DSA) layer and a glass slide layer fabricated by the laser-etching method, which is typically simple, rapid (∼3 min per chip), and cost effective (∼$0.2 per chip). Specially designed “U shape” reaction cells are covered by an adhesive tape that serves as an on-off switch, enabling the simple operation of the assay. As a proof of concept, we employed the developed meter chip for the quantification of bovine catalase in raw milk samples to detect catalase concentrations as low as 20 μg/mL. The meter chip has great potential to detect various target analytes for a wide range of POC applications. - Highlights: • The meter chip is a standalone point-of-care diagnostic tool with visible readouts of quantification results. • A fast and low cost fabrication protocol (~3 min and ~$0.2 per chip) of meter chip was proposed. • The chip may hold the potential for rapid scaning of bovine mastitis in cattle farms for food safety control.

Review of Ontario Hydro Pickering 'A' and Bruce 'A' nuclear generating stations' accident analyses

International Nuclear Information System (INIS)

Serdula, K.J.

1988-01-01

Deterministic safety analysis for the Pickering 'A' and Bruce 'A' nuclear generating stations were reviewed. The methodology used in the evaluation and assessment was based on the concept of 'N' critical parameters defining an N-dimensional safety parameter space. The reviewed accident analyses were evaluated and assessed based on their demonstrated safety coverage for credible values and trajectories of the critical parameters within this N-dimensional safety parameter space. The reported assessment did not consider probability of occurrence of event. The reviewed analyses were extensive for potential occurrence of accidents under normal steady-state operating conditions. These analyses demonstrated an adequate assurance of safety for the analyzed conditions. However, even for these reactor conditions, items have been identified for consideration of review and/or further study, which would provide a greater assurance of safety in the event of an accident. Accident analyses based on a plant in a normal transient operating state or in an off-normal condition but within the allowable operating envelope are not as extensive. Improvements in demonstrations and/or justifications of safety upon potential occurrence of accidents would provide further assurance of adequacy of safety under these conditions. Some events under these conditions have not been analyzed because of their judged low probability; however, accident analyses in this area should be considered. Recommendations are presented relating to these items; it is also recommended that further study is needed of the Pickering 'A' special safety systems

Introduction to safety theory

International Nuclear Information System (INIS)

Meyna, A.

1982-01-01

After a general introduction to safety theory, safety characteristics are defined and quantified. This is followed by a calculation of the safety characteristics of simple, safety-relevant systems in general and in consideration of common-mode errors. The qualitative and quantitative role of human errors is discussed for various models, and a simple man-machine model is developed for investigation of common-mode errors and human error. The main part of the paper deals with safety analysis in complex systems. After a general review, the common inductive and deductive methods of analysis are presented and commented on and their fields of application discussed. Analytical and simulation codes are presented as methods of evaluation for big, complex event trees - i.e. ''hazard trees in the sense of safety engineering (as a subset of safety relevance). After a basic classification and mathematical formulation of Markovian processes, the author shows that these may be used successfully for calculation of safety characteristics if transition rates are constant and if the number of system states is limited. (orig./RW) [de

Quantitative security and safety analysis with attack-fault trees

NARCIS (Netherlands)

Kumar, Rajesh; Stoelinga, Mariëlle Ida Antoinette

2017-01-01

Cyber physical systems, like power plants, medical devices and data centers have to meet high standards, both in terms of safety (i.e. absence of unintentional failures) and security (i.e. no disruptions due to malicious attacks). This paper presents attack fault trees (AFTs), a formalism that

Patient participation in patient safety and nursing input - a systematic review.

Science.gov (United States)

Vaismoradi, Mojtaba; Jordan, Sue; Kangasniemi, Mari

2015-03-01

This systematic review aims to synthesise the existing research on how patients participate in patient safety initiatives. Ambiguities remain about how patients participate in routine measures designed to promote patient safety. Systematic review using integrative methods. Electronic databases were searched using keywords describing patient involvement, nursing input and patient safety initiatives to retrieve empirical research published between 2007 and 2013. Findings were synthesized using the theoretical domains of Vincent's framework for analysing risk and safety in clinical practice: "patient", "healthcare provider", "task", "work environment", "organisation & management". We identified 17 empirical research papers: four qualitative, one mixed-method and 12 quantitative designs. All 17 papers indicated that patients can participate in safety initiatives. Improving patient participation in patient safety necessitates considering the patient as a person, the nurse as healthcare provider, the task of participation and the clinical environment. Patients' knowledge, health conditions, beliefs and experiences influence their decisions to engage in patient safety initiatives. An important component of the management of long-term conditions is to ensure that patients have sufficient knowledge to participate. Healthcare providers may need further professional development in patient education and patient care management to promote patient involvement in patient safety, and ensure that patients understand that they are 'allowed' to inform nurses of adverse events or errors. A healthcare system characterised by patient-centredness and mutual acknowledgement will support patient participation in safety practices. Further research is required to improve international knowledge of patient participation in patient safety in different disciplines, contexts and cultures. Patients have a significant role to play in enhancing their own safety while receiving hospital care. This

WALKABILITY IN HISTORIC URBAN SPACES: TESTING THE SAFETY AND SECURITY IN MARTYRS' SQUARE IN TRIPOLI

Directory of Open Access Journals (Sweden)

Khairi M. Al-bashir Abdulla

2017-11-01

Full Text Available Much of the built environment design literature focuses on a composite of walkable spaces variables such as density, diversity, and destination accessibility.  One of the most effective factors in walkability is “safety and security”. There is an evident gap in understanding the perceived ability of Libyan public spaces to support walkability. This paper aims to investigate the effectiveness of “walkability” in traditional Libyan urban spaces and analyse the relationship between walking, a safe and secure environment, and its impact on a heritage site in Tripoli city centre. The perceived personal safety of 140 users of the heritage site “Martyrs' Square” were measured; this research is studying the quality of environment and users’ interaction with their environmental issues relating to the study area. Mixed methods were used in this research: this study used both quantitative and qualitative methods to gather information; the quantitative took the form of a questionnaire; and the qualitative took the form of observations. Analysis of quantitative data was conducted with SPSS software; the survey was conducted from August 2016 to September 2016. The results of this study are useful for urban planning, to classify the walkable urban space elements, which could improve the level of walkability in Libyan cities and create sustainable and liveable urban spaces.

SRP reactor safety evolution

International Nuclear Information System (INIS)

Rankin, D.B.

1984-01-01

The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

Nirex safety assessment research programme: 1987/88

International Nuclear Information System (INIS)

George, D.; Hodgkinson, D.P.

1987-01-01

The Nirex Safety Assessment Research programme's objective is to provide information for the radiological safety case for disposing low-level and intermediate-level radioactive wastes in underground repositories. The programme covers a wide range of experimental studies and mathematical modelling for the near and far field. It attempts to develop a quantitative understanding of events and processes which have an impact on the safety of radioactive waste disposal. (U.K.)

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

Quantitative assessment of probability of failing safely for the safety instrumented system using reliability block diagram method

International Nuclear Information System (INIS)

Jin, Jianghong; Pang, Lei; Zhao, Shoutang; Hu, Bin

2015-01-01

Highlights: • Models of PFS for SIS were established by using the reliability block diagram. • The more accurate calculation of PFS for SIS can be acquired by using SL. • Degraded operation of complex SIS does not affect the availability of SIS. • The safe undetected failure is the largest contribution to the PFS of SIS. - Abstract: The spurious trip of safety instrumented system (SIS) brings great economic losses to production. How to ensure the safety instrumented system is reliable and available has been put on the schedule. But the existing models on spurious trip rate (STR) or probability of failing safely (PFS) are too simplified and not accurate, in-depth studies of availability to obtain more accurate PFS for SIS are required. Based on the analysis of factors that influence the PFS for the SIS, using reliability block diagram method (RBD), the quantitative study of PFS for the SIS is carried out, and gives some application examples. The results show that, the common cause failure will increase the PFS; degraded operation does not affect the availability of the SIS; if the equipment was tested and repaired one by one, the unavailability of the SIS can be ignored; the corresponding occurrence time of independent safe undetected failure should be the system lifecycle (SL) rather than the proof test interval and the independent safe undetected failure is the largest contribution to the PFS for the SIS

NPP Temelin safety analysis reports and PSA status

International Nuclear Information System (INIS)

Mlady, O.

1999-01-01

To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

International Nuclear Information System (INIS)

Heard, F.J.

1997-01-01

The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

Quantitative analysis of the effect of complex internals on LMFBR containment during energetic accidents

International Nuclear Information System (INIS)

Zeuch, W.R.; Wang, C.Y.

1985-01-01

This paper discusses the effects of complex internals on the containment response of large LMFBRs during energetic accidents. Results of a series of analyses with the ALICE-II code demonstrate quantitative structural and hydrodynamic effects from parametric variation of reactor internal designs. Effects of various upper internal structure treatments, structural stiffness of the upper internal structure and core support structure, and the location and dimensions of internal components are examined. Results indicate that reduction of primary containment loads can be accomplished through such means as confinement of the core region and avoiding over-strengthened, rigid internals. A study of the beneficial and adverse parameters involved in primary containment response should be helpful in optimizing designs for safety purposes

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

Quantitative risk analysis offshore-Human and organizational factors

International Nuclear Information System (INIS)

Espen Skogdalen, Jon; Vinnem, Jan Erik

2011-01-01

Quantitative Risk Analyses (QRAs) are one of the main tools for risk management within the Norwegian and UK oil and gas industry. Much criticism has been given to the limitations related to the QRA-models and that the QRAs do not include human and organizational factors (HOF-factors). Norway and UK offshore legislation and guidelines require that the HOF-factors are included in the QRAs. A study of 15 QRAs shows that the factors are to some extent included, and there are large differences between the QRAs. The QRAs are categorized into four levels according to the findings. Level 1 QRAs do not describe or comment on the HOF-factors at all. Relevant research projects have been conducted to fulfill the requirements of Level 3 analyses. At this level, there is a systematic collection of data related to HOF. The methods are systematic and documented, and the QRAs are adjusted. None of the QRAs fulfill the Level 4 requirements. Level 4 QRAs include the model and describe the HOF-factors as well as explain how the results should be followed up in the overall risk management. Safety audits by regulatory authorities are probably necessary to point out the direction for QRA and speed up the development.

The Gutenberg English Poetry Corpus: Exemplary Quantitative Narrative Analyses

Directory of Open Access Journals (Sweden)

Arthur M. Jacobs

2018-04-01

Full Text Available This paper describes a corpus of about 3,000 English literary texts with about 250 million words extracted from the Gutenberg project that span a range of genres from both fiction and non-fiction written by more than 130 authors (e.g., Darwin, Dickens, Shakespeare. Quantitative narrative analysis (QNA is used to explore a cleaned subcorpus, the Gutenberg English Poetry Corpus (GEPC, which comprises over 100 poetic texts with around two million words from about 50 authors (e.g., Keats, Joyce, Wordsworth. Some exemplary QNA studies show author similarities based on latent semantic analysis, significant topics for each author or various text-analytic metrics for George Eliot’s poem “How Lisa Loved the King” and James Joyce’s “Chamber Music,” concerning, e.g., lexical diversity or sentiment analysis. The GEPC is particularly suited for research in Digital Humanities, Computational Stylistics, or Neurocognitive Poetics, e.g., as training and test corpus for stimulus development and control in empirical studies.

A quantitative approach to the risk perception associated with nuclear safety

International Nuclear Information System (INIS)

Black, S.

2015-01-01

Subjective risk perception associated with nuclear safety is hard-wired into the general public psyche; but as real as this 'feels', and as much as it requires to be respected in a democracy, misguided risk perception on nuclear safety can create its own perils for humans. The objective of this paper is to create a better understanding of the phenomena of risk perception associated with nuclear safety presented by journalistic media. It will attempt to quantify the manifestation of risk perception associated with nuclear safety by providing comparison between the media coverage of nuclear and industrial accidents of similar magnitude. It will utilise the Fog Index, a mathematical formula that defines the readability of an article, allowing for an unbiased numerical comparison on 'readability' to be derived. Fog Index is expressed as: Fog Index = 0.4(N/S + 100*L/N), where N is the number of words in the article, S is the number of sentences and L is the number of words with 3 syllables or more. To provide consistency, the medium chosen to compare industrial accidents are reports extracted from 'The Times' newspaper, written at the time of the accidents and concerning Chernobyl and Bhopal disasters. 'The Times' is respected newspaper, written for a knowledgeable audience who have an in-depth interest in the news from the UK and abroad; subsequently this causes it to have a relatively high Fog index, compared to its tabloid counterparts. The higher the Fog Index, the more education the reader requires to fully understand the article, a Fog Index of 12 is the limit for the majority of the general public. Research found that reporting of nuclear safety accidents has a Fog Index of approximately 14 while it was only of 10 for Bhopal accident. These values go someway in demonstrating that the complexity of media information on nuclear safety transferred via journalistic media is beyond what can reasonably be expected to be

Safety climate and injuries: an examination of theoretical and empirical relationships.

Science.gov (United States)

Beus, Jeremy M; Payne, Stephanie C; Bergman, Mindy E; Arthur, Winfred

2010-07-01

Our purpose in this study was to meta-analytically address several theoretical and empirical issues regarding the relationships between safety climate and injuries. First, we distinguished between extant safety climate-->injury and injury-->safety climate relationships for both organizational and psychological safety climates. Second, we examined several potential moderators of these relationships. Meta-analyses revealed that injuries were more predictive of organizational safety climate than safety climate was predictive of injuries. Additionally, the injury-->safety climate relationship was stronger for organizational climate than for psychological climate. Moderator analyses revealed that the degree of content contamination in safety climate measures inflated effects, whereas measurement deficiency attenuated effects. Additionally, moderator analyses showed that as the time period over which injuries were assessed lengthened, the safety climate-->injury relationship was attenuated. Supplemental meta-analyses of specific safety climate dimensions also revealed that perceived management commitment to safety is the most robust predictor of occupational injuries. Contrary to expectations, the operationalization of injuries did not meaningfully moderate safety climate-injury relationships. Implications and recommendations for future research and practice are discussed.

Risk assessment of safety data link and network communication in digital safety feature control system of nuclear power plant

International Nuclear Information System (INIS)

Lee, Sang Hun; Son, Kwang Seop; Jung, Wondea; Kang, Hyun Gook

2017-01-01

Highlights: • Safety data communication risk assessment framework and quantitative scheme were proposed. • Fault-tree model of ESFAS unavailability due to safety data communication failure was developed. • Safety data link and network risk were assessed based on various ESF-CCS design specifications. • The effect of fault-tolerant algorithm reliability of safety data network on ESFAS unavailability was assessed. - Abstract: As one of the safety-critical systems in nuclear power plants (NPPs), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for the transmission of safety component actuation signals from the group controllers to loop controllers to effectively accommodate various safety-critical field controllers. Since data communication failure risk in the ESF-CCS has yet to be fully quantified, the ESF-CCS employing data communication systems have not been applied in NPPs. This study therefore developed a fault tree model to assess the data link and data network failure-induced unavailability of a system function used to generate an automated control signal for accident mitigation equipment. The current aim is to provide risk information regarding data communication failure in a digital safety feature control system in consideration of interconnection between controllers and the fault-tolerant algorithm implemented in the target system. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to data link and network failure and its risk effect on safety signal generation failure. This study is expected to provide insight into the risk assessment of safety-critical data communication in a digitalized NPP instrumentation and control system.

Development of a procedure for qualitative and quantitative evaluation of human factors as a part of probabilistic safety assessments of nuclear power plants. Part A

International Nuclear Information System (INIS)

Richei, A.

1998-01-01

The objective of this project is the development of a procedure for the qualitative and quantitative evaluation of human factors in the probabilistic safety assessment for nuclear power plants. The Human Error Rate Assessment and Optimizing System (HEROS) is introduced. The evaluation of a task with HEROS is realized in the three evaluation levels, i.e. 'Management Structure', 'Working Environment' and 'Man-Machine-Interface'. The developed expert system uses the fuzzy set theory for an assessment. For the evaluation of cognitive tasks evaluation criteria are derived also. The validation of the procedure is based on three examples, reflecting the common practice of probabilistic safety assessments and including problems, which cannot, respectively - only insufficiently - be evaluated with the established human risk analysis procedures. HERO applications give plausible and comprehensible results. (orig.) [de

General principles of nuclear safety management related to research reactor decommissioning

International Nuclear Information System (INIS)

Banciu, Ortenzia; Vladescu, Gabriela

2003-01-01

The paper contents the general principles applicable to the decommissioning of research reactors to ensure a proper nuclear safety management, during both decommissioning activities and post decommissioning period. The main objective of decommissioning is to ensure the protection of workers, population and environment against all radiological and non-radiological hazards that could result after a reactor shutdown and dismantling. In the same time, it is necessary, by some proper provisions, to limit the effect of decommissioning for the future generation, according to the new Romanian, IAEA and EU Norms and Regulations. Assurance of nuclear safety during decommissioning process involves, in the first step, to establish of some safety principles and requirements to be taken into account during whole process. In the same time, it is necessary to perform a series of analyses to ensure that the whole process is conducted in a planned and safe manner. The general principles proposed for a proper management of safety during research reactor decommissioning are as follows: - Set-up of all operations included in a Decommissioning Plan; - Set-up and qualitative evaluation of safety problems, which could appear during normal decommissioning process, both radiological and nonradiological risks for workers and public; - Set-up of accident list related to decommissioning process the events that could appear both due to some abnormal working conditions and to some on-site and off-site events like fires, explosions, flooding, earthquake, etc.); - Development and qualitative/ quantitative evaluation of scenarios for each incidents; - Development (and evaluation) of safety indicator system. The safety indicators are the most important tools used to assess the level of nuclear safety during decommissioning process, to discover the weak points and to establish safety measures. The paper contains also, a safety case evaluation (description of facility according to the decommissioning

Safety in relation to risk and benefit

International Nuclear Information System (INIS)

Siddall, E.

1985-01-01

The proper definition and quantification of human safety is discussed and from this basis the historical development of our present very high standard of safety is traced. It is shown that increased safety is closely associated with increased wealth, and the quantitative relationship between then is derived from different sources of evidence. When this factor is applied to the production of wealth by industry, a safety benefit is indicated which exceeds the asserted risks by orders of magnitude. It is concluded that present policies and attitudes in respect to the safety of industry may be diametrically wrong. (orig.) [de

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

International Nuclear Information System (INIS)

Khazanov, A.L.

2005-01-01

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

Gene Set Analyses of Genome-Wide Association Studies on 49 Quantitative Traits Measured in a Single Genetic Epidemiology Dataset

Directory of Open Access Journals (Sweden)

Jihye Kim

2013-09-01

Full Text Available Gene set analysis is a powerful tool for interpreting a genome-wide association study result and is gaining popularity these days. Comparison of the gene sets obtained for a variety of traits measured from a single genetic epidemiology dataset may give insights into the biological mechanisms underlying these traits. Based on the previously published single nucleotide polymorphism (SNP genotype data on 8,842 individuals enrolled in the Korea Association Resource project, we performed a series of systematic genome-wide association analyses for 49 quantitative traits of basic epidemiological, anthropometric, or blood chemistry parameters. Each analysis result was subjected to subsequent gene set analyses based on Gene Ontology (GO terms using gene set analysis software, GSA-SNP, identifying a set of GO terms significantly associated to each trait (pcorr < 0.05. Pairwise comparison of the traits in terms of the semantic similarity in their GO sets revealed surprising cases where phenotypically uncorrelated traits showed high similarity in terms of biological pathways. For example, the pH level was related to 7 other traits that showed low phenotypic correlations with it. A literature survey implies that these traits may be regulated partly by common pathways that involve neuronal or nerve systems.

Safety climate and attitude as evaluation measures of organizational safety.

Science.gov (United States)

Isla Díaz, R; Díaz Cabrera, D

1997-09-01

The main aim of this research is to develop a set of evaluation measures for safety attitudes and safety climate. Specifically it is intended: (a) to test the instruments; (b) to identify the essential dimensions of the safety climate in the airport ground handling companies; (c) to assess the quality of the differences in the safety climate for each company and its relation to the accident rate; (d) to analyse the relationship between attitudes and safety climate; and (e) to evaluate the influences of situational and personal factors on both safety climate and attitude. The study sample consisted of 166 subjects from three airport companies. Specifically, this research was centered on ground handling departments. The factor analysis of the safety climate instrument resulted in six factors which explained 69.8% of the total variance. We found significant differences in safety attitudes and climate in relation to type of enterprise.

Using Inequality Measures to Incorporate Environmental Justice into Regulatory Analyses

Science.gov (United States)

Abstract: Formally evaluating how specific policy measures influence environmental justice is challenging, especially in the context of regulatory analyses in which quantitative comparisons are the norm. However, there is a large literature on developing and applying quantitative...

Demonstration of inherent safety features of HTGRs using the HTTR

International Nuclear Information System (INIS)

Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

2004-01-01

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) quantitatively as well as providing the core and plant transient data for validation of HTGR analysis codes for safety evaluation. The safety demonstration test are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase test simulating reactivity insertion events and coolant flow reduction events stared in FY 2002. Post-test analyses have been conducted to reproduced the test results by using the core and plant dynamics analysis code, ACCORD and Monte Carlo code, MVP. The analysis results agreed fairly well with the test results of a control rod withdrawal test simulating reactivity insertion, and gas circulators trip test simulating coolant flow reduction, at power levels of 50% and 30% of the rated power, respectively. It is shown that improvement of the ACCORD code by taking into consideration vertical and horizontal temperature distribution gives better analysis results in the control rod withdrawal test. The fist phase safety demonstration tests will continue until FY 2005, and the second phase tests are planned to be started in FY 2006. (author)

Patient participation in patient safety still missing: Patient safety experts' views.

Science.gov (United States)

Sahlström, Merja; Partanen, Pirjo; Rathert, Cheryl; Turunen, Hannele

2016-10-01

The aim of this study was to elicit patient safety experts' views of patient participation in promoting patient safety. Data were collected between September and December in 2014 via an electronic semi-structured questionnaire and interviews with Finnish patient safety experts (n = 21), then analysed using inductive content analysis. Patient safety experts regarded patients as having a crucial role in promoting patient safety. They generally deemed the level of patient safety as 'acceptable' in their organizations, but reported that patient participation in their own safety varied, and did not always meet national standards. Management of patient safety incidents differed between organizations. Experts also suggested that patient safety training should be increased in both basic and continuing education programmes for healthcare professionals. Patient participation in patient safety is still lacking in clinical practice and systematic actions are needed to create a safety culture in which patients are seen as equal partners in the promotion of high-quality and safe care. © 2016 John Wiley & Sons Australia, Ltd.

Predictive microbiology: Quantitative science delivering quantifiable benefits to the meat industry and other food industries.

Science.gov (United States)

McMeekin, T A

2007-09-01

Predictive microbiology is considered in the context of the conference theme "chance, innovation and challenge", together with the impact of quantitative approaches on food microbiology, generally. The contents of four prominent texts on predictive microbiology are analysed and the major contributions of two meat microbiologists, Drs. T.A. Roberts and C.O. Gill, to the early development of predictive microbiology are highlighted. These provide a segue into R&D trends in predictive microbiology, including the Refrigeration Index, an example of science-based, outcome-focussed food safety regulation. Rapid advances in technologies and systems for application of predictive models are indicated and measures to judge the impact of predictive microbiology are suggested in terms of research outputs and outcomes. The penultimate section considers the future of predictive microbiology and advances that will become possible when data on population responses are combined with data derived from physiological and molecular studies in a systems biology approach. Whilst the emphasis is on science and technology for food safety management, it is suggested that decreases in foodborne illness will also arise from minimising human error by changing the food safety culture.

What price safety. A probabilistic cost-benefit evaluaton of existing engineered safety features

International Nuclear Information System (INIS)

O'Donnell, E.P.

1978-01-01

The paper provides a method for performing quantitative cost-benefit evaluations for nuclear safety concerns involving accidents of low probability and potentially large consequences. It presents an application of the method to ECCS, containment, emergency power system and hydrogen recombiner system. This evaluation provides a valuable assessment of the relative cost effectiveness of these features in reducing accident risk. It also provides insight into the sensitivity of cost-benefit calculations to the manner in which safety features are sequantially added in design. (author)

Nurse safety outcomes: old problem, new solution - the differentiating roles of nurses' psychological capital and managerial support.

Science.gov (United States)

Brunetto, Yvonne; Xerri, Matthew; Farr-Wharton, Ben; Shacklock, Kate; Farr-Wharton, Rod; Trinchero, Elisabetta

2016-11-01

The aim of this study was to examine the impacts of nurses' psychological capital and managerial support, plus specific safety interventions (managerial safety priorities, safety training satisfaction), on nurses' in-role safety performance. Most hospitals in industrialized countries have adopted selective (often the least costly) aspects of safety, usually related to safety policies. However, patient safety remains a challenge in many countries. Research shows that training can be used to upskill employees in psychological capital, with statistically significant organizational and employee benefits, but this area is under-researched in nursing. Data were collected using a survey-based, self-report strategy. The emerging patterns of data were then compared with the findings of previous research. Quantitative survey data were collected during 2014 from 242 nurses working in six Australian hospitals. Two models were tested and analysed using covariance-based Structural Equation Modelling. Psychological capital and safety training satisfaction were important predictors of nurses' in-role safety performance and as predictors of nurses' perceptions of whether management implements what it espouses about safety ('managerial safety priorities'). Managerial support accounted for just under a third of psychological capital and together, psychological capital and managerial support, plus satisfaction with safety training, were important to nurses' perceptions of in-role safety performance. Organizations are likely to benefit from upskilling nurses and their managers to increase nurses' psychological capital and managerial support, which then will enhance nurses' satisfaction with training and in-role safety performance perceptions. © 2016 John Wiley & Sons Ltd.

Quantitative clinical radiobiology

International Nuclear Information System (INIS)

Bentzen, S.M.

1993-01-01

Based on a series of recent papers, a status is given of our current ability to quantify the radiobiology of human tumors and normal tissues. Progress has been made in the methods of analysis. This includes the introduction of 'direct' (maximum likelihood) analysis, incorporation of latent-time in the analyses, and statistical approaches to allow for the many factors of importance in predicting tumor-control probability of normal-tissue complications. Quantitative clinical radiobiology of normal tissues is reviewed with emphasis on fractionation sensitivity, repair kinetics, regeneration, latency, and the steepness of dose-response curves. In addition, combined modality treatment, functional endpoints, and the search for a correlation between the occurrence of different endpoints in the same individual are discussed. For tumors, quantitative analyses of fractionation sensitivity, repair kinetics, reoxygenation, and regeneration are reviewed. Other factors influencing local control are: Tumor volume, histopathologic differentiation and hemoglobin concentration. Also, the steepness of the dose-response curve for tumors is discussed. Radiobiological strategies for improving radiotherapy are discussed with emphasis on non-standard fractionation and individualization of treatment schedules. (orig.)

Probabilistic safety analysis for fire events for the NPP Isar 2

International Nuclear Information System (INIS)

Schmaltz, H.; Hristodulidis, A.

2007-01-01

The 'Probabilistic Safety Analysis for Fire Events' (Fire-PSA KKI2) for the NPP Isar 2 was performed in addition to the PSA for full power operation and considers all possible events which can be initiated due to a fire. The aim of the plant specific Fire-PSA was to perform a quantitative assessment of fire events during full power operation, which is state of the art. Based on simplistic assumptions referring to the fire induced failures, the influence of system- and component-failures on the frequency of the core damage states was analysed. The Fire-PSA considers events, which will result due to fire-induced failures of equipment on the one hand in a SCRAM and on the other hand in events, which will not have direct operational effects but because of the fire-induced failure of safety related installations the plant will be shut down as a precautionary measure. These events are considered because they may have a not negligible influence on the frequency of core damage states in case of failures during the plant shut down because of the reduced redundancy of safety related systems. (orig.)

Organic Tanks Safety Program: Advanced organic analysis FY 1996 progress report

International Nuclear Information System (INIS)

1996-09-01

Major focus during the first part of FY96 was to evaluate using organic functional group concentrations to screen for energetics. Fourier transform infrared and Raman spectroscopy would be useful screening tools for determining C-H and COO- organic content in tank wastes analyzed in a hot cell. These techniques would be used for identifying tanks of potential safety concern that may require further analysis. Samples from Tanks 241-C-106 and -C-204 were analyzed; the major organic in C-106 was B2EHPA and in C-204 was TBP. Analyses of simulated wastes were also performed for the Waste Aging Studies Task; organics formed as a result of degradation were identified, and the original starting components were monitored quantitatively. Sample analysis is not routine and required considerable methods adaptation and optimization. Several techniques have been evaluated for directly analyzing chelator and chelator fragments in tank wastes: matrix-assisted laser desorption/ionization time-of-flight mass spectrometry and liquid chromatography with ultraviolet detection using Cu complexation. Although not directly funded by the Tanks Safety Program, the success of these techniques have implications for both the Flammable Gas and Organic Tanks Safety Programs

Probabilistic safety assessment

International Nuclear Information System (INIS)

Hoertner, H.; Schuetz, B.

1982-09-01

For the purpose of assessing applicability and informativeness on risk-analysis methods in licencing procedures under atomic law, the choice of instruments for probabilistic analysis, the problems in and experience gained in their application, and the discussion of safety goals with respect to such instruments are of paramount significance. Naturally, such a complex field can only be dealt with step by step, making contribution relative to specific problems. The report on hand shows the essentials of a 'stocktaking' of systems relability studies in the licencing procedure under atomic law and of an American report (NUREG-0739) on 'Quantitative Safety Goals'. (orig.) [de

Safety re-assessment of AECL test and research reactors

International Nuclear Information System (INIS)

Winfield, D.J.

1990-01-01

Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

Does Employee Safety Matter for Patients Too? Employee Safety Climate and Patient Safety Culture in Health Care.

Science.gov (United States)

Mohr, David C; Eaton, Jennifer Lipkowitz; McPhaul, Kathleen M; Hodgson, Michael J

2015-04-22

We examined relationships between employee safety climate and patient safety culture. Because employee safety may be a precondition for the development of patient safety, we hypothesized that employee safety culture would be strongly and positively related to patient safety culture. An employee safety climate survey was administered in 2010 and assessed employees' views and experiences of safety for employees. The patient safety survey administered in 2011 assessed the safety culture for patients. We performed Pearson correlations and multiple regression analysis to examine the relationships between a composite measure of employee safety with subdimensions of patient safety culture. The regression models controlled for size, geographic characteristics, and teaching affiliation. Analyses were conducted at the group level using data from 132 medical centers. Higher employee safety climate composite scores were positively associated with all 9 patient safety culture measures examined. Standardized multivariate regression coefficients ranged from 0.44 to 0.64. Medical facilities where staff have more positive perceptions of health care workplace safety climate tended to have more positive assessments of patient safety culture. This suggests that patient safety culture and employee safety climate could be mutually reinforcing, such that investments and improvements in one domain positively impacts the other. Further research is needed to better understand the nexus between health care employee and patient safety to generalize and act upon findings.

Using US EPA’s Chemical Safety for Sustainability’s Comptox Chemistry Dashboard and Tools for Bioactivity, Chemical and Toxicokinetic Modeling Analyses (Course at 2017 ISES Annual Meeting)

Science.gov (United States)

Title: Using US EPA’s Chemical Safety for Sustainability’s Comptox Chemistry Dashboard and Tools for Bioactivity, Chemical and Toxicokinetic Modeling Analyses • Class format: half-day (4 hours) • Course leader(s): Barbara A. Wetmore and Antony J. Williams,...

What is the value and impact of quality and safety teams? A scoping review

Directory of Open Access Journals (Sweden)

Norris Jill M

2011-08-01

Full Text Available Abstract Background The purpose of this study was to conduct a scoping review of the literature about the establishment and impact of quality and safety team initiatives in acute care. Methods Studies were identified through electronic searches of Medline, Embase, CINAHL, PsycINFO, ABI Inform, Cochrane databases. Grey literature and bibliographies were also searched. Qualitative or quantitative studies that occurred in acute care, describing how quality and safety teams were established or implemented, the impact of teams, or the barriers and/or facilitators of teams were included. Two reviewers independently extracted data on study design, sample, interventions, and outcomes. Quality assessment of full text articles was done independently by two reviewers. Studies were categorized according to dimensions of quality. Results Of 6,674 articles identified, 99 were included in the study. The heterogeneity of studies and results reported precluded quantitative data analyses. Findings revealed limited information about attributes of successful and unsuccessful team initiatives, barriers and facilitators to team initiatives, unique or combined contribution of selected interventions, or how to effectively establish these teams. Conclusions Not unlike systematic reviews of quality improvement collaboratives, this broad review revealed that while teams reported a number of positive results, there are many methodological issues. This study is unique in utilizing traditional quality assessment and more novel methods of quality assessment and reporting of results (SQUIRE to appraise studies. Rigorous design, evaluation, and reporting of quality and safety team initiatives are required.

Safety and licensing analyses for the Fort St. Vrain HTGR

International Nuclear Information System (INIS)

Ball, S.J.; Conklin, J.C.; Harrington, R.M.; Cleveland, J.C.; Clapp, N.E. Jr.

1982-01-01

The Oak Ridge National Laboratory (ORNL) safety analysis program for the HTGR includes development and verification of system response simulation codes, and applications of these codes to specific Fort St. Vrain reactor licensing problems. Licensing studies addressed the oscillation problems and the concerns about large thermal stresses in the core support blocks during a postulated accident

PWR plant transient analyses using TRAC-PF1

International Nuclear Information System (INIS)

Ireland, J.R.; Boyack, B.E.

1984-01-01

This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

Method of safety evaluation in nuclear power plants

International Nuclear Information System (INIS)

Kuraszkiewicz, P.; Zahn, P.

1988-01-01

A novel quantitative technique for evaluating safety of subsystems of nuclear power plants based on expert estimations is presented. It includes methods of mathematical psychology recognizing the effect of subjective factors in the expert estimates and, consequently, contributes to further objectification of evaluation. It may be applied to complementing probabilistic safety assessment. As a result of such evaluations a characteristic 'safety of nuclear power plants' is obtained. (author)

Review of accident analyses of RB experimental reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-01-01

The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

Review of accident analyses of RB experimental reactor

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2003-01-01

Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

The role of risk assessment and safety analysis in integrated safety assessments

International Nuclear Information System (INIS)

Niall, R.; Hunt, M.; Wierman, T.E.

1990-01-01

To ensure that the design and operation of both nuclear and non- nuclear hazardous facilities is acceptable, and meets all societal safety expectations, a rigorous deterministic and probabilistic assessment is necessary. An approach is introduced, founded on the concept of an ''Integrated Safety Assessment.'' It merges the commonly performed safety and risk analyses and uses them in concert to provide decision makers with the necessary depth of understanding to achieve ''adequacy.'' 3 refs., 1 fig

Quantitative analyses at baseline and interim PET evaluation for response assessment and outcome definition in patients with malignant pleural mesothelioma

Energy Technology Data Exchange (ETDEWEB)

Lopci, Egesta; Chiti, Arturo [Humanitas Research Hospital, Nuclear Medicine Department, Rozzano, Milan (Italy); Zucali, Paolo Andrea; Perrino, Matteo; Gianoncelli, Letizia; Lorenzi, Elena; Gemelli, Maria; Santoro, Armando [Humanitas Research Hospital, Oncology, Rozzano (Italy); Ceresoli, Giovanni Luca [Humanitas Gavazzeni, Oncology, Bergamo (Italy); Giordano, Laura [Humanitas Research Hospital, Biostatistics, Rozzano (Italy)

2015-04-01

Quantitative analyses on FDG PET for response assessment are increasingly used in clinical studies, particularly with respect to tumours in which radiological assessment is challenging and complete metabolic response is rarely achieved after treatment. A typical example is malignant pleural mesothelioma (MPM), an aggressive tumour originating from mesothelial cells of the pleura. We present our results concerning the use of semiquantitative and quantitative parameters, evaluated at the baseline and interim PET examinations, for the prediction of treatment response and disease outcome in patients with MPM. We retrospectively analysed data derived from 131 patients (88 men, 43 women; mean age 66 years) with MPM who were referred to our institution for treatment between May 2004 and July 2013. Patients were investigated using FDG PET at baseline and after two cycles of pemetrexed-based chemotherapy. Responses were determined using modified RECIST criteria based on the best CT response after treatment. Disease control rate, progression-free survival (PFS) and overall survival (OS) were calculated for the whole population and were correlated with semiquantitative and quantitative parameters evaluated at the baseline and interim PET examinations; these included SUV{sub max}, total lesion glycolysis (TLG), percentage change in SUV{sub max} (ΔSUV{sub max}) and percentage change in TLG (ΔTLG). Disease control was achieved in 84.7 % of the patients, and median PFS and OS for the entire cohort were 7.2 and 14.3 months, respectively. The log-rank test showed a statistically significant difference in PFS between patients with radiological progression and those with partial response (PR) or stable disease (SD) (1.8 vs. 8.6 months, p < 0.001). Baseline SUV{sub max} and TLG showed a statistically significant correlation with PFS and OS (p < 0.001). In the entire population, both ΔSUV{sub max} and ΔTLG were correlated with disease control based on best CT response (p < 0

RB research reactor safety report

International Nuclear Information System (INIS)

Sotic, O.; Pesic, M.; Vranic, S.

1979-04-01

This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

Accuracy of quantitative visual soil assessment

Science.gov (United States)

van Leeuwen, Maricke; Heuvelink, Gerard; Stoorvogel, Jetse; Wallinga, Jakob; de Boer, Imke; van Dam, Jos; van Essen, Everhard; Moolenaar, Simon; Verhoeven, Frank; Stoof, Cathelijne

2016-04-01

farmers carried out quantitative visual observations all independently from each other. All observers assessed five sites, having a sand, peat or clay soil. For almost all quantitative visual observations the spread of observed values was low (coefficient of variation background of the observer, might influence the outcome of visual assessment of some soil properties. In countries where soil analyses can easily be carried out, VSA might be a good replenishment to available soil chemical analyses, and in countries where it is not feasible to carry out soil analyses, VSA might be a good start to assess soil quality.

Safety goals for commercial nuclear power plants

International Nuclear Information System (INIS)

Roe, J.W.

1988-01-01

In its official policy statement on safety goals for the operation of nuclear power plants, the Nuclear Regulatory Commission (NRC) set two qualitative goals, supported by two quantitative objectives. These goals are that (1) individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health; and (2) societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks. As an alternative, this study proposes four quantitative safety goals for nuclear power plants. It begins with an analysis of the NRC's safety-goal development process, a key portion of which was devoted to delineating criteria for evaluating goal-development methods. Based on this analysis, recommendations for revision of the NRC's basic benchmarks for goal development are proposed. Using the revised criteria, NRC safety goals are evaluated, and the alternative safety goals are proposed. To further support these recommendations, both the NRC's goals and the proposed goals are compared with the results of three major probabilistic risk assessment studies. Finally, the potential impact of these recommendations on nuclear safety is described

Safety case plan 2008

International Nuclear Information System (INIS)

2008-07-01

Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy) Posiva is preparing to submit the construction license application for a spent fuel repository by the end of the year 2012. The long-term safety section supporting the license application is based on a safety case, which, according to the internationally adopted definition, is a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. In 2005, Posiva presented a plan to prepare such a safety case. The present report provides a revised plan of the safety case contents mentioned above. The update of the safety case plan takes into account the recommendations made by the Radiation and Nuclear Safety Authority (STUK) about improving the focus and further developing the plan. Accordingly, particular attention is given to the quality management of the safety case work, the management of uncertainties and the scenario methodology. The quality management is based on the ISO 9001:2000 standard process thinking enhanced with special features arising from STUK's YVL Guides. The safety case production process is divided into four main sub-processes. The conceptualisation and methodology sub-process defines the framework for the assessment. The critical data handling and modelling sub-process links Posiva's main technical and scientific activities to the production of the safety case. The assessment sub-process analyses the consequences of the evolution of the disposal system in various scenarios, classified either as part of the expected evolution or as disruptive scenarios. The compliance and confidence sub-process is responsible for final evaluation of compliance of the assessment results with the regulatory criteria and the overall confidence in the safety case. As in the previous safety case plan, the safety case will be based on several reports, but

Nuclear safety as applied to space power reactor systems

International Nuclear Information System (INIS)

Cummings, G.E.

1987-01-01

Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety

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

Food and feed safety assessment: the importance of proper sampling.

Science.gov (United States)

Kuiper, Harry A; Paoletti, Claudia

2015-01-01

The general principles for safety and nutritional evaluation of foods and feed and the potential health risks associated with hazardous compounds are described as developed by the Food and Agriculture Organization (FAO) and the World Health Organization (WHO) and further elaborated in the European Union-funded project Safe Foods. We underline the crucial role of sampling in foods/feed safety assessment. High quality sampling should always be applied to ensure the use of adequate and representative samples as test materials for hazard identification, toxicological and nutritional characterization of identified hazards, as well as for estimating quantitative and reliable exposure levels of foods/feed or related compounds of concern for humans and animals. The importance of representative sampling is emphasized through examples of risk analyses in different areas of foods/feed production. The Theory of Sampling (TOS) is recognized as the only framework within which to ensure accuracy and precision of all sampling steps involved in the field-to-fork continuum, which is crucial to monitor foods and feed safety. Therefore, TOS must be integrated in the well-established FAO/WHO risk assessment approach in order to guarantee a transparent and correct frame for the risk assessment and decision making process.

Ferrocyanide safety program cyanide speciation studies FY 1993 annual report

International Nuclear Information System (INIS)

Bryan, S.A.; Pool, K.H.; Bryan, S.L.; Sell, R.L.; Thomas, L.M.P.

1993-09-01

This report summarizes Pacific Northwest Laboratory's (PNL) FY 1993 progress toward developing and implementing methods to identify and quantify cyanide species in ferrocyanide tank waste. Currently, there are 24 high-level waste storage tanks at the US Department of Energy's (DOE) Hanford Site that have been placed on a Ferrocyanide Tank Watchlist because they contain an estimated 1000 g-moles or more of precipitated ferrocyanide. This amount of ferrocyanide is of concern because the consequences of a potential explosion may exceed those reported previously in safety analyses. To bound the safety concern, methods are needed to definitively measure and quantitate the amount of ferrocyanides present within actual waste tanks to a lower limit of at least 0.1 wt % up to approximately 15 wt %. The target analyte concentration for cyanide in waste is approximately 0.1 to 15 wt % (as CN) in the original undiluted sample. After dissolution of the original sample and appropriate dilutions, the concentration range of interest in the analytical solutions can vary between 0.001 to 0.1 wt % (as CN)

Development of safety function assessment trees for pressurized heavy water reactor LP/SD operations

International Nuclear Information System (INIS)

Yang, Hui Chang; Chung, Chang Hyun; Kim, Ki Yong; Jee, Moon Hak; Sung, Chang Kyoung

2003-01-01

The objective of Configuration Risk Management Program(CRMP) is to maintain the safety level by assuring the defense-in-depth of nuclear power plant while the configurations are changed during plant operations, especially for the LP/SD. Such a safety purpose can be achieved by establishing the risk monitoring programs with both quantitative and qualitative features. Generally, the quantitative risk evaluation models, i.e., PRA models are used for the risk evaluation during full power operation, and the qualitative risk evaluation models such as safety function assessment trees are used. Through this study, safety function assessment trees were developed

Development and application of an integrated evaluation framework for preventive safety applications

NARCIS (Netherlands)

Scholliers, J.; Joshi, S.; Gemou, M.; Hendriks, F.; Ljung Aust, M.; Luoma, J.; Netto, M.; Engstrom, J.; Leanderson Olsson, S.; Kutzner, R.; Tango, F.; Amditis, A.J.; Blosseville, J.M.; Bekiaris, E.

2011-01-01

Preventive safety functions help drivers avoid or mitigate accidents. No quantitative methods have been available to evaluate the safety impact of these systems. This paper describes a framework for the assessment of preventive and active safety functions, which integrates procedures for technical

Modeling Logistic Performance in Quantitative Microbial Risk Assessment

NARCIS (Netherlands)

Rijgersberg, H.; Tromp, S.O.; Jacxsens, L.; Uyttendaele, M.

2010-01-01

In quantitative microbial risk assessment (QMRA), food safety in the food chain is modeled and simulated. In general, prevalences, concentrations, and numbers of microorganisms in media are investigated in the different steps from farm to fork. The underlying rates and conditions (such as storage

Additional methodology development for statistical evaluation of reactor safety analyses

International Nuclear Information System (INIS)

Marshall, J.A.; Shore, R.W.; Chay, S.C.; Mazumdar, M.

1977-03-01

The project described is motivated by the desire for methods to quantify uncertainties and to identify conservatisms in nuclear power plant safety analysis. The report examines statistical methods useful for assessing the probability distribution of output response from complex nuclear computer codes, considers sensitivity analysis and several other topics, and also sets the path for using the developed methods for realistic assessment of the design basis accident

System theory and safety models in Swedish, UK, Dutch and Australian road safety strategies.

Science.gov (United States)

Hughes, B P; Anund, A; Falkmer, T

2015-01-01

Road safety strategies represent interventions on a complex social technical system level. An understanding of a theoretical basis and description is required for strategies to be structured and developed. Road safety strategies are described as systems, but have not been related to the theory, principles and basis by which systems have been developed and analysed. Recently, road safety strategies, which have been employed for many years in different countries, have moved to a 'vision zero', or 'safe system' style. The aim of this study was to analyse the successful Swedish, United Kingdom and Dutch road safety strategies against the older, and newer, Australian road safety strategies, with respect to their foundations in system theory and safety models. Analysis of the strategies against these foundations could indicate potential improvements. The content of four modern cases of road safety strategy was compared against each other, reviewed against scientific systems theory and reviewed against types of safety model. The strategies contained substantial similarities, but were different in terms of fundamental constructs and principles, with limited theoretical basis. The results indicate that the modern strategies do not include essential aspects of systems theory that describe relationships and interdependencies between key components. The description of these strategies as systems is therefore not well founded and deserves further development. Copyright © 2014 Elsevier Ltd. All rights reserved.

Heat transfer calculations for the High Flux Isotope Reactor (HFIR). Technical specifications: bases for safety limits and limiting safety system settings

International Nuclear Information System (INIS)

Sims, T.M.; Swanks, J.H.

1977-09-01

Heat transfer analyses, in support of the preparation of the HFIR technical specifications, were made to establish the bases for the safety limits and limiting safety system settings applicable to the HFIR. The results of these analyses, along with the detailed bases, are presented

Semi-quantitative prediction of a multiple API solid dosage form with a combination of vibrational spectroscopy methods.

Science.gov (United States)

Hertrampf, A; Sousa, R M; Menezes, J C; Herdling, T

2016-05-30

Quality control (QC) in the pharmaceutical industry is a key activity in ensuring medicines have the required quality, safety and efficacy for their intended use. QC departments at pharmaceutical companies are responsible for all release testing of final products but also all incoming raw materials. Near-infrared spectroscopy (NIRS) and Raman spectroscopy are important techniques for fast and accurate identification and qualification of pharmaceutical samples. Tablets containing two different active pharmaceutical ingredients (API) [bisoprolol, hydrochlorothiazide] in different commercially available dosages were analysed using Raman- and NIR Spectroscopy. The goal was to define multivariate models based on each vibrational spectroscopy to discriminate between different dosages (identity) and predict their dosage (semi-quantitative). Furthermore the combination of spectroscopic techniques was investigated. Therefore, two different multiblock techniques based on PLS have been applied: multiblock PLS (MB-PLS) and sequential-orthogonalised PLS (SO-PLS). NIRS showed better results compared to Raman spectroscopy for both identification and quantitation. The multiblock techniques investigated showed that each spectroscopy contains information not present or captured with the other spectroscopic technique, thus demonstrating that there is a potential benefit in their combined use for both identification and quantitation purposes. Copyright © 2016 Elsevier B.V. All rights reserved.

Inherent and passive safety measures in accelerator driven systems: a safety strategy for ADS

International Nuclear Information System (INIS)

Maschek, W.; Rineiski, A.; Morita, K.; Flad, M.

2001-01-01

The efficiency of Accelerator Driven Systems (ADSs) for the transmutation and incineration of nuclear waste is strongly related to the utilization of so-called dedicated fuels. In the ideal case these fuels should consist of pure TRUs without fertile materials as 238 U or 232 Th to achieve highest incineration/transmutation rates. Dedicated fuels still have to be developed and programs are under way for their fabrication, irradiation and testing. These fertile-free fuels may suffer from deteriorated thermal or thermo-mechanical properties, as a reduced melting point, reduced thermal conductivity or even thermal instability. First analyses have shown that the use of dedicated fuels may lead to a strong deterioration of the safety parameters of the reactor core as e.g. the void worth, the Doppler or the kinetics quantities as neutron generation time and β eff . In addition, a dedicated core may contain multiple ''critical'' fuel masses, resulting in a considerable recriticality potential. Current knowledge on these dedicated fuels suggests that ''critical'' reactors may not be feasible, because of safety reasons. However, for ADSs, the salient hope has been promoted that due to the subcriticality of the system the poor safety features of such fuels could be coped with. Analyses are presented which show potential safety problems for such dedicated cores. Respecting the results of these analyses a safety strategy is proposed along the lines of defense approach in analogy with ideas formerly developed for fast reactors. Inherent and passive safety measures are integrated into the various defense lines. (author)

Problems with quantification of safety culture

International Nuclear Information System (INIS)

Kozuh, M.; Mavko, B.

1995-01-01

For the qualitative part of the method for the Safety Culture assessment we quantitative part was developed based on expert judgement and statistical methods. The quantitative assessment should go in parallel with the qualitative part already presented. The essential part is based on expert opinion which organizational factors are the most important for certain risk significant components and how well are they implemented. The problems with getting the ratings are described in the paper. (author)

Elastic fibers in human skin: quantitation of elastic fibers by computerized digital image analyses and determination of elastin by radioimmunoassay of desmosine.

Science.gov (United States)

Uitto, J; Paul, J L; Brockley, K; Pearce, R H; Clark, J G

1983-10-01

The elastic fibers in the skin and other organs can be affected in several disease processes. In this study, we have developed morphometric techniques that allow accurate quantitation of the elastic fibers in punch biopsy specimens of skin. In this procedure, the elastic fibers, visualized by elastin-specific stains, are examined through a camera unit attached to the microscope. The black and white images sensing various gray levels are then converted to binary images after selecting a threshold with an analog threshold selection device. The binary images are digitized and the data analyzed by a computer program designed to express the properties of the image, thus allowing determination of the volume fraction occupied by the elastic fibers. As an independent measure of the elastic fibers, alternate tissue sections were used for assay of desmosine, an elastin-specific cross-link compound, by a radioimmunoassay. The clinical applicability of the computerized morphometric analyses was tested by examining the elastic fibers in the skin of five patients with pseudoxanthoma elasticum or Buschke-Ollendorff syndrome. In the skin of 10 healthy control subjects, the elastic fibers occupied 2.1 +/- 1.1% (mean +/- SD) of the dermis. The volume fractions occupied by the elastic fibers in the lesions of pseudoxanthoma elasticum or Buschke-Ollendorff syndrome were increased as much as 6-fold, whereas the values in the unaffected areas of the skin in the same patients were within normal limits. A significant correlation between the volume fraction of elastic fibers, determined by computerized morphometric analyses, and the concentration of desmosine, quantitated by radioimmunoassay, was noted in the total material. These results demonstrate that computerized morphometric techniques are helpful in characterizing disease processes affecting skin. This methodology should also be applicable to other tissues that contain elastic fibers and that are affected in various heritable and

Application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal. Final report; Anwendung und Weiterentwicklung von Modellen fuer Endlagersicherheitsanalysen auf die Freigabe radioaktiver Stoffe zur Deponierung. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Artmann, Andreas; Larue, Juergen; Seher, Holger; Weiss, Dietmar

2014-08-15

The project of application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal is aimed to study the long-term safety using repository-specific simulation programs with respect to radiation exposure for different scenarios. It was supposed to investigate whether the 10 micro Sv criterion can be guaranteed under consideration of human intrusion scenarios. The report covers the following issues: selection and identification of models and codes and the definition of boundary conditions; applicability of conventional repository models for long-term safety analyses; modeling results for the pollutant release and transport and calculation of radiation exposure; determination of the radiation exposure.

Development of safety analysis technology for LMR

International Nuclear Information System (INIS)

Hahn, Do Hee; Kwon, Y. M.; Kim, K. D.

2000-05-01

The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well

Development of safety analysis technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

2000-05-01

The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

Uncertainty and conservatism in safety evaluations based on a BEPU approach

International Nuclear Information System (INIS)

Yamaguchi, A.; Mizokami, S.; Kudo, Y.; Hotta, A.

2009-01-01

Atomic Energy Society of Japan has published 'Standard Method for Safety Evaluation using Best Estimate Code Based on Uncertainty and Scaling Analyses with Statistical Approach' to be applied to accidents and AOOs in the safety evaluation of LWRs. In this method, hereafter named as the AESJ-SSE (Statistical Safety Evaluation) method, identification and quantification of uncertainties will be performed and then a combination of the best estimate code and the evaluation of uncertainty propagation will be performed. Uncertainties are categorized into bias and variability. In general, bias is related to our state-of-knowledge on uncertainty objects (modeling, scaling, input data, etc.) while variability reflects stochastic features involved in these objects. Considering many kinds of uncertainties in thermal-hydraulics models and experimental databases show variabilities that will be strongly influenced by our state of knowledge, it seems reasonable that these variabilities are also related to state-of-knowledge. The design basis events (DBEs) that are employed for licensing analyses form a main part of the given or prior conservatism. The regulatory acceptance criterion is also regarded as the prior conservatism. In addition to these prior conservatisms, a certain amount of the posterior conservatism is added with maintaining intimate relationships with state-of-knowledge. In the AESJ-SSE method, this posterior conservatism can be incorporated into the safety evaluation in a combination of the following three ways, (1) broadening ranges of variability relevant to uncertainty objects, (2) employing more disadvantageous biases relevant to uncertainty objects and (3) adding an extra bias to the safety evaluation results. Knowing implemented quantitative bases of uncertainties and conservatism, the AESJ-SSE method provides a useful ground for rational decision-making. In order to seek for 'the best estimation' as well as reasonably setting the analytical margin, a degree

PWR reload safety evaluation methodology

International Nuclear Information System (INIS)

Doshi, P.K.; Chapin, D.L.; Love, D.S.

1993-01-01

The current practice for WWER safety analysis is to prepare the plant Safety Analysis Report (SAR) for initial plant operation. However, the existing safety analysis is typically not evaluated for reload cycles to confirm that all safety limits are met. In addition, there is no systematic reanalysis or reevaluation of the safety analyses after there have been changes made to the plant. The Westinghouse process is discussed which is in contrast to this and in which the SAR conclusions are re-validated through evaluation and/or analysis of each reload cycle. (Z.S.)

Safety, safety case and society - Lessons from the experience of the Forum on Stakeholder Confidence and other NEA initiatives

International Nuclear Information System (INIS)

Pescatore, Claudio

2014-01-01

A vast amount of literature on radioactive waste management (RWM) and its governance is available on the web page of the Radioactive Waste Management Committee of the OECD Nuclear Energy Agency (NEA), in particular on the pages of the Forum on Stakeholder Confidence (FSC), the Reversibility and Retrievability (R and R) Project and the Project on Records, Knowledge and Memory (RK and M) Preservation across Generations. The FSC literature alone likely represents the largest collection of literature on RWM governance presently available on any single site. The safety case developed for any deep geological repository project deals with technical safety. A license is to be granted based on the repository being, after closure, safe 'by itself', i.e. without the need to watch it, independent of the existence of the implementer, regulator and others. The main legal requirement of the safety case is that it needs to show convincingly that the technical regulatory criteria are met. The latter are both qualitative and quantitative. Qualitative criteria are technical, but not in a strong sense, e.g. one requirement may simply be the use of 'sound technical and managerial principles'. The safety case also needs to argue robustness upon human intrusion. The human intrusion analyses, however, are only used to make a qualitative judgement on the robustness of the system. The international guidance suggests that their results need not be tested, by the authorities, for compliance against a numerical yardstick. The technical regulator will have an important role in decision making, but others aside from the technical regulator will also play a decision-making role in the development of a repository project and with regard to its safety. For instance, the technical regulator is largely removed from the initial choice of site. Safety nowadays is brought about by a system of actors comprising the implementer, technical regulators, specialist groups in various advisory roles and the

Risk, fear and public safety

International Nuclear Information System (INIS)

Siddall, E.

1981-04-01

Part 1 of the paper advocates a rational approach to public safety based on unbiassed quantitative assessment of overall risks and benefits of any technological activity. It shows that improved safety should be attainable at less cost than is the case at present. Part 2 offers an explanation of why so little has been achieved in this direction and outlines the major errors in present practices. Part 3 suggests what might realistically be done towards the achievement of some of the possible benefits. Factors which are important in the study of safety and evidence supporting the arguments are discussed in six appendices. It is urged that the scientific and technological community should improve its understanding of safety as a specialization and should endeavour to lead rather than follow in our present political system

To dimension safety valves. Probabilist study

International Nuclear Information System (INIS)

Noel, Robert; Couvreur, Denis

1982-01-01

The gauge of safety valves of a steam pressure apparatus is usually determined according to an operating situation envelope which it is admitted covers all that can happen in reality. For the safety of the dryer-superheaters of turbines in nuclear power stations, Electricite de France and Alsthom-Atlantique made a reliability study; its method is exposed and the results are discussed. Such a study is heavy going and complex, but in return it permits a better quantitative understanding of the various dimension and operating parameters of an installation which condition its safety. It is therefore a source of progress [fr

Calculational framework for safety analyses of non-reactor nuclear facilities

International Nuclear Information System (INIS)

Coleman, J.R.

1994-01-01

A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks

Everyday life and social relations in home-living patients with mild Alzheimer's disease and their caregivers: Quantitative and qualitative analyses

DEFF Research Database (Denmark)

Sørensen, Lisbeth Villemoes

2007-01-01

Everyday life and social relations in home-living patients with mild Alzheimer’s disease (AD) and their caregivers: quantitative and qualitative analyses. This PhD project was carried out between April 2004 and March 2007 during my employment as project coordinator in the Memory Disorder Research...... and the presence of neuropsychiatric symptoms. In the second study, data were collected using semi-structured research interviews with 11 patients before their participation in the DAISY intervention programme. Grounded theory analysis of the interview data revealed that the basic social psychological problem...... faced by the patients was: their awareness of decline in personal dignity and value. Coping strategies used to meet these problems were adaptations to the altered situation in order to maintain a feeling of well-being. In the third study, data were collected using individual semi-structured research...

Impact of a health safety warning and prior authorisation on the use of piroxicam: a time-series study.

Science.gov (United States)

Carracedo-Martínez, Eduardo; Pia-Morandeira, Agustin; Figueiras, Adolfo

2012-03-01

The aim of this study was to assess the quantitative changes in systemic use of piroxicam after the issue of a health safety warning about its risks and the subsequent implementation of prior authorisation. We determined the number of monthly daily defined doses/1000 inhabitants/day (DHDs) of piroxicam in the period 2005-2008 in a health area in Spain. The data were analysed graphically, and the impact of the safety warning and introduction of prior authorisation were estimated by using segmented regression analysis. The graph showed that the number of DHDs of piroxicam was stable both before and after the health safety warning but registered a very marked decrease after implementation of prior authorisation, after which DHDs of piroxicam remained stable at a 98% inferior level compared with previous to prior authorisation. Segmented regression analysis showed no statistically significant immediate jump in piroxicam utilisation after the safety warning nor a change in the slope afterwards, but it did show a significant immediate jump after prior authorisation. Population exposure to systemic piroxicam remained unaffected by a previous health safety warning but declined sharply after the introduction of prior authorisation. Copyright © 2012 John Wiley & Sons, Ltd.

Quantitative risk analysis of urban flooding in lowland areas

NARCIS (Netherlands)

Ten Veldhuis, J.A.E.

2010-01-01

Urban flood risk analyses suffer from a lack of quantitative historical data on flooding incidents. Data collection takes place on an ad hoc basis and is usually restricted to severe events. The resulting data deficiency renders quantitative assessment of urban flood risks uncertain. The study

A quantitative assessment of organizational factors affecting safety using a system dynamics model

International Nuclear Information System (INIS)

Yoo, J. K.; Yoon, T. S.

2003-01-01

The purpose of this study is to develop a system dynamics model for the assessment of organizational and human factors in the nuclear power plant safety. Previous studies are classified into two major approaches. One is the engineering approach such as ergonomics and Probabilistic Safety Assessment (PSA). The other is socio-psychology one. Both have contributed to find organizational and human factors and increased nuclear safety However, since these approaches assume that the relationship among factors is independent they do not explain the interactions between factors or variables in NPP's. To overcome these restrictions, a system dynamics model, which can show causal relations between factors and quantify organizational and human factors, has been developed. Operating variables such as degree of leadership, adjustment of number of employee, and workload in each department, users can simulate various situations in nuclear power plants in the organization side. Through simulation, user can get an insight to improve safety in plants and to find managerial tools in the organization and human side

Risk analyses of nuclear power plants

International Nuclear Information System (INIS)

Jehee, J.N.T.; Seebregts, A.J.

1991-02-01

Probabilistic risk analyses of nuclear power plants are carried out by systematically analyzing the possible consequences of a broad spectrum of causes of accidents. The risk can be expressed in the probabilities for melt down, radioactive releases, or harmful effects for the environment. Following risk policies for chemical installations as expressed in the mandatory nature of External Safety Reports (EVRs) or, e.g., the publication ''How to deal with risks'', probabilistic risk analyses are required for nuclear power plants

Quantitative nature of overexpression experiments

Science.gov (United States)

Moriya, Hisao

2015-01-01

Overexpression experiments are sometimes considered as qualitative experiments designed to identify novel proteins and study their function. However, in order to draw conclusions regarding protein overexpression through association analyses using large-scale biological data sets, we need to recognize the quantitative nature of overexpression experiments. Here I discuss the quantitative features of two different types of overexpression experiment: absolute and relative. I also introduce the four primary mechanisms involved in growth defects caused by protein overexpression: resource overload, stoichiometric imbalance, promiscuous interactions, and pathway modulation associated with the degree of overexpression. PMID:26543202

Object-oriented fault tree evaluation program for quantitative analyses

Science.gov (United States)

Patterson-Hine, F. A.; Koen, B. V.

1988-01-01

Object-oriented programming can be combined with fault free techniques to give a significantly improved environment for evaluating the safety and reliability of large complex systems for space missions. Deep knowledge about system components and interactions, available from reliability studies and other sources, can be described using objects that make up a knowledge base. This knowledge base can be interrogated throughout the design process, during system testing, and during operation, and can be easily modified to reflect design changes in order to maintain a consistent information source. An object-oriented environment for reliability assessment has been developed on a Texas Instrument (TI) Explorer LISP workstation. The program, which directly evaluates system fault trees, utilizes the object-oriented extension to LISP called Flavors that is available on the Explorer. The object representation of a fault tree facilitates the storage and retrieval of information associated with each event in the tree, including tree structural information and intermediate results obtained during the tree reduction process. Reliability data associated with each basic event are stored in the fault tree objects. The object-oriented environment on the Explorer also includes a graphical tree editor which was modified to display and edit the fault trees.

Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

2015-05-15

IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea.

Development of a Novel Nuclear Safety Culture Evaluation Method for an Operating Team Using Probabilistic Safety Analysis

International Nuclear Information System (INIS)

Han, Sangmin; Lee, Seung Min; Seong, Poong Hyun

2015-01-01

IAEA defined safety culture as follows: 'Safety Culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. Also, celebrated behavioral scientist, Cooper, defined safety culture as,'safety culture is that observable degree of effort by which all organizational members direct their attention and actions toward improving safety on a daily basis' with his internal psychological, situational, and behavioral context model. With these various definitions and criteria of safety culture, several safety culture assessment methods have been developed to improve and manage safety culture. To develop a new quantitative safety culture evaluation method for an operating team, we unified and redefined safety culture assessment items. Then we modeled a new safety culture evaluation by adopting level 1 PSA concept. Finally, we suggested the criteria to obtain nominal success probabilities of assessment items by using 'operational definition'. To validate the suggested evaluation method, we analyzed the collected audio-visual recording data collected from a full scope main control room simulator of a NPP in Korea

New developments in quantitative risk assessment of campylobacteriosis

DEFF Research Database (Denmark)

Havelaar, Arie; Nauta, Maarten

meat to ready-to-eat foods is the main pathway of consumer exposure. Undercooking appears to be of minor importance. However, this conclusion may need to be reconsidered in the light of increasing consumption of minced meat preparations. Five QMRA models have been compared in detail, and detailed......Quantitative microbiological risk assessment (QMRA) is now broadly accepted as an important decision support tool in food safety risk management. It has been used to support decision making at the global level (Codex Alimentarius, FAO and WHO), at the European level (European Food Safety Authority...

Experiences in assessing safety culture

International Nuclear Information System (INIS)

Spitalnik, J.

2002-01-01

Based on several Safety Culture self-assessment applications in nuclear organisations, the paper stresses relevant aspects to be considered when programming an assessment of this type. Reasons for assessing Safety Culture, basic principles to take into account, necessary resources, the importance of proper statistical analyses, the feed-back of results, and the setting up of action plans to enhance Safety Culture are discussed. (author)

Quantitative approaches in climate change ecology

DEFF Research Database (Denmark)

Brown, Christopher J.; Schoeman, David S.; Sydeman, William J.

2011-01-01

Contemporary impacts of anthropogenic climate change on ecosystems are increasingly being recognized. Documenting the extent of these impacts requires quantitative tools for analyses of ecological observations to distinguish climate impacts in noisy data and to understand interactions between...... climate variability and other drivers of change. To assist the development of reliable statistical approaches, we review the marine climate change literature and provide suggestions for quantitative approaches in climate change ecology. We compiled 267 peer‐reviewed articles that examined relationships...

MELCOR 1.8.2 Analyses in Support of ITER's RPrS

International Nuclear Information System (INIS)

Brad J Merrill

2008-01-01

The International Thermonuclear Experimental Reactor (ITER) Program is performing accident analyses for ITER's 'Rapport Preliminaire de Surete' (Report Preliminary on Safety - RPrS) with a modified version of the MELCOR 1.8.2 code. The RPrS is an ITER safety document required in the ITER licensing process to obtain a 'Decret Autorisation de Construction' (a Decree Authorizing Construction - DAC) for the ITER device. This report documents the accident analyses performed by the US with the MELCOR 1.8.2 code in support of the ITER RPrS effort. This work was funded through an ITER Task Agreement for MELCOR Quality Assurance and Safety Analyses. Under this agreement, the US was tasked with performing analyses for three accident scenarios in the ITER facility. Contained within the text of this report are discussions that identify the cause of these accidents, descriptions of how these accidents are likely to proceed, the method used to analyze the consequences of these accidents, and discussions of the transient thermal hydraulic and radiological release results for these accidents

Probabilistic safety analyses. Status and further development of methods and models, applications

International Nuclear Information System (INIS)

Berg, H.P.; Schott, H.

1992-12-01

The report describes the topics of the deterministic and probabilistic approach. The PSA is used in order to investigate event sequences beyond design limits; in particular the expected frequency of core melting is important. The basis of PSA is described including its limits. Moreover, the current state of the art of science and technology in the field of PSA including the so-called 'living PSA' are explained. Some measures which result in order to improve the safety of a nuclear power plant from the German Risk-Study are shown. An overview is given on the status of PSA in periodic safety reviews in German nuclear power plants. Moreover, the main topics of running investigations are presented. (orig.) [de

The use of safety indicators in the assessment of radioactive waste disposal

International Nuclear Information System (INIS)

Wingefors, S.; Westerlind, M.; Gera, F.

1999-01-01

The most widely used criteria for disposal are limits or constraints on individual dose or risk, and these have been introduced in most national legal frameworks. There is general agreement that future generations have the right to the same level of protection as the current generation. Even if quantitative criteria corresponding to the required level of protection can be (and have been) defined, it is a great challenge to demonstrate compliance with these criteria. The difficulties are to large extent due to the long time-scales needed to be considered in radioactive waste disposal. The future cannot be predicted in detail but instead different scenarios, with different probabilities of occurrence, must be assessed. Some parts of a disposal system can be predicted or analysed with high confidence for very long periods of time, e.g. geological formations, while for example the evolution of the biosphere, and in particular the society, become quite uncertain within less than one thousand years. Thus, there may be considerable uncertainty in doses (or risks) derived from the safety assessment of a repository. Due to these unavoidable uncertainties it is believed advantageous to use multiple approaches in the safety assessment and to identify different indicators for the repository safety ('multiple-lines-of-reasoning'). The most fundamental safety indicators are dose/risk but complementary indicators have been suggested, in particular flux and environmental concentration of radionuclides. This presentation is focussed on fluxes and concentrations as complementary safety indicators. Other safety indicators, e.g. transfer times, are mentioned only briefly

Independent assessment for new nuclear reactor safety

Directory of Open Access Journals (Sweden)

D'Auria Francesco

2017-01-01

Full Text Available A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs. Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry. The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach.

Independent assessment for new nuclear reactor safety

International Nuclear Information System (INIS)

D'Auria, F.; Glaeser, H.; Debrecin, N.

2017-01-01

A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty (BEPU) approach. (authors)

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

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF 6 ; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

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

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

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

International Nuclear Information System (INIS)

2010-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

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

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation.? read more The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are

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

International Nuclear Information System (INIS)

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

2008-01-01

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

A probabilistic safety assessment of in-pile test loop in HWRR

International Nuclear Information System (INIS)

Cao Xuewu; Li Zhaohuan

1991-07-01

The PSA methodology has been applied to the in-pile test loop which is installed in the Heavy Water Research Reactor (HWRR). This loop is designed and operated for fuel assembly testing of the Qinshan PWR plant. This analysis is to assess the safety and to evaluate the design of this operating loop. The procedure and models are similar to a PSA on nuclear power plant. The major contents in the analysis consist of the familiarization of the object, the investigation and selection of accident initiators, setting events and fault trees, data collections, quantitative calculations, qualitative and result analyses and final conclusion. This analysis is only limited to the initiators of in-pile loop itself and possible errors made by operators during normal operation. The accident occurence is less than 10 -4 a -1 which may be recommended as an acceptance risk for safety operation of an in-pile test loop. Finally, suggestions have been raised to improve the design of test loop, especially in reducing operation errors by local operators

A bottom-up approach in estimating the measurement uncertainty and other important considerations for quantitative analyses in drug testing for horses.