Human Factors Evaluation of Man-Machine Interface for Periodic Safety Review of Yonggwang Unit no. 1, 2

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Hee; Lee, Jung Woon; Park, Jae Chang (and others)

2006-01-15

This report describes the research results of human factors assessment on the MMI(Man Machine Interface) equipment as part of Periodic Safety Review(PSR) of Yonggwang Unit no. 1, 2. As MMI is a key factor among human factors to be reviewed in PSR, we reviewed the MMI components of nuclear power plants in aspect of human factors engineering. The availability, suitability, and effectiveness of the MMI devices were chosen to be reviewed. The MMI devices were investigated through the review of design documents related to the MMI, survey of control panels, evaluation of experts, and experimental assessment. Checklists were used to perform this assessment and record the review results. The items mentioned by the expert comments to review in detail in relation with task procedures were tested by experiments with operators' participation. For some questionable issues arisen during this MMI review, operator workload and possibility of errors in operator actions were analysed. The reviewed MMI devices contain MCR(Main Control Room), SPDS(Safety Parameter Display System), RSP(Remote Shutdown Panel), and the selected LCBs(Local Control Boards) importantly related to safety. As results of the assessments, any significant problem challenging the safety was not found on human factors in the MMI devices. However, several small items to be changed and improved in suitability of MMI devices were discovered. An action plan is recommended to accommodate the suggestions and review comments. It will enhance the plant safety on MMI area.

Proposal of a risk-factor-based analytical approach for integrating occupational health and safety into project risk evaluation.

Science.gov (United States)

Badri, Adel; Nadeau, Sylvie; Gbodossou, André

2012-09-01

Excluding occupational health and safety (OHS) from project management is no longer acceptable. Numerous industrial accidents have exposed the ineffectiveness of conventional risk evaluation methods as well as negligence of risk factors having major impact on the health and safety of workers and nearby residents. Lack of reliable and complete evaluations from the beginning of a project generates bad decisions that could end up threatening the very existence of an organization. This article supports a systematic approach to the evaluation of OHS risks and proposes a new procedure based on the number of risk factors identified and their relative significance. A new concept called risk factor concentration along with weighting of risk factor categories as contributors to undesirable events are used in the analytical hierarchy process multi-criteria comparison model with Expert Choice(©) software. A case study is used to illustrate the various steps of the risk evaluation approach and the quick and simple integration of OHS at an early stage of a project. The approach allows continual reassessment of criteria over the course of the project or when new data are acquired. It was thus possible to differentiate the OHS risks from the risk of drop in quality in the case of the factory expansion project. Copyright © 2011 Elsevier Ltd. All rights reserved.

Problems of nuclear power plant safety evaluation

International Nuclear Information System (INIS)

Suchomel, J.

1977-01-01

Nuclear power plant safety is discussed with regard to external effects on the containment and to the human factor. As for external effects, attention is focused on shock waves which may be due to explosions or accidents in flammable material transport and storage, to missiles, and to earthquake effects. The criteria for evaluating nuclear power plant safety in different countries are shown. Factors are discussed affecting the reliability of man with regard to his behaviour in a loss-of-coolant accident in the power plant. Different types of PWR containments and their functions are analyzed, mainly in case of accident. Views are discussed on the role of destructive accidents in the overall evaluation of fast reactor safety. Experiences are summed up gained with the operation of WWER reactors with respect to the environmental impact of the nuclear power plants. (Z.M.)

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

International Nuclear Information System (INIS)

Ebisawa, K.; Abe, K.; Muramatsu, K.; Itoh, M.; Kohno, K.; Tanaka, T.

1994-01-01

This paper presents a method for evaluating 'response factors' of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design resonse to actual response. This method has the following characteristic features: (1) The components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components. This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups. (orig.)

Evaluation of design safety factors for time-dependent buckling

International Nuclear Information System (INIS)

Stone, C.M.; Nickell, R.E.

1977-02-01

The ASME Boiler and Pressure Vessel Code rules concerning time-dependent (creep) buckling for Class 1 nuclear components have recently been changed. Previous requirements for a factor of ten on service life have been replaced with a factor of safety of 1.5 on loading for load-controlled buckling. This report examines the supposed equivalence of the two rules from the standpoint of materials behavior--specifically, the secondary creep strain rate exponent. The comparison is made using results obtained numerically for an axially-loaded, cylindrical shell with varying secondary creep exponents. A computationally efficient scheme for analyzing creep buckling problems is also presented

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Human Factors and Safety Culture in Maritime Safety (revised

Directory of Open Access Journals (Sweden)

Heinz Peter Berg

2013-09-01

Full Text Available As in every industry at risk, the human and organizational factors constitute the main stakes for maritime safety. Furthermore, several events at sea have been used to develop appropriate risk models. The investigation on maritime accidents is, nowadays, a very important tool to identify the problems related to human factor and can support accident prevention and the improvement of maritime safety. Part of this investigation should in future also be near misses. Operation of ships is full of regulations, instructions and guidelines also addressing human factors and safety culture to enhance safety. However, even though the roots of a safety culture have been established, there are still serious barriers to the breakthrough of the safety management. One of the most common deficiencies in the case of maritime transport is the respective monitoring and documentation usually lacking of adequacy and excellence. Nonetheless, the maritime area can be exemplified from other industries where activities are ongoing to foster and enhance safety culture.

Development of the safety evaluation system in the respects of organizational factors and workers' consciousness. Pt. 4. Application of the system for contract companies

International Nuclear Information System (INIS)

Hasegawa, Naoko; Hirose, Ayako; Hayase, Kenichi; Tsuge, Tadafumi; Sasou, Kunihide; Takano, Kenichi

2003-01-01

The purpose of our study is to develop a safety evaluation system which clarifies the safety level of an organization. As a basic method of evaluation using a questionnaire had been established, now that the generalization is needed for the system. Hence, this paper is intended to consider the applicability of the system for contract companies. Subjects were workers who belonged to contract companies engaging in the maintenance of power plants in regular inspections. The following results were obtained: 1) The Comprehensive Safety Index (CSI) taking into account individual and organizational factors was identified using the principal component analysis. 2) The validity of CSI was confirmed with significant correlations between the CSI score and the rate of accidents. 3) Careful consideration should be provided for individual factors especially when evaluating the safety level of subcontract companies. 4) It seemed necessary to take into account the influence of parent companies and occupational hazards level. 5) The comparison among different industries should be avoided because of the difference in organizational structures and subjects of attention for keeping safety. (author)

Identification of road user related risk factors, Deliverable 5.1 of the H2020 project SafetyCube (Safety CaUsation, Benefits and Efficiency).

NARCIS (Netherlands)

Filtness, A. & Papadimitriou, E. (Eds.) Leskovšek, B. Focant, N. Martensen, H. Sgarra, V. Usami, D.S. Soteropoulos, A. Stadlbauer, S. Theofilatos, A. Yannis, G. Ziakopoulos, A. Diamandouros, K. Durso, C. Goldenbeld, C. Loenis, B. Schermers, G. Petegem, J.-H. van Elvik, R. Hesjevoll, I.S. Quigley, C. & Papazikou, E.

2017-01-01

The present Deliverable (D5.1) describes the identification and evaluation of infrastructure related risk factors. It outlines the results of Task 5.1 of WP5 of SafetyCube, which aimed to identify and evaluate infrastructure related risk factors and related road safety problems by (i) presenting a

A guideline for comprehensive evaluation of a licensee's effort to cultivate safety culture

International Nuclear Information System (INIS)

Makino, Maomi; Ishii, Yoichi

2009-01-01

The nuclear industry in Japan had held excellent performance in safety in the world during 90's. However recent events stem from organizational factors and defects of safety culture are pointed out in their contexts. In order to reduce accidents caused by organizational factors, the Japanese Regulatory body NISA (Nuclear and Industrial Safety Agency) decided to evaluate a licensee's effort for the cultivation of safety culture, and to order all licensses to add the provision of cultivating safety culture to their safety preservation rules. The inspection for the new safety preservation rules started in December, 2007. For a measure of evaluation by resident inspectors, NISA and the Japan Nuclear Energy Safety Organization (JNES) prepared a guideline for the prevention of degradation of safety culture and organizational climate. In this guideline, 14 items were defined as the components of the safety culture or as the viewpoints to evaluate the effort made to prevent any degradation of safety culture and organizational climate in the daily safety preservation activities. The 14 items are also used to establish the method to comprehensively evaluate the effort to prevent degradation of safety culture and organizational climate. This method consists of 10 steps: two steps to taken prior to start of the evaluation, two steps to be taken during the evaluation period, 5 steps to be taken during a comprehensive evaluation period and a final step to be taken for comprehensive findings for safety culture. This paper mainly describes the viewpoints to evaluate comprehensively a licensee's effort for cultivation of safety culture. (author)

Scale development of safety management system evaluation for the airline industry.

Science.gov (United States)

Chen, Ching-Fu; Chen, Shu-Chuan

2012-07-01

The airline industry relies on the implementation of Safety Management System (SMS) to integrate safety policies and augment safety performance at both organizational and individual levels. Although there are various degrees of SMS implementation in practice, a comprehensive scale measuring the essential dimensions of SMS is still lacking. This paper thus aims to develop an SMS measurement scale from the perspective of aviation experts and airline managers to evaluate the performance of company's safety management system, by adopting Schwab's (1980) three-stage scale development procedure. The results reveal a five-factor structure consisting of 23 items. The five factors include documentation and commands, safety promotion and training, executive management commitment, emergency preparedness and response plan and safety management policy. The implications of this SMS evaluation scale for practitioners and future research are discussed. Copyright © 2012 Elsevier Ltd. All rights reserved.

Statistical Hot Channel Factors and Safety Limit CHFR/OFIR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byeonghee; Park, Suki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The fuel integrity of research reactors are usually judged by comparing the critical heat flux ratio (CHFR) and the maximum fuel temperature (MFT) with the safety limits. Onset of flow instability ratio (OFIR) can also be used for the examination with CHFR. Hot channel factors (HCFs) are incorporated when calculating the CHFR/OFIR and MFT, to consider the uncertainties of fuel properties and thermo-hydraulic variables affecting them. The HCFs and safety limit CHFR is sometimes estimated to include too much conservatism, deteriorating the design flexibilities and operating margins. In this paper, a statistical estimation of HCFs and the safety limit CHFR/OFIR is presented by a random sampling of uncertainty parameters. A 15MW pool type research reactor is selected as the sample reactor for the estimation. The HCFs and the safety limit CHFR/OFIR of a 15MW pool type research reactor are evaluated statistically. The parameters affecting the HCF and the safety limit CHFR/OFIR are listed and their uncertainties are estimated. The relevant parameter uncertainties are sampled randomly and the HCFs and the safety limits are evaluated from them. The HCFs and the safety limit CHFR/OFIR with 95% probability are smaller than those estimated deterministically because the statistical evaluation convolute the correlation uncertainties and the other uncertainties in probabilistic way, whereas the deterministic evaluation simply multiply them.

A tool for safety evaluations of road improvements.

Science.gov (United States)

Peltola, Harri; Rajamäki, Riikka; Luoma, Juha

2013-11-01

Road safety impact assessments are requested in general, and the directive on road infrastructure safety management makes them compulsory for Member States of the European Union. However, there is no widely used, science-based safety evaluation tool available. We demonstrate a safety evaluation tool called TARVA. It uses EB safety predictions as the basis for selecting locations for implementing road-safety improvements and provides estimates of safety benefits of selected improvements. Comparing different road accident prediction methods, we demonstrate that the most accurate estimates are produced by EB models, followed by simple accident prediction models, the same average number of accidents for every entity and accident record only. Consequently, advanced model-based estimates should be used. Furthermore, we demonstrate regional comparisons that benefit substantially from such tools. Comparisons between districts have revealed significant differences. However, comparisons like these produce useful improvement ideas only after taking into account the differences in road characteristics between areas. Estimates on crash modification factors can be transferred from other countries but their benefit is greatly limited if the number of target accidents is not properly predicted. Our experience suggests that making predictions and evaluations using the same principle and tools will remarkably improve the quality and comparability of safety estimations. Copyright © 2013 Elsevier Ltd. All rights reserved.

Evaluation of Influence Factors within Implementing of Nuclear Safety Culture in Embarking Countries

International Nuclear Information System (INIS)

Situmorang, J.

2016-01-01

The evaluation of the implementation nuclear safety culture at BATAN has been performed. BATAN is Indonesia’s national nuclear energy agency. Nowadays, BATAN is planning to develop an experimental power reactor. To implement the nuclear safety culture BATAN has issued BATAN chairman regulation (Perka BATAN 200). Perka BATAN is the reference for individuals and organizations to implement nuclear safety culture which includes basic principles, mechanisms, assessment, as well as the implementation of the application of safety culture. It covers the establishment of safety policies, program development, program implementation, development and measurement of safety culture. Each facilities within BATAN is expected to well implement a safety culture. The implementation of safety culture is developed by considering the characteristics, attributes and indicators. The characteristics, attributes and indicators referenced are elaborated from the IAEA. The activities to strengthen safety culture are monthly workshop with participants is head of every facilities, safety leadership training and workshop for safety division manager in every facilities. It is also issued a handbook of safety that is distributed to all employees BATAN.

Evaluation of a survey tool to measure safety climate in Australian hospital pharmacy staff.

Science.gov (United States)

Walpola, Ramesh L; Chen, Timothy F; Fois, Romano A; Ashcroft, Darren M; Lalor, Daniel J

Safety climate evaluation is increasingly used by hospitals as part of quality improvement initiatives. Consequently, it is necessary to have validated tools to measure changes. To evaluate the construct validity and internal consistency of a survey tool to measure Australian hospital pharmacy patient safety climate. A 42 item cross-sectional survey was used to evaluate the patient safety climate of 607 Australian hospital pharmacy staff. Survey responses were initially mapped to the factor structure previously identified in European community pharmacy. However, as the data did not adequately fit the community pharmacy model, participants were randomly split into two groups with exploratory factor analysis performed on the first group (n = 302) and confirmatory factor analyses performed on the second group (n = 305). Following exploratory factor analysis (59.3% variance explained) and confirmatory factor analysis, a 6-factor model containing 28 items was obtained with satisfactory model fit (χ 2 (335) = 664.61 p  0.643) and model nesting between the groups (Δχ 2 (22) = 30.87, p = 0.10). Three factors (blame culture, organisational learning and working conditions) were similar to those identified in European community pharmacy and labelled identically. Three additional factors (preoccupation with improvement; comfort to question authority; and safety issues being swept under the carpet) highlight hierarchical issues present in hospital settings. This study has demonstrated the validity of a survey to evaluate patient safety climate of Australian hospital pharmacy staff. Importantly, this validated factor structure may be used to evaluate changes in safety climate over time. Copyright © 2016 Elsevier Inc. All rights reserved.

14 CFR 31.25 - Factor of safety.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Factor of safety. 31.25 Section 31.25... STANDARDS: MANNED FREE BALLOONS Strength Requirements § 31.25 Factor of safety. (a) Except as specified in paragraphs (b) and (c) of this section, the factor of safety is 1.5. (b) A factor of safety of at least five...

Evaluation of reactor safety

Energy Technology Data Exchange (ETDEWEB)

NONE

1960-04-15

Although the operation of nuclear reactors has a remarkably good record of safety, the prevention of possible reactor accidents is one of the major factors that atomic planners have to contend with. At the same time, excessive caution may breed an attitude that hampers progress, either by resisting new development or by demanding unnecessarily elaborate and expensive precautions out of proportion to the actual hazards involved. The best course obviously is to determine the possible dangers and adopt adequate measures for their prevention, providing of course, for a reasonable margin of error in judging the hazards and the effectiveness of the measures. The greater the expert understanding and thoroughness with which this is done, the narrower need the margin be. This is the basic idea behind the evaluation of reactor safety

Safety factors for neutron fluences in NPP safety assessment

International Nuclear Information System (INIS)

Demekhin, V.L.; Bukanov, V.N.; Il'kovich, V.V.; Pugach, A.M.

2016-01-01

In accordance with global practice and a number of existing regulations, the use of conservative approach is required for the calculations related to nuclear safety assessment of NPP. It implies the need to consider the determination of neutron fluence errors that is rather complicated. It is proposed to carry out the consideration by the way of multiplying the neutron fluences obtained with transport calculations by safety factors. The safety factor values are calculated by the developed technique based on the theory of errors, features of the neutron transport calculation code and the results obtained with the code. It is shown that the safety factor value is equal 1.18 with the confidence level of not less than 0.95 for the majority of VVER-1000 reactor places where neutron fluences are determined by MCPV code, and its maximum value is 1.25

Influence of organizational factors on safety

International Nuclear Information System (INIS)

Haber, S.B.; Metlay, D.S.; Crouch, D.A.

1990-01-01

There is a need for a better understanding of exactly how organizational management factors at a nuclear power plant (NPP) affect plant safety performance, either directly or indirectly, and how these factors might be observed, measured, and evaluated. The purpose of this research project is to respond to that need by developing a general methodology for characterizing these organizational and management factors, systematically collecting information on their status and integrating that information into various types of evaluative activities. Research to date has included the development of the Nuclear Organization and Management Analysis Concept (NOMAC) of a NPP, the identification of key organizational and management factors, and the identification of the methods for systematically measuring and analyzing the influence of these factors on performance. Most recently, two field studies, one at a fossil fuel plant and the other at a NPP, were conducted using the developed methodology. Results are presented from both studies highlighting the acceptability, practicality, and usefulness of the methods used to assess the influence of various organizational and management factors including culture, communication, decision-making, standardization, and oversight. 6 refs., 3 figs., 1 tab

Application of Mixed Group Decision Making to Safety Evaluation of Agricultural Products

Institute of Scientific and Technical Information of China (English)

2012-01-01

In view of the gravity of issues concerning safety of agricultural products and urgency of resolving these issues,after analyzing the problems existing in safety of agricultural products,this article offers a method for evaluating safety of agricultural products on the basis of mixed group decision making.First of all,it introduces the factors influencing safety evaluation of agricultural products;subsequently,given that the judgment matrices offered by the group of experts contain both reciprocal and complementary judgment matrices in the process of jointly participating in evaluation arising from personal preference,it proposes to assemble expert information in order to obtain indicator weight using the OWA operator;finally,the process of evaluating safety of agricultural products is given.

14 CFR 29.303 - Factor of safety.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Factor of safety. 29.303 Section 29.303... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Strength Requirements General § 29.303 Factor of safety. Unless otherwise provided, a factor of safety of 1.5 must be used. This factor applies to external and inertia...

14 CFR 27.303 - Factor of safety.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Factor of safety. 27.303 Section 27.303... STANDARDS: NORMAL CATEGORY ROTORCRAFT Strength Requirements General § 27.303 Factor of safety. Unless otherwise provided, a factor of safety of 1.5 must be used. This factor applies to external and inertia...

Prophylaxis in congenital factor VII deficiency: indications, efficacy and safety. Results from the Seven Treatment Evaluation Registry (STER).

Science.gov (United States)

Napolitano, Mariasanta; Giansily-Blaizot, Muriel; Dolce, Alberto; Schved, Jean F; Auerswald, Guenter; Ingerslev, Jørgen; Bjerre, Jens; Altisent, Carmen; Charoenkwan, Pimlak; Michaels, Lisa; Chuansumrit, Ampaiwan; Di Minno, Giovanni; Caliskan, Umran; Mariani, Guglielmo

2013-04-01

Because of the very short half-life of factor VII, prophylaxis in factor VII deficiency is considered a difficult endeavor. The clinical efficacy and safety of prophylactic regimens, and indications for their use, were evaluated in factor VII-deficient patients in the Seven Treatment Evaluation Registry. Prophylaxis data (38 courses) were analyzed from 34 patients with severe factor VII deficiency (factor VII (24 courses), four received plasma-derived factor VII, and ten received fresh frozen plasma. Prophylactic schedules clustered into "frequent" courses (three times weekly, n=23) and "infrequent" courses (≤ 2 times weekly, n=15). Excluding courses for menorrhagia, "frequent" and "infrequent" courses produced 18/23 (78%) and 5/12 (41%) "excellent" outcomes, respectively; relative risk, 1.88; 95% confidence interval, 0.93-3.79; P=0.079. Long term prophylaxis lasted from 1 to >10 years. No thrombosis or new inhibitors occurred. In conclusion, a subset of patients with factor VII deficiency needed prophylaxis because of severe bleeding. Recombinant activated factor VII schedules based on "frequent" administrations (three times weekly) and a 90 μg/kg total weekly dose were effective. These data provide a rationale for long-term, safe prophylaxis in factor VII deficiency.

14 CFR 23.303 - Factor of safety.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Factor of safety. 23.303 Section 23.303... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES Structure General § 23.303 Factor of safety. Unless otherwise provided, a factor of safety of 1.5 must be used. ...

Partial Safety Factors for Rubble Mound Breakwaters

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Burcharth, H. F.; Christiani, E.

1995-01-01

On the basis of the failure modes formulated in the various subtasks calibration of partial safety factors are described in this paper. The partial safety factors can be used to design breakwaters under quite different design conditions, namely probabilities of failure from 0.01 to 0.4, design...... lifetimes from 20 to 100 years and different qualities of wave data. A code of practice where safety is taken into account using partial safety factors is called a level I code. The partial safety factors are calibrated using First Order Reliability Methods (FORM, see Madsen et al. [1]) where...... in section 3. First Order Reliability Methods are described in section 4, and in section 5 it is shown how partial safety factors can be introduced and calibrated. The format of a code for design and analysis of rubble mound breakwaters is discussed in section 6. The mathematical formulation of the limit...

Prospective safety performance evaluation on construction sites.

Science.gov (United States)

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

2015-05-01

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

Discussion on the Safety Factors of Slopes Recommended for Small Dams

Directory of Open Access Journals (Sweden)

Jan Vrubel

2017-01-01

Full Text Available The design and assessment of the slope stability of small embankment dams is usually not carried out using slope stability calculations but rather by the comparison of proposed or existing dam slopes with those recommended by technical standards or guidelines. Practical experience shows that in many cases the slopes of small dams are steeper than those recommended. However, most of such steeper slopes at existing dams do not exhibit any visible signs of instability, defects or sliding. For the dam owner and also for dam stability engineers, the safety of the slope, expressed e.g. via a factor of safety, is crucial. The aim of this study is to evaluate the safety margin provided by recommended slopes. The factor of safety was evaluated for several dam shape and layout variants via the shear strength reduction method using PLAXIS software. The study covers various dam geometries, dam core and shoulder positions and parameter values of utilised soils. Three load cases were considered: one with a steady state seepage condition and two with different reservoir water level drawdown velocities – standard and critical. As numerous older small dams lack a drainage system, variants with and without a toe drain were assessed. Calculated factors of safety were compared with required values specified by national standards and guidelines.

14 CFR 25.303 - Factor of safety.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Factor of safety. 25.303 Section 25.303... STANDARDS: TRANSPORT CATEGORY AIRPLANES Structure General § 25.303 Factor of safety. Unless otherwise specified, a factor of safety of 1.5 must be applied to the prescribed limit load which are considered...

Impact of Geotechnical Factors on the Safety of Low Embankment Dams From the Perspective of Technical and Safety Supervision

Directory of Open Access Journals (Sweden)

Kasana Andrej

2015-03-01

Full Text Available Our research deals with a broad spectrum of problems concerning the variability of geotechnical factors and their influence on the safety of the biggest group of dam constructions in Slovakia, i.e., low earthfill dams. Its specific aim is the observation of their risk factors by using our experience and knowledge gained while working in the sector of technical and safety supervision. To achieve the aims of a research thesis, we analyzed 39 low earthfill dams. We performed observations and documented their conditions with the aim of clarifying the risk factors. After an analysis of the information materials that characterize dams and after a statistical analysis of the measurement results in situ, including measurements from technical and safety supervision databases, we performed an analysis by using mathematical modeling to evaluate the safety of the dam constructions. Out of the total number of 39 dam constructions, an analysis of the stability of the dam slopes was performed on 37 dams, and deformation problems were analyzed on 28 of the dams. Filtration problems were analyzed at 26 dams, and a complete evaluation of the intensity of filtration movements was performed on 19 of the constructions.

Verification of Overall Safety Factors In Deterministic Design Of Model Tested Breakwaters

DEFF Research Database (Denmark)

Burcharth, H. F.

2001-01-01

The paper deals with concepts of safety implementation in design. An overall safety factor concept is evaluated on the basis of a reliability analysis of a model tested rubble mound breakwater with monolithic super structure. Also discussed are design load identification and failure mode limit...

10CFR50.59 safety evaluations

International Nuclear Information System (INIS)

Grime, L.; Page, E.

1987-01-01

As a plant changes from the design phase to the operational phase, new regulations and standards apply. One such regulation is 10CFR50.59 on safety evaluations. Once an operating license is issued, it is mandatory to submit all applicable changes, tests, and experiments to the safety evaluation process. As preparation for this transition, Detroit Edison had procedures in place and conducted personnel training. Reviews of the safety engineering were conducted by the on-site review board. The off-site board delegated detailed reviews of most safety evaluations to the independent safety evaluation group (ISEG). The on-site group review included presentation of complete design packages by engineers. The ISEG and off-site review group's activity focused on safety evaluation. This paper addresses industry trends that were studied, Detroit Edison's recent actions, and industry issues related to 10CFR50.59 safety evaluations

A hierarchical factor analysis of a safety culture survey.

Science.gov (United States)

Frazier, Christopher B; Ludwig, Timothy D; Whitaker, Brian; Roberts, D Steve

2013-06-01

Recent reviews of safety culture measures have revealed a host of potential factors that could make up a safety culture (Flin, Mearns, O'Connor, & Bryden, 2000; Guldenmund, 2000). However, there is still little consensus regarding what the core factors of safety culture are. The purpose of the current research was to determine the core factors, as well as the structure of those factors that make up a safety culture, and establish which factors add meaningful value by factor analyzing a widely used safety culture survey. A 92-item survey was constructed by subject matter experts and was administered to 25,574 workers across five multi-national organizations in five different industries. Exploratory and hierarchical confirmatory factor analyses were conducted revealing four second-order factors of a Safety Culture consisting of Management Concern, Personal Responsibility for Safety, Peer Support for Safety, and Safety Management Systems. Additionally, a total of 12 first-order factors were found: three on Management Concern, three on Personal Responsibility, two on Peer Support, and four on Safety Management Systems. The resulting safety culture model addresses gaps in the literature by indentifying the core constructs which make up a safety culture. This clarification of the major factors emerging in the measurement of safety cultures should impact the industry through a more accurate description, measurement, and tracking of safety cultures to reduce loss due to injury. Copyright © 2013 National Safety Council and Elsevier Ltd. All rights reserved.

Preparation of the requirements for the safety regulation related to human and organizational factors

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The outline of the project in the current fiscal year is to investigate and analyze issues associated with Human and Organizational Factors involved in incidents of nuclear facilities, and to study and develop evaluation methods of these countermeasures. The guideline to evaluate licensee's safety culture and root cause analysis (RCA) had been developed for further improving safety on nuclear power plants at 2007. These guidelines have been used at regulatory inspection since that time. Based on experience of using these existing guidelines, some activities for improving guidelines are now under investigation; these are selecting candidate quantitative indicators for safety culture evaluation and researching good practices for RCA issues. JNES implemented human factor analysis about 18 domestic events including the Fukushima Dai-ichi nuclear power plant accident. (author)

Safer electronic health records safety assurance factors for EHR resilience

CERN Document Server

Sittig, Dean F

2015-01-01

This important volume provide a one-stop resource on the SAFER Guides along with the guides themselves and information on their use, development, and evaluation. The Safety Assurance Factors for EHR Resilience (SAFER) guides, developed by the editors of this book, identify recommended practices to optimize the safety and safe use of electronic health records (EHRs). These guides are designed to help organizations self-assess the safety and effectiveness of their EHR implementations, identify specific areas of vulnerability, and change their cultures and practices to mitigate risks.This book pr

Development of HANARO human factors management plan and evaluation of BCS display

International Nuclear Information System (INIS)

Oh, I. S.; Lee, J. W.; Lee, Y. H.

2004-01-01

In this study, human factors evaluation of BCS display design was performed. We adopted the suitability of design elements of BCS display as human factors evaluation measure. And, we also adopted guideline based evaluation, field survey and expert evaluation as evaluation method. The checklist was utilized for the evaluation, and the results of evaluation were well arranged in the evaluation format. We did not find out the HED (Human Engineering Discrepancy) impede safety of HANARO, except some necessary items to improve during short periods. We also provide some items of improvement for the enhancement of safety and operator's performance in the aspect of long periods. If the proposed improvement items were completely fulfilled, the more improved safety of HANARO will be secured

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.)

A probabilistic bridge safety evaluation against floods.

Science.gov (United States)

Liao, Kuo-Wei; Muto, Yasunori; Chen, Wei-Lun; Wu, Bang-Ho

2016-01-01

To further capture the influences of uncertain factors on river bridge safety evaluation, a probabilistic approach is adopted. Because this is a systematic and nonlinear problem, MPP-based reliability analyses are not suitable. A sampling approach such as a Monte Carlo simulation (MCS) or importance sampling is often adopted. To enhance the efficiency of the sampling approach, this study utilizes Bayesian least squares support vector machines to construct a response surface followed by an MCS, providing a more precise safety index. Although there are several factors impacting the flood-resistant reliability of a bridge, previous experiences and studies show that the reliability of the bridge itself plays a key role. Thus, the goal of this study is to analyze the system reliability of a selected bridge that includes five limit states. The random variables considered here include the water surface elevation, water velocity, local scour depth, soil property and wind load. Because the first three variables are deeply affected by river hydraulics, a probabilistic HEC-RAS-based simulation is performed to capture the uncertainties in those random variables. The accuracy and variation of our solutions are confirmed by a direct MCS to ensure the applicability of the proposed approach. The results of a numerical example indicate that the proposed approach can efficiently provide an accurate bridge safety evaluation and maintain satisfactory variation.

Method of calculating the safety factor profile on the HT-7 tokamak

International Nuclear Information System (INIS)

Zhang Xianmei; Lu Yuancheng; Wan Baonian

2001-01-01

A method of calculating the safety factor profile on the HT-7 tokamak has been described. It is derived from Maxwell's equations, among which the authors mainly use two of them: one is the magnetic field diffusion equation, and the other is Ampere's Law. This method can be also used to evaluate the safety factor on other devices with a circular cross sections. It is helpful to the study of the plasma MHD behavior on the HT-7 tokamak

A probabilistic analysis method to evaluate the effect of human factors on plant safety

International Nuclear Information System (INIS)

Ujita, H.

1987-01-01

A method to evaluate the effect of human factors on probabilistic safety analysis (PSA) is developed. The main features of the method are as follows: 1. A time-dependent multibranch tree is constructed to treat time dependency of human error probability. 2. A sensitivity analysis is done to determine uncertainty in the PSA due to branch time of human error occurrence, human error data source, extraneous act probability, and human recovery probability. The method is applied to a large-break, loss-of-coolant accident of a boiling water reactor-5. As a result, core melt probability and risk do not depend on the number of time branches, which means that a small number of branches are sufficient. These values depend on the first branch time and the human error probability

A Study on the Holding Capacity Safety Factors for Torpedo Anchors

Directory of Open Access Journals (Sweden)

Luís V. S. Sagrilo

2012-01-01

Full Text Available The use of powerful numerical tools based on the finite-element method has been improving the prediction of the holding capacity of fixed anchors employed by the offshore oil industry. One of the main achievements of these tools is the reduction of the uncertainty related to the holding capacity calculation of these anchors. Therefore, it is also possible to reduce the values of the associated design safety factors, which have been calibrated relying on models with higher uncertainty, without impairing the original level of structural safety. This paper presents a study on the calibration of reliability-based safety factors for the design of torpedo anchors considering the statistical model uncertainty evaluated using results from experimental tests and their correspondent finite-element-based numerical predictions. Both working stress design (WSD and load and resistance factors design (LRFD design methodologies are investigated. Considering the WSD design methodology, the single safety is considerably lower than the value typically employed in the design of torpedo anchors. Moreover, a LRFD design code format for torpedo anchors is more appropriate since it leads to designs having less-scattered safety levels around the target value.

Relationship between organizational factors, safety culture and PSA in nuclear power plant operations

International Nuclear Information System (INIS)

Joksimovich, V.; Orvis, D.D.

1997-01-01

There are four nuclear safety imperatives or ''4Ms'': machine (hardware, design, QA/QC), milieux (operating conditions, environment, natural phenomena), man (human reliability) and management (organizational and management influences). Nuclear safety evaluations as well as evolution of its most powerful tool, Probabilistic Safety Assessment (PSA), followed chronologically the 4M constituents. The nuclear industry worldwide, and the nuclear safety regulators in particular, have been preoccupied with the first M almost to the point of obsession with belated and only intuitive interest in the third and fourth M (human dimension). Human factors or economics in the nuclear industry was an afterthought. Human reliability was essentially born in the aftermath of the Three Mile Island (TMI) accident. Impact of organizational factors on nuclear safety is only in the early stages of R and D. This paper describes some of the concepts being pursued by APG to link organizational factors and safety culture to Human Reliability Analysis (HRA) and to integrate such into probabilistic safety assessment (PSA), e.g. [APG, 1993]. (author). 11 refs, 4 figs, 1 tab

Evaluation of periodic safety status analyses

International Nuclear Information System (INIS)

Faber, C.; Staub, G.

1997-01-01

In order to carry out the evaluation of safety status analyses by the safety assessor within the periodical safety reviews of nuclear power plants safety goal oriented requirements have been formulated together with complementary evaluation criteria. Their application in an inter-disciplinary coopertion covering the subject areas involved facilitates a complete safety goal oriented assessment of the plant status. The procedure is outlined briefly by an example for the safety goal 'reactivity control' for BWRs. (orig.) [de

Organizational factors in nuclear safety

International Nuclear Information System (INIS)

Wilpert, Bernhard

2000-01-01

The overall picture of factors which contributed to the event presents a panorama of a NPP where organizational and managerial characteristics were intricately intertwined and emerged as crucial for a general deterioration of the plant's capabilities to continually correct its deficiencies and optimize its operations. In the following author shall attempt to first cover various important efforts to modeling organizational factors relevant to safety. The second part of my presentation will offer an attempt towards an integrative model. The third part concludes with an agenda for research and practice. Most of the twelve different approaches above attempt to consider safety relevant organizational factors by way of pragmatic classifications. Together with their sub-categories we can count close to 160 different factors on various levels of abstraction. This is tantamount to say that most approaches lack systematic theoretical underpinnings. Thus then arises the question whether we need to develop a generic model, which promises to encompass these three major approaches altogether. Practical issues emerge particularly in the domain of organizational development, i.e. the goal oriented efforts to change the structures and the functioning of nuclear operations in such a way that the desired outputs in terms safety and reliability result in a sustained fashion. Again, these practical concerns are intimately related to developments and advances in theory and methodology. Only a close cooperation among scientists from various disciplines and of practitioners holds the promise of adequately understanding and use of organizational factors in future improving the safety record of nuclear industry worldwide. (S.Y.)

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

Directory of Open Access Journals (Sweden)

Zi-wu Fan

2009-06-01

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

Evaluating the Effectiveness of an Educational Intervention to Improve the Patient Safety Attitudes of Intern Pharmacists.

Science.gov (United States)

Walpola, Ramesh L; Fois, Romano A; McLachlan, Andrew J; Chen, Timothy F

2017-02-25

Objective. To evaluate the effectiveness of a face-to-face educational intervention in improving the patient safety attitudes of intern pharmacists. Methods. A patient safety education program was delivered to intern pharmacists undertaking The University of Sydney Intern Training Program in 2014. Their patient safety attitudes were evaluated immediately prior to, immediately after, and three-months post-intervention. Underlying attitudinal factors were identified using exploratory factor analysis. Changes in factor scores were examined using analysis of variance. Results. Of the 120 interns enrolled, 95 (78.7%) completed all three surveys. Four underlying attitudinal factors were identified: attitudes towards addressing errors, questioning behaviors, blaming individuals, and reporting errors. Improvements in all attitudinal factors were evident immediately after the intervention. However, only improvements in attitudes towards blaming individuals involved in errors were sustained at three months post-intervention. Conclusion. The educational intervention was associated with short-term improvements in pharmacist interns' patient safety attitudes. However, other factors likely influenced their attitudes in the longer term.

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

Survey of factors associated with nurses' perception of patient safety.

Science.gov (United States)

Park, Sun A; Lee, Sui Jin; Choi, Go Un

2011-01-01

To describe the nurses' perception of hospital organization related to cultural issues on the safety of the patient and reporting medical errors. In addition, to identify factors associated with the safety of the patient and the nurse. A survey conducted during December 2008-Jannuary 2009, with 126 nurses using the Korean version of the AHRQ patient safety survey, a self-report 5-point Likert scale. Stata 10.0 was used for descriptive analysis, ANOVA (Analysis of variance) and logistic regression. National Cancer Center in Korea. The means for a working environment related to patient safety was 3.4 (±0.62). The associated factors of duration were at a present hospital, a special area, and direct contact with patients. Among organizational culture factors related to patient safety, the means were 3.81(±0.54) for the boss/manager's perception of patient safety and 3.37(±0.49) for the cooperation/collaboration between units. The frequent number of errors reported by nurses were 1~2(22.2%) times over the past 12 months. For incidence reporting, the items that the 'nurses perceived for communication among clinicians as fair' had a means of 3.23(±0.40) and the 'overall evaluation of patient safety was a good' 3.34(±0.73). The nurses' perception of cooperation and collaboration between units were associated with the direct contact between the patient and the nurse. The frequency of incidence reporting was associated with the duration of working hours at the present hospital and also their work experience. The nurses' perception of hospital environment, organizational culture, and incidence reporting was above average and mostly associated with organizational culture.

Study on Fuzzy Comprehensive Evaluation Model for the Safety of Mine Belt Conveyor

Directory of Open Access Journals (Sweden)

Gong Xiaoyan

2017-01-01

Full Text Available To improve the situation of the frequent failures of mine belt conveyor during operation, a model was used to evaluate the safety of mine belt conveyor. Based on the foundation of collecting and analyzing a large quantity of fault information of belt conveyor in the nationwide coal mine, the fault tree model of belt conveyor has been built, then the safety evaluation index system was established by analyzing and removing some secondary indicators. Furthermore, the weighted value of safety evaluation indexs was determined by analytic hierarchy process(AHP, and the single factor fuzzy evaluation matrix was constructed by experts grading method. Additionally, the model was applied in evaluating the security of belt conveyor in Nanliang coal mine. The results shows the security level is recognized to the “general”, which means that this model can be adopted widely in evaluating the safety of mine belt conveyor.

Geoscientific evaluation factors and criteria for siting and site evaluation. Progress report

International Nuclear Information System (INIS)

Stroem, A.; Ericsson, Lars O.; Svemar, C.; Almen, K.E.; Andersson, Johan

1999-03-01

The purposes of the present report are to: present the work that has been done to identify the parameters that need to be determined in a geoscientific site investigation and that serve as the basis for the work with geoscientific evaluation factors; give a progress report from the project that was initiated in 1997 named Siting Factors and Criteria for Site Evaluation, with an emphasis on definitions, outline and structure for the execution of the work; present geoscientific requirements on function both general and in detail in the form of an example for the discipline rock mechanics; present geoscientific evaluation factors associated with different stages in the siting work in the form of an example for the discipline hydrogeochemical composition; present plans for further work as regards criteria for site evaluation in different siting stages. The project is under way, and this is to be regarded as a progress report since e.g. criteria for site evaluation will be presented at a later date. The long-term performance and safety of the deep repository must always be evaluated by means of an integrated safety assessment. The work with factors and criteria can never take the place of such an assessment, but can provide guidance regarding its outcome. Requirements and preferences regarding the function of the rock in the deep repository have been clarified in this progress report. What is new here is the structuring that has been carried out, with a classification into different geoscientific disciplines, and the formalism that has been given to the terms requirement, preference and function. Based on fundamental safety and construction functions, requirements on function have been specified for the disciplines geology, thermal properties, hydro-geology, rock mechanics, chemistry and transport properties. Furthermore, function analyses have been identified by means of which it is possible to concretize requirements on function and which geoscientific parameters are

Geoscientific evaluation factors and criteria for siting and site evaluation. Progress report

Energy Technology Data Exchange (ETDEWEB)

Stroem, A.; Ericsson, Lars O.; Svemar, C. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Almen, K.E. [KEA GEO-konsult AB (Sweden); Andersson, Johan [Golder Grundteknik KB (Sweden)

1999-03-01

The purposes of the present report are to: present the work that has been done to identify the parameters that need to be determined in a geoscientific site investigation and that serve as the basis for the work with geoscientific evaluation factors; give a progress report from the project that was initiated in 1997 named Siting Factors and Criteria for Site Evaluation, with an emphasis on definitions, outline and structure for the execution of the work; present geoscientific requirements on function both general and in detail in the form of an example for the discipline rock mechanics; present geoscientific evaluation factors associated with different stages in the siting work in the form of an example for the discipline hydrogeochemical composition; present plans for further work as regards criteria for site evaluation in different siting stages. The project is under way, and this is to be regarded as a progress report since e.g. criteria for site evaluation will be presented at a later date. The long-term performance and safety of the deep repository must always be evaluated by means of an integrated safety assessment. The work with factors and criteria can never take the place of such an assessment, but can provide guidance regarding its outcome. Requirements and preferences regarding the function of the rock in the deep repository have been clarified in this progress report. What is new here is the structuring that has been carried out, with a classification into different geoscientific disciplines, and the formalism that has been given to the terms requirement, preference and function. Based on fundamental safety and construction functions, requirements on function have been specified for the disciplines geology, thermal properties, hydro-geology, rock mechanics, chemistry and transport properties. Furthermore, function analyses have been identified by means of which it is possible to concretize requirements on function and which geoscientific parameters are

Evaluation on safety issues of SMART

International Nuclear Information System (INIS)

Kim, W. S.; Seol, K. W.; Yoon, Y. K.; Lee, J. H.

2001-01-01

Safety issues on the SMART were evaluated in the light of the compliance with the Ministerial Ordinance of Technical Requirements applying to Nuclear Installations, which was recently revised. Evaluation concludes that regulatory requirements associated with following items have to be developed as the licensing criteria for the SMART: (1) proving the safety of design or materials different form existing reactors; (2) coping with beyond design basis accidents; (3) rulemaking on the safety of reactor safeguard vessel ; (4) ensuring integrity of steam generator tubes; and (5) classifying equipment based on their safety significance. Appropriate actions including implementation of new requirements under development should be taken for safety issues such as diversity of reactivity control and in-service inspection of steam generator tubes that are not complied with the current Technical Requirements. Safety level of the SMART design will be evaluated further by the more detailed assessment according to the Technical Requirements, and additional safety issues will be identified and resolved, if it necessary

LNG Safety Assessment Evaluation Methods

Energy Technology Data Exchange (ETDEWEB)

Muna, Alice Baca [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-05-01

Sandia National Laboratories evaluated published safety assessment methods across a variety of industries including Liquefied Natural Gas (LNG), hydrogen, land and marine transportation, as well as the US Department of Defense (DOD). All the methods were evaluated for their potential applicability for use in the LNG railroad application. After reviewing the documents included in this report, as well as others not included because of repetition, the Department of Energy (DOE) Hydrogen Safety Plan Checklist is most suitable to be adapted to the LNG railroad application. This report was developed to survey industries related to rail transportation for methodologies and tools that can be used by the FRA to review and evaluate safety assessments submitted by the railroad industry as a part of their implementation plans for liquefied or compressed natural gas storage ( on-board or tender) and engine fueling delivery systems. The main sections of this report provide an overview of various methods found during this survey. In most cases, the reference document is quoted directly. The final section provides discussion and a recommendation for the most appropriate methodology that will allow efficient and consistent evaluations to be made. The DOE Hydrogen Safety Plan Checklist was then revised to adapt it as a methodology for the Federal Railroad Administration’s use in evaluating safety plans submitted by the railroad industry.

Human Factors Evaluation of Procedures for Periodic Safety Review of Yonggwang Unit no. 1, 2

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Hee; Lee, Jung Woon; Park, Jae Chang (and others)

2006-01-15

This report describes the results of human factors assessment on the plant operating procedures as part of Periodic Safety Review(PSR) of Yonggwang Nuclear Power Plant Unit no. 1, 2. The suitability of item and appropriateness of format and structure in the key operating procedures of nuclear power plants were investigated by the review of plant operating experiences and procedure documents, field survey, and experimental assessment on some part of procedures. A checklist was used to perform this assessment and record the review results. The reviewed procedures include EOP(Emergency Operating Procedures), GOP(General Operating Procedures), AOP(Abnormal Operating Procedures), and management procedures of some technical departments. As results of the assessments, any significant problem challenging the safety was not found on the human factors in the operating procedures. However, several small items to be changed and improved were discovered. An action plan is recommended to accommodate the suggestions and review comments. It will enhance the plant safety on the operating procedure.

Multimethods approach to safety-parameter-display evaluation

International Nuclear Information System (INIS)

Banks, W.W.; Blackman, H.S.; Gertman, D.I.; Petersen, R.J.

1982-01-01

The Human Factors Engineering Office of EG and G Idaho performed this NRC-funded study to assist the NRC in objectively assessing licensee-developed safety parameter display (SPD) formats and designs. The purpose of this study was to quantitatively measure the degree to which a tachistoscopic method of display evaluation would correlate with the results of a multidimensional rating approach to display evaluation. Results of the following three experiments will be presented; (a) tachistoscopic, (b) multidimensional rating scale, and (c) the combined results of a and b. The test material for all experiments consisted of three multivariate data display formats all under development as SPDs for reactor control rooms presenting safety parameter display data at the loss-of-fluid test (LOFT) facility. The three display formats studied were stars, deviation bar graphs, and meters. Eighteen adult volunteers were used as subjects. All were currently qualified reactor operators from the LOFT reactor plant, with a mean of 9.4 years reactor operating experience

Organizational factors influencing improvements in safety

International Nuclear Information System (INIS)

Marcus, A.; Nichols, M.L.; Olson, J.; Osborn, R.; Thurber, J.

1992-01-01

Research reported here seeks to identify the key organizational factors that influence safety-related performance indicators in nuclear power plants over time. It builds upon organizational factors identified in NUREG/CR-5437, and begins to develop a theory of safety-related performance and performance improvement based on economic and behavioral theories of the firm. Central to the theory are concepts of past performance, problem recognition, resource availability, resource allocation, and business strategies that focus attention. Variables which reflect those concepts are combined in statistical models and tested for their ability to explain scrams, safety system actuations, significant events, safety system failures, radiation exposure, and critical hours. Results show the performance indicators differ with respect to the sets of variables which serve as the best predictors of future performance, and past performance is the most consistent predictor of future performance

Development of a draft of human factors safety review procedures for the Korean Next Generation Reactor

International Nuclear Information System (INIS)

Lee, Jung Woon; Moon, B. S.; Park, J. C.; Lee, Y. H.; Oh, I. S.; Lee, H. C.

2000-02-01

In this study, a draft of Human Factors Engineering (HFE) Safety Review Procedures (SRP) was developed for the safety review of KNGR based on HFE Safety and Regulatory Requirements and Guidelines (SRRG). This draft includes acceptance criteria, review procedure, and evaluation findings for the areas of review including HFE program management, human factors analyses, human factors design, and HFE verification and validation, based on section 15.1 'human factors engineering design process' and 15.2 'control room human factors engineering' of KNGR specific safety requirements and chapter 15 'human factors engineering' of KNGR safety regulatory guides. For the effective review, human factors concerns or issues related to advanced HSI design that have been reported so far should be extensively examined. In this study, a total of 384 human factors issues related to the advanced HSI design were collected through our review of a total of 145 documents. A summary of each issue was described and the issues were identified by specific features of HSI design. These results were implemented into a database system

Development of the safety evaluation system in the respects of organizational factors and workers' consciousness. Pt. 5. Application of the system for industries except electric power industry

International Nuclear Information System (INIS)

Hasegawa, Naoko; Hirose, Ayako; Hayase, Kenichi; Sasou Kunihide; Takano, Kenichi

2004-01-01

The purpose of our study is to develop a safety evaluation system which clarifies the safety level of an organization. As a basic method of evaluation using a questionnaire had been established, now that the generalization is needed for the system. Hence, this paper is intended to verify the applicability of the system for eight manufacture industries. The investigation using a questionnaire was conducted for 125 factories' workers. The following results were obtained: 1) The Comprehensive Safety Index (CSI) taking into account individual and organizational factors was identified using the principal component analysis. 2) Although the criterion-related validity of CSI was confirmed for some industries, ti will be necessary for the advancement of the system's reliability to compile more data into the system. 3) According to the result of investigations on safety management in secure companies and the causes of current industrial accidents, it was clarified that the CSI had the content validity. 4) It seemed possible to evaluate the safety level using two different industries' data if there were similarities between the industries in the score of the CSI and the aspects to which were attached importance for the improvement of the safety. (author)

Patient safety in the operating room: an intervention study on latent risk factors

Directory of Open Access Journals (Sweden)

van Beuzekom Martie

2012-06-01

Full Text Available Abstract Background Patient safety is one of the greatest challenges in healthcare. In the operating room errors are frequent and often consequential. This article describes an approach to a successful implementation of a patient safety program in the operating room, focussing on latent risk factors that influence patient safety. We performed an intervention to improve these latent risk factors (LRFs and increase awareness of patient safety issues amongst OR staff. Methods Latent risk factors were studied using a validated questionnaire applied to the OR staff before and after an intervention. A pre-test/post-test control group design with repeated measures was used to evaluate the effects of the interventions. The staff from one operating room of an university hospital acted as the intervention group. Controls consisted of the staff of the operating room in another university hospital. The outcomes were the changes in LRF scores, perceived incident rate, and changes in incident reports between pre- and post-intervention. Results Based on pre-test scores and participants’ key concerns about organizational factors affecting patient safety in their department the intervention focused on the following LRFs: Material Resources, Training and Staffing Recourses. After the intervention, the intervention operating room - compared to the control operating room - reported significantly fewer problems on Material Resources and Staffing Resources and a significantly lower score on perceived incident rate. The contribution of technical factors to incident causation decreased significantly in the intervention group after the intervention. Conclusion The change of state of latent risk factors can be measured using a patient safety questionnaire aimed at these factors. The change of the relevant risk factors (Material and Staffing resources concurred with a decrease in perceived and reported incident rates in the relevant categories. We conclude that

Criticality safety evaluation in Tokai Reprocessing Plant

International Nuclear Information System (INIS)

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

2000-04-01

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

Organizational factors and nuclear power plant safety

International Nuclear Information System (INIS)

Haber, S.B.

1995-01-01

There are many organizations in our society that depend on human performance to avoid incidents involving significant adverse consequences. As our culture and technology have become more sophisticated, the management of risk on a broad basis has become more and more critical. The safe operation of military facilities, chemical plants, airlines, and mass transit, to name a few, are substantially dependent on the performance of the organizations that operate those facilities. The nuclear power industry has, within the past 15 years, increased the attention given to the influence of human performance in the safe operation of nuclear power plants (NPP). While NPPs have been designed through engineering disciplines to intercept and mitigate events that could cause adverse consequences, it has been clear from various safety-related incidents that human performance also plays a dominant role in preventing accidents. Initial efforts following the 1979 Three Mile Island incident focused primarily on ergonomic factors (e.g., the best design of control rooms for maximum performance). Greater attention was subsequently directed towards cognitive processes involved in the use of NPP decision support systems and decision making in general, personnel functions such as selection systems, and the influence of work scheduling and planning on employees' performance. Although each of these approaches has contributed to increasing the safety of NPPS, during the last few years, there has been a growing awareness that particular attention must be paid to how organizational processes affect NPP personnel performance, and thus, plant safety. The direct importance of organizational factors on safety performance in the NPP has been well-documented in the reports on the Three Mile Island and Chernobyl accidents as well as numerous other events, especially as evaluated by the U.S. Nuclear Regulatory Commission (NRC)

Site evaluation for nuclear installations. Safety requirements

International Nuclear Information System (INIS)

2003-01-01

This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Siting, which was issued in 1988 as Safety Series No. 50-C-S (Rev. 1). It takes account of developments relating to site evaluations for nuclear installations since the Code on Siting was last revised. These developments include the issuing of the Safety Fundamentals publication on The Safety of Nuclear Installations, and the revision of various safety standards and other publications relating to safety. Requirements for site evaluation are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear installations. It is recognized that there are steady advances in technology and scientific knowledge, in nuclear safety and in what is considered adequate protection. Safety requirements change with these advances and this publication reflects the present consensus among States. This Safety Requirements publication was prepared under the IAEA programme on safety standards for nuclear installations. It establishes requirements and provides criteria for ensuring safety in site evaluation for nuclear installations. The Safety Guides on site evaluation listed in the references provide recommendations on how to meet the requirements established in this Safety Requirements publication. The objective of this publication is to establish the requirements for the elements of a site evaluation for a nuclear installation so as to characterize fully the site specific conditions pertinent to the safety of a nuclear installation. The purpose is to establish requirements for criteria, to be applied as appropriate to site and site-installation interaction in operational states and accident conditions, including those that could lead to emergency measures for: (a) Defining the extent of information on a proposed site to be presented by the applicant; (b) Evaluating a proposed site to ensure that the site

Guide for understanding and evaluation of safety culture

International Nuclear Information System (INIS)

2008-01-01

This report was the guide of understanding and evaluation of safety culture. Operator's activities for enhancement of safety culture in nuclear installations became an object of safety regulation in the management system. Evaluation of operator's activities (including top management's involvement) to prevent degradation of safety culture and organization climate in daily works needed understanding of safety culture and diversity of operator's activities. This guide was prepared to check indications of degradation of safety culture and organization climate in operator's activities in daily works and encourage operator's activities to enhance safety culture improvement and good practice. Comprehensive evaluation of operator's activities to prevent degradation of safety culture and organization climate would be performed from the standpoints of 14 safety culture elements such as top management commitment, clear plan and implementation of upper manager, measures to avoid wrong decision making, questioning attitude, reporting culture, good communications, accountability and openness, compliance, learning system, activities to prevent accidents or incidents beforehand, self-assessment or third party evaluation, work management, change management and attitudes/motivation. Element-wise examples and targets for evaluation were attached with evaluation check tables. (T. Tanaka)

The role of psychological factors in workplace safety.

Science.gov (United States)

Kotzé, Martina; Steyn, Leon

2013-01-01

Workplace safety researchers and practitioners generally agree that it is necessary to understand the psychological factors that influence people's workplace safety behaviour. Yet, the search for reliable individual differences regarding psychological factors associated with workplace safety has lead to sparse results and inconclusive findings. The aim of this study was to investigate whether there are differences between the psychological factors, cognitive ability, personality and work-wellness of employees involved in workplace incidents and accidents and/or driver vehicle accidents and those who are not. The study population (N = 279) consisted of employees employed at an electricity supply organisation in South Africa. Mann-Whitney U-test and one-way ANOVA were conducted to determine the differences in the respective psychological factors between the groups. These results showed that cognitive ability did not seem to play a role in workplace incident/accident involvement, including driver vehicle accidents, while the wellness factors burnout and sense of coherence, as well as certain personality traits, namely conscientiousness, pragmatic and gregariousness play a statistically significant role in individuals' involvement in workplace incidents/accidents/driver vehicle accidents. Safety practitioners, managers and human resource specialists should take cognisance of the role of specifically work-wellness in workplace safety behaviour, as management can influence these negative states that are often caused by continuously stressful situations, and subsequently enhance work place safety.

Development of a draft of human factors safety review procedures for the Korean next generation reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Woon; Moon, B. S.; Park, J. C.; Lee, Y. H.; Oh, I. S.; Lee, H. C. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-02-01

In this study, a draft of human factors engineering (HFE) safety review procedures (SRP) was developed for the safety review of KNGR based on HFE Safety and Regulatory Requirements and Guidelines (SRRG). This draft includes acceptance criteria, review procedure, and evaluation findings for the areas of review including HFE Program Management, Human Factors Analyses, Human Factors Design, and HFE Verification and Validation, based on Section 15.1 'Human Factors Engineering Design Process' and 15.2 'Control Room Human Factors Engineering' of KNGR Specific Safety Requirements and Chapter 15 'Human Factors Engineering' of KNGR Safety Regulatory Guides. For the effective review, human factors concerns or issues related to advanced HSI design that have been reported so far should be extensively examined. In this study, a total of 384 human factors issues related to the advanced HSI design were collected through our review of a total of 145 documents. A summary of each issue was described and the issues were identified by specific features of HSI design. These results were implemented into a database system. 8 refs., 2 figs. (Author)

Human factors in safety and business management.

Science.gov (United States)

Vogt, Joachim; Leonhardt, Jorg; Koper, Birgit; Pennig, Stefan

2010-02-01

Human factors in safety is concerned with all those factors that influence people and their behaviour in safety-critical situations. In aviation these are, for example, environmental factors in the cockpit, organisational factors such as shift work, human characteristics such as ability and motivation of staff. Careful consideration of human factors is necessary to improve health and safety at work by optimising the interaction of humans with their technical and social (team, supervisor) work environment. This provides considerable benefits for business by increasing efficiency and by preventing incidents/accidents. The aim of this paper is to suggest management tools for this purpose. Management tools such as balanced scorecards (BSC) are widespread instruments and also well known in aviation organisations. Only a few aviation organisations utilise management tools for human factors although they are the most important conditions in the safety management systems of aviation organisations. One reason for this is that human factors are difficult to measure and therefore also difficult to manage. Studies in other domains, such as workplace health promotion, indicate that BSC-based tools are useful for human factor management. Their mission is to develop a set of indicators that are sensitive to organisational performance and help identify driving forces as well as bottlenecks. Another tool presented in this paper is the Human Resources Performance Model (HPM). HPM facilitates the integrative assessment of human factors programmes on the basis of a systematic performance analysis of the whole system. Cause-effect relationships between system elements are defined in process models in a first step and validated empirically in a second step. Thus, a specific representation of the performance processes is developed, which ranges from individual behaviour to system performance. HPM is more analytic than BSC-based tools because HPM also asks why a certain factor is

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

The human factors and the safety of experimentation reactors

International Nuclear Information System (INIS)

Jeffroy, F.; Delaporte-Normier, M.L.

2007-01-01

Inside IRSN (Institute for Radiological protection and Nuclear Safety), the mission of the Human Factors Group is to assess the way operators of nuclear installations take into account the risks related to human activities. In the last few years, IRSN has been involved in the safety analysis of different installations where Cea develops research programs, in particular experimental reactors. The first part of this article presents the methodology used by IRSN to evaluate how operators take into account risks related to human activities. This methodology is made up of 4 steps: 1) the identification of the human activities that convey a risk for the installation nuclear safety (safety-sensitive activities), for instance in the case of the Masurca reactor, it has been shown that errors made during the manufacturing of fuel tubes can lead to a criticality accident; 2) listing all the dispositions or arrangements taken to make human safety-sensitive activities more reliable; 3) checking the efficiency of such dispositions or arrangements; and 4) assessing the ability of the operators to generate the adequate dispositions or arrangements. The second part highlights the necessity to develop inside these research installations an organisation that facilitates cooperation between experimenters and operators

Organizational factors affecting safety implementation in food companies in Thailand.

Science.gov (United States)

Chinda, Thanwadee

2014-01-01

Thai food industry employs a massive number of skilled and unskilled workers. This may result in an industry with high incidences and accident rates. To improve safety and reduce the accident figures, this paper investigates factors influencing safety implementation in small, medium, and large food companies in Thailand. Five factors, i.e., management commitment, stakeholders' role, safety information and communication, supportive environment, and risk, are found important in helping to improve safety implementation. The statistical analyses also reveal that small, medium, and large food companies hold similar opinions on the risk factor, but bear different perceptions on the other 4 factors. It is also found that to improve safety implementation, the perceptions of safety goals, communication, feedback, safety resources, and supervision should be aligned in small, medium, and large companies.

Safety performance evaluation using proactive indicators in a selected industry

Directory of Open Access Journals (Sweden)

Abolfazl Barkhordari

2015-03-01

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

Nuclear safety culture evaluation model based on SSE-CMM

International Nuclear Information System (INIS)

Yang Xiaohua; Liu Zhenghai; Liu Zhiming; Wan Yaping; Peng Guojian

2012-01-01

Safety culture, which is of great significance to establish safety objectives, characterizes level of enterprise safety production and development. Traditional safety culture evaluation models emphasis on thinking and behavior of individual and organization, and pay attention to evaluation results while ignore process. Moreover, determining evaluation indicators lacks objective evidence. A novel multidimensional safety culture evaluation model, which has scientific and completeness, is addressed by building an preliminary mapping between safety culture and SSE-CMM's (Systems Security Engineering Capability Maturity Model) process area and generic practice. The model focuses on enterprise system security engineering process evaluation and provides new ideas and scientific evidences for the study of safety culture. (authors)

Determination of engineering safety factor -routine in Hungary (a methodology for the normal operation local power engineering safety factors)

International Nuclear Information System (INIS)

Szecsenyi, Z.; Korpas, L.; Bona, G.; Kereszturi, A.

2010-01-01

From the late nineties Paks Nuclear Power Plant-in collaboration with KFKI Atomic Energy Research Institute (KFKI AEKI)- is developing a system for determining the normal operation local power engineering safety factors. The system is based on a Monte Carlo sampling of the uncertain model input parameters. Additionally, the comparison of the calculation to the in-core measurements plays essential role for determining some important input parameters. By using new fuel types and the corresponding more recent detailed technological data, the applied method is being improved from time to time. Presently, the actually used and authorized engineering safety factors at Paks NPP are determined by using this method. In the paper, the system.s main properties are described (not going beyond the possible extent). The main points are as follows:-Mathematical definition of the engineering safety factor;-Sources of the uncertainties;-Input error propagation method constituting the basis of the system;-Flow-chart of the subsequent steps of the determination Finally, in the paper the engineering safety factors values of some selected parameters are presented as examples for demonstration of the capability of the method. (Authors)

Evaluation of the Pharmacy Safety Climate Questionnaire in European community pharmacies.

NARCIS (Netherlands)

Phipps, D.L.; Bie, J. de; Herborg, H.; Guerreiro, M.; Eickhoff, C.; Fernandes-Llimos, F.; Bouvy, M.L.; Rossing, C.; Mueller, U.; Ashcroft, D.M.

2012-01-01

Objective: To evaluate the internal reliability, factor structure and construct validity of the Pharmacy Safety Climate Questionnaire (PSCQ) when applied to a pan-European sample of community pharmacies. Design: A cross-sectional survey design was used. Setting: Community pharmacies in Denmark,

Safety culture evaluation and asset root cause analysis

International Nuclear Information System (INIS)

Okrent, D.; Xiong, Y.

1995-01-01

This paper examines the role of organizational and management factors in nuclear power plant safety through the use of operating experiences. The ASSET (Assessment of Safety Significant Events Team) reports of thirteen plants (total thirty events) have been analyzed in term of twenty organizational dimensions (factors) identified by Brookhaven National Laboratory and Pennsylvania State University. For three plants detailed results are reported in this paper. The results of thirteen plants are summarized in the form of a table. The study tends to confirm that organizational and management factors play an important role in plant safety. The twenty organizational dimensions and their definitions, in general, were adequate in this study. Formalization, Safety Culture, Technical Knowledge, Training, Roles-Responsibilities and Problem Identification appear to be key organizational factors which influence the safety of nuclear power plants studied

Establishment and prioritization of relevant factors to the safety of fuel cycle facilities non reactor through dynamics archetypes evaluation

International Nuclear Information System (INIS)

Sousa, Anna Leticia Barbosa de

2012-01-01

The present work aims to establish and prioritize factors that are important to the safety of nuclear fuel cycle facilities in order to model, analyze and design safety as a physical system, employing systemic models in an innovative way. This work takes into consideration the fact that models that use adaptations of methodologies for nuclear reactors will not properly work due to the specificities of fuel cycle facilities. Based on the fundamentals of the theory of systems, the four levels of system thinking, and the relationship of eight socio technical factors, a mental model has been developed for safety management in the nuclear fuel cycle context. From this conceptual model, safety archetypes were constructed in order to identify and highlight the processes of change and decision making that allow the system to migrate to a state of loss of safety. After that, stock and flow diagrams were created so that their behavior could be assessed by the system's dynamics. The results from the analysis using the model that simulates the dynamic behavior of the variables (socio technical factors) indicated, as expected, that the system's dynamics proved to be an appropriate and efficient tool for modeling fuel cycle safety as an emergent property. (author)

Preliminary safety evaluation (PSE) for Sodium Storage Facility at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Bowman, B.R.

1994-01-01

This evaluation was performed for the Sodium Storage Facility (SSF) which will be constructed at the Fast Flux Test Facility (FFTF) in the area adjacent to the South and West Dump Heat Exchanger (DHX) pits. The purpose of the facility is to allow unloading the sodium from the FFTF plant tanks and piping. The significant conclusion of this Preliminary Safety Evaluation (PSE) is that the only Safety Class 2 components are the four sodium storage tanks and their foundations. The building, because of its imminent risk to the tanks under an earthquake or high winds, will be Safety Class 3/2, which means the building has a Safety Class 3 function with the Safety Class 2 loads of seismic and wind factored into the design

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)

[Validation of a questionnaire to evaluate patient safety in clinical laboratories].

Science.gov (United States)

Giménez Marín, Ángeles; Rivas-Ruiz, Francisco

2012-01-01

The aim of this study was to prepare, pilot and validate a questionnaire to evaluate patient safety in the specific context of clinical laboratories. A specific questionnaire on patient safety in the laboratory, with 62 items grouped into six areas, was developed, taking into consideration the diverse human and laboratory contextual factors which may contribute to producing errors. A pilot study of 30 interviews was carried out, including validity and reliability analyses using principal components factor analysis and Cronbach's alpha. Subsequently, 240 questionnaires were sent to 21 hospitals, followed by a test-retest of 41 questionnaires with the definitive version. The sample analyzed was composed of 225 questionnaires (an overall response rate of 80%). Of the 62 items initially assessed, 17 were eliminated due to non-compliance with the criteria established before the principal components factor analysis was performed. For the 45 remaining items, 12 components were identified, with an cumulative variance of 69.5%. In seven of the 10 components with two or more items, Cronbach's alpha was higher than 0.7. The questionnaire items assessed in the test-retest were found to be stable. We present the first questionnaire with sufficiently proven validity and reliability for evaluating patient safety in the specific context of clinical laboratories. This questionnaire provides a useful instrument to perform a subsequent macrostudy of hospital clinical laboratories in Spain. The questionnaire can also be used to monitor and promote commitment to patient safety within the search for continuous quality improvement. Copyright © 2011 SESPAS. Published by Elsevier Espana. All rights reserved.

Squale: evaluation criteria of functioning safety

International Nuclear Information System (INIS)

Deswarte, Y.; Kaaniche, M.; Benoit, P.

1998-05-01

The SQUALE (security, safety and quality evaluation for dependable systems) project is part of the ACTS (advanced communications, technologies and services) European program. Its aim is to develop confidence evaluation criteria to test the functioning safety of systems. All industrial sectors that use critical applications (nuclear, railway, aerospace..) are concerned. SQUALE evaluation criteria differ from the classical evaluation methods: they are independent of the application domains and industrial sectors, they take into account the overall functioning safety attributes, and they can progressively change according to the level of severity required. In order to validate the approach and to refine the criteria, a first experiment is in progress with the METEOR automatic underground railway and another will be carried out on a telecommunication system developed by Bouygues company. (J.S.)

How to evaluate the effectiveness of safety assessment in the area of human factors?

International Nuclear Information System (INIS)

Rolina, G.; Moisdon, J.C.; Jeffroy, F.

2007-01-01

The Three Mile Island nuclear reactor accident in 1979 led to a new approach regarding safety that includes a better consideration of man and his activities. A few years later, with the set up of a group of specialists at Electricite de France and at the Institute for Radiological Protection and Nuclear Safety, a new player appeared at France's nuclear safety organisation: the assessment expert specialising in human factors (HF). The improvement of man-machine interfaces was one of the first projects undertaken by the HF experts, the majority of whom specialise in ergonomics. A review of the literature and analysis of the archives, revealed that the specialists' scope of investigation has since increased; so that organisation is also the subject of HF assessment. However, this area is not one of consensual or established knowledge; neither researchers nor specialists can agree on a model of safe organisation. What then can we say about effectiveness of HF assessment? How can we define the criteria of effectiveness of a safety assessment production system in this area? The question is the subject of original research based on collaboration between the scientific management centre (CGS) of the Ecole des Mines in Paris and the section for the study of human factors (SEFH) at IRSN. To address this question, the CGS team monitors some assessments to which SEFH contributes. In other words, it attends different meetings on framing, technical instruction, reporting, taking notes and collecting related documents (minutes of meetings,...). It carries out additional interviews with different parties involved in assessment in order to ascertain their point of view. A sample of five assessments was defined to cover a varied number of situations encountered by the team of HF experts. The type of facility, the operator and the subject concerned are some of the variables integrated for this choice

The PIANC Safety Factor System for Breakwaters

DEFF Research Database (Denmark)

Burcharth, H. F.

2000-01-01

The paper presents a summary of the recommendations for implementation of safety in breakwater designs given by the PIANC PTC IT Working Group No 12 on Analysis of Rubble Mound Breakwaters with Vertical and Inclined Concrete Walls. The working groups developed for the most important failure modes...... a system of partial safety factors which facilitate design to any target safety level....

Safety Evaluation of Roundabouts in Georgia

Science.gov (United States)

2018-02-28

Several previous studies have documented significant safety benefits of roundabouts in the United Sates. However, the safety benefits for a given roundabout may vary depending on factors such as the familiarity of the driving population to roundabout...

Analysis of factors influencing safety management for metro construction in China.

Science.gov (United States)

Yu, Q Z; Ding, L Y; Zhou, C; Luo, H B

2014-07-01

With the rapid development of urbanization in China, the number and size of metro construction projects are increasing quickly. At the same time, and increasing number of accidents in metro construction make it a disturbing focus of social attention. In order to improve safety management in metro construction, an investigation of the participants' perspectives on safety factors in China metro construction has been conducted to identify the key safety factors, and their ranking consistency among the main participants, including clients, consultants, designers, contractors and supervisors. The result of factor analysis indicates that there are five key factors which influence the safety of metro construction including safety attitude, construction site safety, government supervision, market restrictions and task unpredictability. In addition, ANOVA and Spearman rank correlation coefficients were performed to test the consistency of the means rating and the ranking of safety factors. The results indicated that the main participants have significant disagreement about the importance of safety factors on more than half of the items. Suggestions and recommendations on practical countermeasures to improve metro construction safety management in China are proposed. Copyright © 2013 Elsevier Ltd. All rights reserved.

Evaluation of Safety Culture Implementation and Socialization Results

International Nuclear Information System (INIS)

Situmorang, Johnny

2003-01-01

Evaluation of safety culture implementation and socialization results has been perform. Evaluation is carried out with specifying safety culture indicators, namely: Meeting between management and employee, system for incidents analysis, training activities related to improving safety, meeting with regulator, contractors, surveys on behavioural attitudes, and resources allocated to promote safety culture. Evaluation is based on observation and visiting the facilities to show the compliance indicator in term of good practices in the frame of safety culture implementation. For three facilities of research reactors, Kartini Yogyakarta, TRIGA Mark II Bandung and MPR-GAS Serpong, implementation of safety culture is considered good enough and progressive. Furthermore some indicator should be considered more intensive, for example the allocated resources, self assesment based on own questionnaire in the frame of improving the safety culture implementation. (author)

Safety evaluation of large ventilation networks

International Nuclear Information System (INIS)

Barrocas, M.; Pruchon, P.; Robin, J.P.; Rouyer, J.L.; Salmon, P.

1981-01-01

For large ventilation networks, it is necessary to make a safety evaluation of their responses to perturbations such as blower failure, unexpected transfers, local pressurization. This evaluation is not easy to perform because of the many interrelationships between the different parts of the networks, interrelationships coming from the circulations of workers and matetials between cells and rooms and from the usefulness of air transfers through zones of different classifications. This evaluation is all the more necessary since new imperatives in energy savings push for minimizing the air flows, which tends to render the network more sensitive to perturbations. A program to evaluate safety has been developed by the Service de Protection Technique in cooperation with operators and designers of big nuclear facilities and the first applications presented here show the weak points of the installation studied from the safety view point

Patient Safety Culture Survey in Pediatric Complex Care Settings: A Factor Analysis.

Science.gov (United States)

Hessels, Amanda J; Murray, Meghan; Cohen, Bevin; Larson, Elaine L

2017-04-19

Children with complex medical needs are increasing in number and demanding the services of pediatric long-term care facilities (pLTC), which require a focus on patient safety culture (PSC). However, no tool to measure PSC has been tested in this unique hybrid acute care-residential setting. The objective of this study was to evaluate the psychometric properties of the Nursing Home Survey on Patient Safety Culture tool slightly modified for use in the pLTC setting. Factor analyses were performed on data collected from 239 staff at 3 pLTC in 2012. Items were screened by principal axis factoring, and the original structure was tested using confirmatory factor analysis. Exploratory factor analysis was conducted to identify the best model fit for the pLTC data, and factor reliability was assessed by Cronbach alpha. The extracted, rotated factor solution suggested items in 4 (staffing, nonpunitive response to mistakes, communication openness, and organizational learning) of the original 12 dimensions may not be a good fit for this population. Nevertheless, in the pLTC setting, both the original and the modified factor solutions demonstrated similar reliabilities to the published consistencies of the survey when tested in adult nursing homes and the items factored nearly identically as theorized. This study demonstrates that the Nursing Home Survey on Patient Safety Culture with minimal modification may be an appropriate instrument to measure PSC in pLTC settings. Additional psychometric testing is recommended to further validate the use of this instrument in this setting, including examining the relationship to safety outcomes. Increased use will yield data for benchmarking purposes across these specialized settings to inform frontline workers and organizational leaders of areas of strength and opportunity for improvement.

Status of Nuclear Safety evaluation in China

International Nuclear Information System (INIS)

Tian Jiashu

1999-01-01

Chinese nuclear safety management and control follows international practice, the regulations are mainly from IAEA with the Chinese condition. The regulatory body is National Nuclear Safety Administration (NNSA). The nuclear safety management, surveillance, safety review and evaluation are guided by NNSA with technical support by several units. Beijing Review Center of Nuclear Safety is one of these units, which was founded in 1987 within Beijing Institute of nuclear Engineering (BINE), co-directed by NNSA and BINE, it is the first technical support team to NNSA. Most of the safety reviews and evaluations of Chinese nuclear installations has been finished by this unit. It is described briefly in this paper that the NNSA's main function and organization, regulations on the nuclear safety, procedure of application and issuing of license, the main activities performed by Beijing Review Center of Nuclear Safety, the situation of severe accident analyses in China, etc. (author)

Plutonium Finishing Plant safety evaluation report

International Nuclear Information System (INIS)

1995-01-01

The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE's independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91

Safety evaluation of Tokai reprocessing plant (TRP). Report of safety evaluation of Tokai reprocessing plant

International Nuclear Information System (INIS)

Yamauchi, Takamichi; Maki, Akira; Nojiri, Ichiro

1999-02-01

The fire and explosion incident of the bituminization facility happened in March 1997 although JNC had taken enough care of the safety of TRP. JNC reflected on it and decided to evaluate the safety of TRP voluntarily. This evaluation has included five activities, that is, (1) confirmation of the structure and organization of TRP, (2) research of the data for operation, radiation and maintenance of TRP, (3) research of reflection of the accidents and troubles which have happened at the past, (4) evaluation on the prevention system, (5) evaluation on the mitigation system. We publish this report to contribute to inheritance of accumulated knowledge and techniques from generation to generation, and remind us of lesson from the fire and explosion incident of the bituminization. (author)

mathematical models for prediction of safety factors for a simply

African Journals Online (AJOL)

HOD

Keywords: reliability, code calibration, load factor, safety factor, design, steel beam. 1. INTRODUCTION ... safety factors for the design of a simply supported steel beam using regression .... 5 design criteria for a solid timber portal frame.

SafetyNet. Human factors safety training on the Internet

DEFF Research Database (Denmark)

Hauland, G.; Pedrali, M.

2002-01-01

This report describes user requirements to an Internet based distance learning system of human factors training, i.e. the SafetyNet prototype, within the aviation (pilots and air traffic control), maritime and medical domains. User requirements totraining have been elicited through 19 semi...

Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S.; Lee, M. S.; Kim, T. H.

2016-01-01

This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified

Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S. [KINS, Daejeon (Korea, Republic of); Lee, M. S.; Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

2016-05-15

This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified.

Evaluation of Pre-marketing Factors to Predict Post-marketing Boxed Warnings and Safety Withdrawals.

Science.gov (United States)

Schick, Andreas; Miller, Kathleen L; Lanthier, Michael; Dal Pan, Gerald; Nardinelli, Clark

2017-06-01

An important goal in drug regulation is understanding serious safety issues with new drugs as soon as possible. Achieving this goal requires us to understand whether information provided during the Food and Drug Administration (FDA) drug review can predict serious safety issues that are usually identified after the product is approved. However, research on this topic remains understudied. In this paper, we examine whether any pre-marketing drug characteristics are associated with serious post-marketing safety actions. We study this question using an internal FDA database containing every new small molecule drug submitted to the FDA's Center for Drug Evaluation and Research (CDER) on or after November 21, 1997, and approved and commercially launched before December 31, 2009. Serious post-marketing safety actions include whether these drugs ever experienced either a post-marketing boxed warning or a withdrawal from the market due to safety concerns. A random effects logistic regression model was used to test whether any pre-marketing characteristics were associated with either post-marketing safety action. A total of 219 new molecular entities were analyzed. Among these drugs, 11 experienced a safety withdrawal and 30 received boxed warnings by July 31, 2016. Contrary to prevailing hypotheses, we find that neither clinical trial sample sizes nor review time windows are associated with the addition of a post-marketing boxed warning or safety withdrawal. However, we do find that new drugs approved with either a boxed warning or priority review are more likely to experience post-marketing boxed warnings. Furthermore, drugs approved with boxed warnings tend to receive post-marketing boxed warnings resulting from new safety information that are unrelated to the original warning. Drugs approved with a boxed warning are 3.88 times more likely to receive a post-marketing boxed warning, while drugs approved with a priority review are 3.51 times more likely to receive a post

Philosophy of safety evaluation on fast breeder reactor

International Nuclear Information System (INIS)

1981-01-01

This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

Failure and factors of safety in piping system design

International Nuclear Information System (INIS)

Antaki, G.A.

1993-01-01

An important body of test and performance data on the behavior of piping systems has led to an ongoing reassessment of the code stress allowables and their safety margin. The codes stress allowables, and their factors of safety, are developed from limits on the incipient yield (for ductile materials), or incipient rupture (for brittle materials), of a test specimen loaded in simple tension. In this paper, we examine the failure theories introduced in the B31 and ASME III codes for piping and their inherent approximations compared to textbook failure theories. We summarize the evolution of factors of safety in ASME and B31 and point out that, for piping systems, it is appropriate to reconsider the concept and definition of factors of safety

TAPS safety evaluation criteria for reload fueling

International Nuclear Information System (INIS)

Mahendra Nath; Veeraraghavan, N.

1976-01-01

To improve operating performance of Tarapur reactors, several proposals are under consideration such as core expansion, change-over to an improved fuel design with lower heat rating, extension of fuel cycle lengths etc., which have a bearing on overall plant operating characteristics and reactor safety. For evaluating safety implications of the various proposals, it is necessary to formulate safety evaluation criteria for reload fuelling. Salient features of these criteria are discussed. (author)

Experiment to evaluate software safety

International Nuclear Information System (INIS)

Soubies, B.; Henry, J.Y.

1994-01-01

The process of licensing nuclear power plants for operation consists of mandatory steps featuring detailed examination of the instrumentation and control system by the safety authorities, including softwares. The criticality of these softwares obliges the manufacturer to develop in accordance with the IEC 880 standard 'Computer software in nuclear power plant safety systems' issued by the International Electronic Commission. The evaluation approach, a two-stage assessment is described in detail. In this context, the IPSN (Institute of Protection and Nuclear Safety), the technical support body of the safety authority uses the MALPAS tool to analyse the quality of the programs. (R.P.). 4 refs

Lyapunov-based distributed control of the safety-factor profile in a tokamak plasma

International Nuclear Information System (INIS)

Bribiesca Argomedo, Federico; Witrant, Emmanuel; Prieur, Christophe; Brémond, Sylvain; Nouailletas, Rémy; Artaud, Jean-François

2013-01-01

A real-time model-based controller is developed for the tracking of the distributed safety-factor profile in a tokamak plasma. Using relevant physical models and simplifying assumptions, theoretical stability and robustness guarantees were obtained using a Lyapunov function. This approach considers the couplings between the poloidal flux diffusion equation, the time-varying temperature profiles and an independent total plasma current control. The actuator chosen for the safety-factor profile tracking is the lower hybrid current drive, although the results presented can be easily extended to any non-inductive current source. The performance and robustness of the proposed control law is evaluated with a physics-oriented simulation code on Tore Supra experimental test cases. (paper)

CRITICALITY SAFETY LIMIT EVALUATION PROGRAM (CSLEP's) AND QUICK SCREENS: ANSWERS TO EXPEDITED PROCESSING LEGACY CRITICALITY SAFETY LIMITS AND EVALUATIONS

International Nuclear Information System (INIS)

TOFFER, H.

2006-01-01

Since the end of the cold war, the need for operating weapons production facilities has faded. Criticality Safety Limits and controls supporting production modes in these facilities became outdated and furthermore lacked the procedure based rigor dictated by present day requirements. In the past, in many instances, the formalism of present day criticality safety evaluations was not applied. Some of the safety evaluations amounted to a paragraph in a notebook with no safety basis and questionable arguments with respect to double contingency criteria. When material stabilization, clean out, and deactivation activities commenced, large numbers of these older criticality safety evaluations were uncovered with limits and controls backed up by tenuous arguments. A dilemma developed: on the one hand, cleanup activities were placed on very aggressive schedules; on the other hand, a highly structured approach to limits development was required and applied to the cleanup operations. Some creative approaches were needed to cope with the limits development process

Centrifuge model test of rock slope failure caused by seismic excitation. Applicability to the stability evaluation method of safety factors against sliding

International Nuclear Information System (INIS)

Ishimaru, Makoto; Kawai, Tadashi

2010-01-01

The purposes of this study are to analyze dynamic failure characteristics of slopes in discontinuous rock mass with brittle fracture by centrifuge model tests and to study applicability to the equivalent linear analysis against dynamic sliding failure of rock slopes. We conducted centrifuge model test using a dip slope model with discontinuities imitated by Teflon sheets. The centrifugal acceleration was 30G, and the acceleration amplitudes of input sin waves were increased gradually at every step. The test results were compared with safety factors of the sliding surface based on the equivalent linear analysis. The following results were obtained: (1) The slope model collapsed when it was excited by the sine wave of 350gal, which was converted to real field scale. (2) Artificial discontinuities considerably affected the collapse, and the type of collapse was plane failure. (3) From response displacement records measured at the slope model, the failure around toe of the slope model probably caused the collapse. (4) The evaluation of safety factors against sliding based on the equivalent linear analysis were conservative compared with the experimental results. (author)

Safety evaluation of food flavorings

International Nuclear Information System (INIS)

Schrankel, Kenneth R.

2004-01-01

Food flavorings are an essential element in foods. Flavorings are a unique class of food ingredients and excluded from the legislative definition of a food additive because they are regulated by flavor legislation and not food additive legislation. Flavoring ingredients naturally present in foods, have simple chemical structures, low toxicity, and are used in very low levels in foods and beverages resulting in very low levels of human exposure or consumption. Today, the overwhelming regulatory trend is a positive list of flavoring substances, e.g. substances not listed are prohibited. Flavoring substances are added to the list following a safety evaluation based on the conditions of intended use by qualified experts. The basic principles for assessing the safety of flavoring ingredients will be discussed with emphasis on the safety evaluation of flavoring ingredients by the Food and Agriculture Organization (FAO) and World Health Organization (WHO) Joint Expert Committee on Food Additives (JECFA) and the US Flavor and Extract Manufacturers Expert Panel (FEXPAN). The main components of the JECFA evaluation process include chemical structure, human intake (exposure), metabolism to innocuous or harmless substances, and toxicity concerns consistent with JECFA principles. The Flavor and Extract Manufacturers Association (FEMA) evaluation is very similar to the JECFA procedure. Both the JECFA and FEMA evaluation procedures are widely recognized and the results are accepted by many countries. This implies that there is no need for developing countries to conduct their own toxicological assessment of flavoring ingredients unless it is an unique ingredient in one country, but it is helpful to survey intake or exposure assessment. The global safety program established by the International Organization of Flavor Industry (IOFI) resulting in one worldwide open positive list of flavoring substances will be reviewed

Road safety effects of porous asphalt: a systematic review of evaluation studies

DEFF Research Database (Denmark)

Elvik, R.; Greibe, Poul

2005-01-01

of eighteen estimates of the effect of porous asphalt on accident rates. No clear effect on road safety of porous asphalt was found. All summary estimates of effect indicated very small changes in accident rates and very few were statistically significant at conventional levels. Studies that have evaluated...... of these changes in risk factors on accident occurrence cannot be predicted. On the whole, the research that has been reported so far regarding road safety effects of porous asphalt is inconclusive. The studies are not of high quality and the findings are inconsistent.......This paper presents a systematic review of studies that have evaluated the effects on road safety of porous asphalt. Porous asphalt is widely used on motorways in Europe, mainly in order to reduce traffic noise and increase road capacity. A meta-analysis was made of six studies, containing a total...

76 FR 35130 - Pipeline Safety: Control Room Management/Human Factors

Science.gov (United States)

2011-06-16

...: Control Room Management/Human Factors AGENCY: Pipeline and Hazardous Materials Safety Administration... the Control Room Management/Human Factors regulations in order to realize the safety benefits sooner... FR 5536). By this amendment to the Control Room Management/Human Factors (CRM) rule, an operator must...

Safety evaluation of advance street name signs

Science.gov (United States)

2009-06-01

The Federal Highway Administration (FHWA) organized a pooled fund study of 26 States to evaluate low-cost safety strategies as part of its strategic highway safety effort. The objective of the pooled fund study was to estimate the safety effectivenes...

Human factors engineering design review acceptance criteria for the safety parameter display

International Nuclear Information System (INIS)

McGevna, V.; Peterson, L.R.

1981-01-01

This report contains human factors engineering design review acceptance criteria developed by the Human Factors Engineering Branch (HFEB) of the Nuclear Regulatory Commission (NRC) to use in evaluating designs of the Safety Parameter Display System (SPDS). These criteria were developed in response to the functional design criteria for the SPDS defined in NUREG-0696, Functional Criteria for Emergency Response Facilities. The purpose of this report is to identify design review acceptance criteria for the SPDS installed in the control room of a nuclear power plant. Use of computer driven cathode ray tube (CRT) displays is anticipated. General acceptance criteria for displays of plant safety status information by the SPDS are developed. In addition, specific SPDS review criteria corresponding to the SPDS functional criteria specified in NUREG-0696 are established

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

International Nuclear Information System (INIS)

Li Tianshu

2013-01-01

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

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.

On applying safety archetypes to the Fukushima accident to identify nonlinear influencing factors

Energy Technology Data Exchange (ETDEWEB)

Sousa, A.L., E-mail: alsousa@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Ribeiro, A.C.O., E-mail: antonio.ribeiro@bayer.com [Bayer Crop Science Brasil S.A., Belford Roxo, RJ (Brazil); Duarte, J.P., E-mail: julianapduarte@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Escola Politecnica. Departamento de Engenharia Nuclear; Frutuoso e Melo, P.F., E-mail: frutuoso@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COOPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear

2013-07-01

Nuclear power plants are typically characterized as high reliable organizations. In other words, they are organizations defined as relatively error free over a long period of time. Another relevant characteristic of the nuclear industry is that safety efforts are credited to design. However, major accidents, like the Fukushima accident, have shown that new tools are needed to identify latent deficiencies and help improve their safety level. Safety archetypes proposed elsewhere (e. g., safety issues stalled in the face of technological advances and eroding safety) consonant with International Atomic Energy Agency (IAEA) efforts are used to examine different aspects of accidents in a systemic perspective of the interaction between individuals, technology and organizational factors. Safety archetypes can help consider nonlinear interactions. Effects are rarely proportional to causes and what happens locally in a system (near the current operating point) often does not apply to distant regions (other system states), so that one has to consider the so-called nonlinear interactions. This is the case, for instance, with human probability failure estimates and safety level identification. In this paper, we discuss the Fukushima accident in order to show how archetypes can highlight nonlinear interactions of factors that influenced it and how to maintain safety levels in order to prevent other accidents. The initial evaluation of the set of archetypes suggested in the literature showed that at least four of them are applicable to the Fukushima accident, as is inferred from official reports on the accident. These are: complacency (that is, the effects of complacency on safety), decreased safety awareness, fixing on symptoms and not the real causes and eroding safety. (author)

On applying safety archetypes to the Fukushima accident to identify nonlinear influencing factors

International Nuclear Information System (INIS)

Sousa, A.L.; Ribeiro, A.C.O.; Duarte, J.P.; Frutuoso e Melo, P.F.

2013-01-01

Nuclear power plants are typically characterized as high reliable organizations. In other words, they are organizations defined as relatively error free over a long period of time. Another relevant characteristic of the nuclear industry is that safety efforts are credited to design. However, major accidents, like the Fukushima accident, have shown that new tools are needed to identify latent deficiencies and help improve their safety level. Safety archetypes proposed elsewhere (e. g., safety issues stalled in the face of technological advances and eroding safety) consonant with International Atomic Energy Agency (IAEA) efforts are used to examine different aspects of accidents in a systemic perspective of the interaction between individuals, technology and organizational factors. Safety archetypes can help consider nonlinear interactions. Effects are rarely proportional to causes and what happens locally in a system (near the current operating point) often does not apply to distant regions (other system states), so that one has to consider the so-called nonlinear interactions. This is the case, for instance, with human probability failure estimates and safety level identification. In this paper, we discuss the Fukushima accident in order to show how archetypes can highlight nonlinear interactions of factors that influenced it and how to maintain safety levels in order to prevent other accidents. The initial evaluation of the set of archetypes suggested in the literature showed that at least four of them are applicable to the Fukushima accident, as is inferred from official reports on the accident. These are: complacency (that is, the effects of complacency on safety), decreased safety awareness, fixing on symptoms and not the real causes and eroding safety. (author)

The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

International Nuclear Information System (INIS)

Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

2006-06-01

Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

Safety indicators as a tool for operational safety evaluation of nuclear power plants

International Nuclear Information System (INIS)

Araujo, Jefferson Borges; Melo, Paulo Fernando Ferreira Frutuoso e; Schirru, Roberto

2009-01-01

Performance indicators have found a wide use in the conventional and nuclear industries. For the conventional industry, the goal is to optimize production, reducing loss of time with accidents, human error and equipment downtimes. In the nuclear industry, nuclear safety is an additional goal. This paper presents a general methodology to the establishment, selection and use of safety indicators for a two loop PWR plant, as Angra 1. The use of performance indicators is not new. The NRC has its own methodology and the IAEA presents methodology suggestions, but there is no detailed documentation about indicators selection, criteria and bases used. Additionally, only the NRC methodology performs a limited integrated evaluation. The study performed identifies areas considered critical for the plant operational safety. For each of these areas, strategic sub-areas are defined. For each strategic sub-area, specific safety indicators are defined. These proposed Safety Indicators are based on the contribution to risk considering a quantitative risk analysis. For each safety indicator, a goal, a bounded interval and proper bases are developed, to allow for a clear and comprehensive individual behavior evaluation. On the establishment of the intervals and boundaries, a probabilistic safety study, operational experience, international and national standards and technical specifications were used. Additionally, an integrated evaluation of the indicators, using expert systems, was done to obtain an overview of the plant general safety. This evaluation uses well-defined and clear rules and weights for each indicator to be considered. These rules were implemented by means of a computational language, on a friendly interface, so that it is possible to obtain a quick response about operational safety. This methodology can be used to identify situations where the plant safety is challenged, by giving a general overview of the plant operational condition. Additionally, this study can

Integrated approach to knowledge acquisition and safety management of complex plants with emphasis on human factors

International Nuclear Information System (INIS)

Kosmowski, K.T.

1998-01-01

In this paper an integrated approach to the knowledge acquisition and safety management of complex industrial plants is proposed and outlined. The plant is considered within a man-technology-environment (MTE) system. The knowledge acquisition is aimed at the consequent reliability evaluation of human factor and probabilistic modeling of the plant. Properly structured initial knowledge is updated in life-time of the plant. The data and knowledge concerning the topology of safety related systems and their functions are created in a graphical CAD system and are object oriented. Safety oriented monitoring of the plant includes abnormal situations due to external and internal disturbances, failures of hard/software components and failures of human factor. The operation and safety related evidence is accumulated in special data bases. Data/knowledge bases are designed in such a way to support effectively the reliability and safety management of the plant. (author)

Partial Safety Factors and Target Reliability Level in Danish Structural Codes

DEFF Research Database (Denmark)

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

2001-01-01

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

Evaluation of the Finnish nuclear safety research program 'SAFIR2010'

International Nuclear Information System (INIS)

2010-01-01

A panel of three members has been asked by the Ministry of Employment and the Economy (MEE) to evaluate SAFIR2010, the Finnish research program on nuclear power plant safety. The program was established for the period 2007-2010 to help maintain expertise in nuclear safety, to integrate young people into the research in order to help assure the future availability of expertise, and to support international collaborations. The program is directed by a Steering Group, appointed by MEE, with representatives from all organizations involved with nuclear safety in Finland. SAFIR2010 has consisted of approximately 30 projects from year to year that fall into eight subject areas: 1. Organization and human factors 2. Automation and control room 3. Fuel and reactor physics 4. Thermal hydraulics 5. Severe accidents 6. Structural safety of reactor circuit 7. Construction safety 8. Probabilistic safety analysis (PSA) For each of these areas there are Reference Groups that provide oversight of the projects within their jurisdiction. The panel carried out its evaluation by reviewing copies of relevant documents and, during a one-week period 17-22 January 2010, meeting with key individuals. The results of the panel are provided as general conclusions, responses to questions posed by MEE, challenges and recommendations and comments on specific projects in each subject area. The general conclusions reflect the panel's view that SAFIR2010 is meeting its objectives and carrying out quality research. The questions addressed are: (a.) Are the achieved results in balance with the funding? Are the results exploited efficiently in practice? (b.) How well does the expertise cover the field? Is the entire SAFIR2010 programme balanced to all different fields in nuclear safety? Does it raise efficiently new experts? (c.) Have the 2006 evaluation results been implemented successfully into SAFIR2010 program? (d.) Challenges and recommendations. In general the panel was very positive about SAFIR

The role of the regulator in promoting and evaluating safety culture. Operating experience feedback programme approach

International Nuclear Information System (INIS)

Perez, S.

2002-01-01

Promoting and Evaluating Safety Culture (S.C.) in Operating Organizations must be one of the main Nuclear Regulator goals to achieve. This can be possible only if each and every one of the regulatory activities inherently involves S.C. It can be seen throughout attitudes, values, uses and practices in both individuals and the whole regulatory organization. One among all the regulatory tools commonly used by regulators to promote and evaluate the commitment of the licensees with safety culture as a whole involves organizational factors and particular attention is directed to the operating organization. This entailed a wide range of activities, including all those related with management of safety performance. Operating Experience Feedback Programme as a tool to enhance safety operation is particularly useful for regulators in the evaluation of the role of S.C. in operating organization. Safety Culture is recognized as a subset of the wider Organizational Culture. Practices that improve organizational effectiveness can also contribute to enhance safety. An effective event investigation methodology is a specific practice, which contributes to a healthy Safety Culture. (author)

Approach to uncertainty evaluation for safety analysis

International Nuclear Information System (INIS)

Ogura, Katsunori

2005-01-01

Nuclear power plant safety used to be verified and confirmed through accident simulations using computer codes generally because it is very difficult to perform integrated experiments or tests for the verification and validation of the plant safety due to radioactive consequence, cost, and scaling to the actual plant. Traditionally the plant safety had been secured owing to the sufficient safety margin through the conservative assumptions and models to be applied to those simulations. Meanwhile the best-estimate analysis based on the realistic assumptions and models in support of the accumulated insights could be performed recently, inducing the reduction of safety margin in the analysis results and the increase of necessity to evaluate the reliability or uncertainty of the analysis results. This paper introduces an approach to evaluate the uncertainty of accident simulation and its results. (Note: This research had been done not in the Japan Nuclear Energy Safety Organization but in the Tokyo Institute of Technology.) (author)

Factors Contribute to Safety Culture in the Manufacturing Industry in Malaysia

OpenAIRE

Ong Choon Hee

2014-01-01

The purpose of this paper is to explain the role of safety culture in the manufacturing industry in Malaysia and identify factors contribute to safety culture. It is suggested in this study that leadership support, management commitment and safety management system are important factors that contribute to safety culture. This study also provides theoretical implications to guide future research and offers practical implications to the managers in the development of safety culture. Given that ...

Criticality Safety Evaluation of Hanford Tank Farms Facility

Energy Technology Data Exchange (ETDEWEB)

WEISS, E.V.

2000-12-15

Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste.

Criticality Safety Evaluation of Hanford Tank Farms Facility

International Nuclear Information System (INIS)

WEISS, E.V.

2000-01-01

Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste

Safety evaluation of a hydrogen fueled transit bus

Energy Technology Data Exchange (ETDEWEB)

Coutts, D.A.; Thomas, J.K.; Hovis, G.L.; Wu, T.T. [Westinghouse Savannah River Co., Aiken, SC (United States)

1997-12-31

Hydrogen fueled vehicle demonstration projects must satisfy management and regulator safety expectations. This is often accomplished using hazard and safety analyses. Such an analysis has been completed to evaluate the safety of the H2Fuel bus to be operated in Augusta, Georgia. The evaluation methods and criteria used reflect the Department of Energy`s graded approach for qualifying and documenting nuclear and chemical facility safety. The work focused on the storage and distribution of hydrogen as the bus motor fuel with emphases on the technical and operational aspects of using metal hydride beds to store hydrogen. The safety evaluation demonstrated that the operation of the H2Fuel bus represents a moderate risk. This is the same risk level determined for operation of conventionally powered transit buses in the United States. By the same criteria, private passenger automobile travel in the United States is considered a high risk. The evaluation also identified several design and operational modifications that resulted in improved safety, operability, and reliability. The hazard assessment methodology used in this project has widespread applicability to other innovative operations and systems, and the techniques can serve as a template for other similar projects.

Evaluation of repository safety

Energy Technology Data Exchange (ETDEWEB)

Sagar, B.; Patrick, W.; Dasgupta, B.; Mohanty, S. [Center for Nuclear Waste Regulatory Analyses, San Antonio (United States)

2002-07-01

The United States high-level waste program requires evaluation of radiological safety during two distinct time intervals. The first interval, commonly referred to as the preclosure period, deals with receipt of waste at the site, transfer into disposal containers, if needed, emplacement in the underground openings, monitoring and maintenance activities, backfill and closure of the underground openings, and decontamination and decommissioning of the surface facilities of the geologic repository. The preclosure period may extend from a few tens of years to as long as a few hundred of years, depending on repository design and societal norms regarding a final decision to permanently seal the repository. During the preclosure or operational period, performance confirmation studies are conducted to provide a basis for updating and reevaluating estimates of postclosure performance and, finally, to provide a basis for a closure decision. The postclosure period during which expected repository performance must meet certain standards may range from ten thousands years, as it does in the United States, to millions of years, as it does in some European nations. Waste handling operations in the preclosure period are to be evaluated in relation to their potential effect on workers, members of general public, and the general environment. During this period, releases of radioactivity are to be monitored and appropriate actions taken whenever established limits are approached or exceeded. Preclosure safety is highly dependent on facility design, operational hardware and automated systems, operational sequences, and reliability of humans involved in operations. Preclosure safety analyses conducted before operations begin play a major role in the design process, selection of equipment, and development of operational procedures. Because of the complexity, duration, and spatial scales of the operations, analyses are conducted using mathematical models implemented in computer codes

Evaluation of repository safety

International Nuclear Information System (INIS)

Sagar, B.; Patrick, W.; Dasgupta, B.; Mohanty, S.

2002-01-01

The United States high-level waste program requires evaluation of radiological safety during two distinct time intervals. The first interval, commonly referred to as the preclosure period, deals with receipt of waste at the site, transfer into disposal containers, if needed, emplacement in the underground openings, monitoring and maintenance activities, backfill and closure of the underground openings, and decontamination and decommissioning of the surface facilities of the geologic repository. The preclosure period may extend from a few tens of years to as long as a few hundred of years, depending on repository design and societal norms regarding a final decision to permanently seal the repository. During the preclosure or operational period, performance confirmation studies are conducted to provide a basis for updating and reevaluating estimates of postclosure performance and, finally, to provide a basis for a closure decision. The postclosure period during which expected repository performance must meet certain standards may range from ten thousands years, as it does in the United States, to millions of years, as it does in some European nations. Waste handling operations in the preclosure period are to be evaluated in relation to their potential effect on workers, members of general public, and the general environment. During this period, releases of radioactivity are to be monitored and appropriate actions taken whenever established limits are approached or exceeded. Preclosure safety is highly dependent on facility design, operational hardware and automated systems, operational sequences, and reliability of humans involved in operations. Preclosure safety analyses conducted before operations begin play a major role in the design process, selection of equipment, and development of operational procedures. Because of the complexity, duration, and spatial scales of the operations, analyses are conducted using mathematical models implemented in computer codes

Simplified probabilistic approach to determine safety factors in deterministic flaw acceptance criteria

International Nuclear Information System (INIS)

Barthelet, B.; Ardillon, E.

1997-01-01

The flaw acceptance rules in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a reliable method of evaluating the safety margins and the integrity of components led Electricite de France to launch a study to link safety factors with requested reliability. A simplified analytical probabilistic approach is developed to analyse the failure risk in Fracture Mechanics. Assuming lognormal distributions of the main random variables, it is possible considering a simple Linear Elastic Fracture Mechanics model, to determine the failure probability as a function of mean values and logarithmic standard deviations. The 'design' failure point can be analytically calculated. Partial safety factors on the main variables (stress, crack size, material toughness) are obtained in relation with reliability target values. The approach is generalized to elastic plastic Fracture Mechanics (piping) by fitting J as a power law function of stress, crack size and yield strength. The simplified approach is validated by detailed probabilistic computations with PROBAN computer program. Assuming reasonable coefficients of variations (logarithmic standard deviations), the method helps to calibrate safety factors for different components taking into account reliability target values in normal, emergency and faulted conditions. Statistical data for the mechanical properties of the main basic materials complement the study. The work involves laboratory results and manufacture data. The results of this study are discussed within a working group of the French in service inspection code RSE-M. (authors)

Safety climate and safety behaviors in the construction industry: The importance of co-workers commitment to safety.

Science.gov (United States)

Schwatka, Natalie V; Rosecrance, John C

2016-06-16

There is growing empirical evidence that as safety climate improves work site safety practice improve. Safety climate is often measured by asking workers about their perceptions of management commitment to safety. However, it is less common to include perceptions of their co-workers commitment to safety. While the involvement of management in safety is essential, working with co-workers who value and prioritize safety may be just as important. To evaluate a concept of safety climate that focuses on top management, supervisors and co-workers commitment to safety, which is relatively new and untested in the United States construction industry. Survey data was collected from a cohort of 300 unionized construction workers in the United States. The significance of direct and indirect (mediation) effects among safety climate and safety behavior factors were evaluated via structural equation modeling. Results indicated that safety climate was associated with safety behaviors on the job. More specifically, perceptions of co-workers commitment to safety was a mediator between both management commitment to safety climate factors and safety behaviors. These results support workplace health and safety interventions that build and sustain safety climate and a commitment to safety amongst work teams.

A Guidebook for Evaluating Organizations in the Nuclear Industry - an example of safety culture evaluation

International Nuclear Information System (INIS)

Oedewald, Pia; Pietikaeinen, Elina; Reiman, Teemu

2011-06-01

Organizations in the nuclear industry need to maintain an overview on their vulnerabilities and strengths with respect to safety. Systematic periodical self assessments are necessary to achieve this overview. This guidebook provides suggestions and examples to assist power companies but also external evaluators and regulators in carrying out organizational evaluations. Organizational evaluation process is divided into five main steps. These are: 1) planning the evaluation framework and the practicalities of the evaluation process, 2) selecting data collection methods and conducting the data acquisition, 3) structuring and analysing the data, 4) interpreting the findings and 5) reporting the evaluation results with possible recommendations. The guidebook emphasises the importance of a solid background framework when dealing with multifaceted phenomena like organisational activities and system safety. The validity and credibility of the evaluation stem largely from the evaluation team's ability to crystallize what they mean by organization and safety when they conduct organisational safety evaluations - and thus, what are the criteria for the evaluation. Another important and often under-considered phase in organizational evaluation is interpretation of the findings. In this guidebook a safety culture evaluation in a Nordic nuclear power plant is presented as an example of organizational evaluation. With the help of the example, challenges of each step in the organizational evaluation process are described. Suggestions for dealing with them are presented. In the case example, the DISC (Design for Integrated Safety culture) model is used as the evaluation framework. The DISC model describes the criteria for a good safety culture and the organizational functions necessary to develop a good safety culture in the organization

Application of the AHP method to analyze the significance of the factors affecting road traffic safety

Directory of Open Access Journals (Sweden)

Justyna SORDYL

2015-06-01

Full Text Available Over the past twenty years, the number of vehicles registered in Poland has grown rapidly. At the same time, a relatively small increase in the length of the road network has been observed. As a result of the limited capacity of available infrastructure, it leads to significant congestion and to increase of the probability of road accidents. The overall level of road safety depends on many factors - the behavior of road users, infrastructure solutions and the development of automotive technology. Thus the detailed assessment of the importance of individual elements determining road safety is difficult. The starting point is to organize the factors by grouping them into categories which are components of the DVE system (driver - vehicle - environment. In this work, to analyze the importance of individual factors affecting road safety, the use of analytic hierarchy process method (AHP was proposed. It is one of the multi-criteria methods which allows us to perform hierarchical analysis of the decision process, by means of experts’ opinions. Usage of AHP method enabled us to evaluate and rank the factors affecting road safety. This work attempts to link the statistical data and surveys in significance analysis of the elements determining road safety.

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

Parameters Evaluation of PLC Dependability and Safety

Directory of Open Access Journals (Sweden)

Juraj Zdansky

2006-01-01

Full Text Available This paper is focused on evaluation of dependability and safety parameters of PLC (Programmable Logic Controller. Achievement of requested level of these parameters is an application assumption for using PLC in control of safety critical processes. Evaluation of these parameters can be made on the base of suitable model and it can be influenced by system architecture when necessary.

The PSA of safety-critical digital I and C system: the determination of important factors and sensitivity analysis

International Nuclear Information System (INIS)

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

2002-01-01

This report is prepared to suggest a practical Probabilistic Safety Assessment (PSA) methodology of safety-critical digital instrumentation and control (I and C) systems. Even though conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it because the result of probabilistic safety assessment plays very important role in proving the safety of a designed system. Microprocessors and software technologies make the digital system very complex and hard to analyze the safety of their applications. The aim of this is: (1) To summarize the factors which should be represented by the model for probabilistic safety assessment and to propose a standpoint of evaluation for digital systems. (2) To quantitatively presents the results of a mathematical case study which examines the analysis framework of the safety of digital systems in the context of the PSA. (3) To show the results of a sensitivity study for some critical factors

Safety evaluation status report for the prototype license application safety analysis report

International Nuclear Information System (INIS)

1989-07-01

The US Nuclear Regulatory Commission (NRC) staff and consultants reviewed a Prototype License Application Safety Analysis Report (PLASAR) submitted by the US Department of Energy (DOE) for the earth-mounded concrete bunker (EMCB) alternative method of low-level radioactive waste disposal. The NRC reviewers relied extensively on the Standard Review Plan (SRP), Rev.1 (NUREG-1200), to evaluate the acceptability of the information provided in the EMCB PLASAR. The NRC staff selected certain review areas in the PLASAR for development of safety evaluation report input to provide examples of safety assessments that are necessary as part of a licensing review. Because of the fictitious nature of the assumed disposal site, and the decision to limit the review to essentially first-round review status, the NRC staff report is labeled a ''Safety Evaluation Status Report'' (SESR). Appendix A comprises the NRC review comments and questions on the information that DOE submitted in the PLASAR. The NRC concentrated its review on the design and operations-related portions of the EMCB PLASAR

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)

An evaluation of an airline cabin safety education program for elementary school children.

Science.gov (United States)

Liao, Meng-Yuan

2014-04-01

The knowledge, attitude, and behavior intentions of elementary school students about airline cabin safety before and after they took a specially designed safety education course were examined. A safety education program was designed for school-age children based on the cabin safety briefings airlines given to their passengers, as well as on lessons learned from emergency evacuations. The course is presented in three modes: a lecture, a demonstration, and then a film. A two-step survey was used for this empirical study: an illustrated multiple-choice questionnaire before the program, and, upon completion, the same questionnaire to assess its effectiveness. Before the program, there were significant differences in knowledge and attitude based on school locations and the frequency that students had traveled by air. After the course, students showed significant improvement in safety knowledge, attitude, and their behavior intention toward safety. Demographic factors, such as gender and grade, also affected the effectiveness of safety education. The study also showed that having the instructor directly interact with students by lecturing is far more effective than presenting the information using only video media. A long-term evaluation, the effectiveness of the program, using TV or video accessible on the Internet to deliver a cabin safety program, and a control group to eliminate potential extraneous factors are suggested for future studies. Copyright © 2013 Elsevier Ltd. All rights reserved.

Sensitivity evaluation of human factors for reliability of the containment spray system

International Nuclear Information System (INIS)

Tsujimura, Yasuhiro; Suzuki, Eiji

1988-01-01

Evaluation of the human reliability is one of the most difficult problems that deal with the safety and reliability of large systems, especially of the Engineered Safety Features (ESF) of the nuclear power plant. Influences of human factors on the reliability of the Containment Spray System in the ESF were estimated by using the FTA method in this paper. As a result, the adequacy of the system structure and the effects of human factors on variations of the design of the system structure were explained. (author)

Improving Safety through Human Factors Engineering.

Science.gov (United States)

Siewert, Bettina; Hochman, Mary G

2015-10-01

Human factors engineering (HFE) focuses on the design and analysis of interactive systems that involve people, technical equipment, and work environment. HFE is informed by knowledge of human characteristics. It complements existing patient safety efforts by specifically taking into consideration that, as humans, frontline staff will inevitably make mistakes. Therefore, the systems with which they interact should be designed for the anticipation and mitigation of human errors. The goal of HFE is to optimize the interaction of humans with their work environment and technical equipment to maximize safety and efficiency. Special safeguards include usability testing, standardization of processes, and use of checklists and forcing functions. However, the effectiveness of the safety program and resiliency of the organization depend on timely reporting of all safety events independent of patient harm, including perceived potential risks, bad outcomes that occur even when proper protocols have been followed, and episodes of "improvisation" when formal guidelines are found not to exist. Therefore, an institution must adopt a robust culture of safety, where the focus is shifted from blaming individuals for errors to preventing future errors, and where barriers to speaking up-including barriers introduced by steep authority gradients-are minimized. This requires creation of formal guidelines to address safety concerns, establishment of unified teams with open communication and shared responsibility for patient safety, and education of managers and senior physicians to perceive the reporting of safety concerns as a benefit rather than a threat. © RSNA, 2015.

Design of Vertical Wall Caisson Breakwaters using Partial Safety Factors

DEFF Research Database (Denmark)

Burcharth, H. F.; Sørensen, John Dalsgaard

1999-01-01

The paper presents a new system for implementation of target reliability in caisson breakwater designs by means of partial safety factors. The development of the system is explained, and tables of partial safety factors are presented for important overall stability failure modes related to caisson...

Safety Culture Evaluation at Research Reactors of Pakistan Atomic Energy Commission

International Nuclear Information System (INIS)

Qamar, M.A.; Saeed, A.; Shah, J.H.

2016-01-01

The concept of safety culture was presented by IAEA in document INSAG-4 (1991), delineated as “assembly of characteristics and attitudes in organizations and individuals which establish that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance”. The purpose of this paper is to describe the evaluation of safety culture at research reactors of the Pakistan Atomic Energy Commission (PAEC). Evaluating the safety culture of a particular organization poses some challenges which can be resolved by using safety culture evaluation models like those of Sachein (1992) and Harber-Barrier(1998). In PAEC, safety culture is the integral part of management system which not only promotes safety culture throughout the organization but also enhances its significance. To strengthen the safety culture, PAEC is also participating in a number of international and regional meetings of IAEA regarding safety culture. PAEC and the national regulator Pakistan Nuclear Regulatory Authority (PNRA) are also arranging workshops, peer reviews, sharing operational experiences and interacting with IAEA missions to enhance its capabilities in the field of safety culture. The Directorate General of Safety (DOS) is a corporate office of PAEC for safety and regulatory matters. DOS is in the process of implementing a program to evaluate safety culture at nuclear installations of PAEC to ensure that safety culture is included as a vital segment of the Integral Management System of the establishment. In this regard, training sessions and lectures on safety culture evaluation are normally conducted in PAEC for awareness and enhancement of the safety culture program. Safety culture is also addressed in PNRA Regulations like PAK-909 and PAK-913. In this paper we will focus on the safety culture evaluation in our research reactors, i.e., PARR-1 and PARR-2. The evaluation results will be based on observations, interviews of employees, group discussions

The Evaluation of the Safety Benefits of Combined Passive and On-Board Active Safety Applications

Science.gov (United States)

Page, Yves; Cuny, Sophie; Zangmeister, Tobias; Kreiss, Jens-Peter; Hermitte, Thierry

2009-01-01

One of the objectives of the European TRACE project (TRaffic Accident Causation in Europe, 2006–2008) was to estimate the proportion of injury accidents that could be avoided and/or the proportion of injury accidents where the severity could be mitigated for on-the-market safety applications, if 100 % of the car fleet would be equipped with them. We have selected for evaluation the Electronic Stability Control (ESC) and the Emergency Brake Assist (EBA) applications. As for passive safety systems, recent cars are designed to offer overall safety protection. Car structure, load limiters, front airbags, side airbags, knee airbags, pretensioners, padding and non aggressive structures in the door panel, the dashboard, the windshield, the seats, and the head rest also contribute to applying more protection. The whole safety package is very difficult to evaluate separately, one element independently segmented from the others. We decided to consider evaluating the effectivenessof the whole passive safety package, This package,, for the sake of simplicity, was the number of stars awarded at the Euro NCAP testing. The challenges were to compare the effectiveness of some safety configuration SC I, with the effectiveness of a different safety configuration SC II. A safety configuration is understood as a package of safety functions. Ten comparisons have been carried out such as the evaluation of the safety benefit of a fifth star given that the car has four stars and an EBA. The main outcome of this analysis is that any addition of a passive or active safety function selected in this analysis is producing increased safety benefits. For example, if all cars were five stars fitted with EBA and ESC, instead of four stars without ESC and EBA, injury accidents would be reduced by 47.2% for severe injuries and 69.5% for fatal injuries. PMID:20184838

Evaluating safety-critical organizations - emphasis on the nuclear industry

Energy Technology Data Exchange (ETDEWEB)

Reiman, Teemu; Oedewald, Pia (VTT, Technical Research Centre of Finland (Finland))

2009-04-15

An organizational evaluation plays a key role in the monitoring, as well as controlling and steering, of the organizational safety culture. If left unattended, organizations have a tendency to gradually drift into a condition where they have trouble identifying their vulnerabilities and mechanisms or practices that create or maintain these vulnerabilities. The aim of an organizational evaluation should be to promote increased understanding of the sociotechnical system and its changing vulnerabilities. Evaluation contributes to organizational development and management. Evaluations are used in various situations, but when the aim is to learn about possible new vulnerabilities, identify organizational reasons for problems, or prepare for future challenges, the organization is most open to genuine surprises and new findings. It is recommended that organizational evaluations should be conducted when - there are changes in the organizational structures - new tools are implemented - when the people report increased workplace stress or a decreased working climate - when incidents and near-misses increase - when work starts to become routine - when weak signals (such as employees voicing safety concerns or other worries, the organization 'feels' different, organizational climate has changed) are perceived. In organizations that already have a high safety level, safety managers work for their successors. This means that they seldom see the results of their successful efforts to improve safety. This is due to the fact that it takes time for the improvement to become noticeable in terms of increased measurable safety levels. The most challenging issue in an organizational evaluation is the definition of criteria for safety. We have adopted a system safety perspective and we state that an organization has a high potential for safety when - safety is genuinely valued and the members of the organization are motivated to put effort on achieving high levels of safety

Evaluating safety-critical organizations - emphasis on the nuclear industry

International Nuclear Information System (INIS)

Reiman, Teemu; Oedewald, Pia

2009-04-01

An organizational evaluation plays a key role in the monitoring, as well as controlling and steering, of the organizational safety culture. If left unattended, organizations have a tendency to gradually drift into a condition where they have trouble identifying their vulnerabilities and mechanisms or practices that create or maintain these vulnerabilities. The aim of an organizational evaluation should be to promote increased understanding of the sociotechnical system and its changing vulnerabilities. Evaluation contributes to organizational development and management. Evaluations are used in various situations, but when the aim is to learn about possible new vulnerabilities, identify organizational reasons for problems, or prepare for future challenges, the organization is most open to genuine surprises and new findings. It is recommended that organizational evaluations should be conducted when - there are changes in the organizational structures - new tools are implemented - when the people report increased workplace stress or a decreased working climate - when incidents and near-misses increase - when work starts to become routine - when weak signals (such as employees voicing safety concerns or other worries, the organization 'feels' different, organizational climate has changed) are perceived. In organizations that already have a high safety level, safety managers work for their successors. This means that they seldom see the results of their successful efforts to improve safety. This is due to the fact that it takes time for the improvement to become noticeable in terms of increased measurable safety levels. The most challenging issue in an organizational evaluation is the definition of criteria for safety. We have adopted a system safety perspective and we state that an organization has a high potential for safety when - safety is genuinely valued and the members of the organization are motivated to put effort on achieving high levels of safety - it is

The likelihood of achieving quantified road safety targets: a binary logistic regression model for possible factors.

Science.gov (United States)

Sze, N N; Wong, S C; Lee, C Y

2014-12-01

In past several decades, many countries have set quantified road safety targets to motivate transport authorities to develop systematic road safety strategies and measures and facilitate the achievement of continuous road safety improvement. Studies have been conducted to evaluate the association between the setting of quantified road safety targets and road fatality reduction, in both the short and long run, by comparing road fatalities before and after the implementation of a quantified road safety target. However, not much work has been done to evaluate whether the quantified road safety targets are actually achieved. In this study, we used a binary logistic regression model to examine the factors - including vehicle ownership, fatality rate, and national income, in addition to level of ambition and duration of target - that contribute to a target's success. We analyzed 55 quantified road safety targets set by 29 countries from 1981 to 2009, and the results indicate that targets that are in progress and with lower level of ambitions had a higher likelihood of eventually being achieved. Moreover, possible interaction effects on the association between level of ambition and the likelihood of success are also revealed. Copyright © 2014 Elsevier Ltd. All rights reserved.

Hazard Management Dealt by Safety Professionals in Colleges: The Impact of Individual Factors

Directory of Open Access Journals (Sweden)

Tsung-Chih Wu

2016-12-01

Full Text Available Identifying, evaluating, and controlling workplace hazards are important functions of safety professionals (SPs. The purpose of this study was to investigate the content and frequency of hazard management dealt by safety professionals in colleges. The authors also explored the effects of organizational factors/individual factors on SPs’ perception of frequency of hazard management. The researchers conducted survey research to achieve the objective of this study. The researchers mailed questionnaires to 200 SPs in colleges after simple random sampling, then received a total of 144 valid responses (response rate = 72%. Exploratory factor analysis indicated that the hazard management scale (HMS extracted five factors, including physical hazards, biological hazards, social and psychological hazards, ergonomic hazards, and chemical hazards. Moreover, the top 10 hazards that the survey results identified that safety professionals were most likely to deal with (in order of most to least frequent were: organic solvents, illumination, other chemicals, machinery and equipment, fire and explosion, electricity, noise, specific chemicals, human error, and lifting/carrying. Finally, the results of one-way multivariate analysis of variance (MANOVA indicated there were four individual factors that impacted the perceived frequency of hazard management which were of statistical and practical significance: job tenure in the college of employment, type of certification, gender, and overall job tenure. SPs within colleges and industries can now discuss plans revolving around these five areas instead of having to deal with all of the separate hazards.

Modelling of safety barriers including human and organisational factors to improve process safety

DEFF Research Database (Denmark)

Markert, Frank; Duijm, Nijs Jan; Thommesen, Jacob

2013-01-01

It is believed that traditional safety management needs to be improved on the aspect of preparedness for coping with expected and unexpected deviations, avoiding an overly optimistic reliance on safety systems. Remembering recent major accidents, such as the Deep Water Horizon, the Texas City....... A valuable approach is the inclusion of human and organisational factors into the simulation of the reliability of the technical system using event trees and fault trees and the concept of safety barriers. This has been demonstrated e.g. in the former European research project ARAMIS (Accidental Risk...

Human factor in the problem of Russian nuclear industry safety

International Nuclear Information System (INIS)

Abramova, V.

2002-01-01

The approach to human factor definition, considered in the paper, consists of recognition of as many as possible factors for developing a complete list of factors, which have influence on mistakes or successful work of NPP personnel. Safety culture is considered as the main factor. The enhancement in nuclear power industry includes an optimization of organizational structures and development of personnel safety attitudes. The organizational factors, as possible root causes for human errors, need to be identified, assessed and improved. The organizational activities taken in Russia are presented

International handbook of evaluated criticality safety benchmark experiments

International Nuclear Information System (INIS)

2010-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span over 55,000 pages and contain 516 evaluations with benchmark specifications for 4,405 critical, near critical, or subcritical configurations, 24 criticality alarm placement / shielding configurations with multiple dose points for each, and 200 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these evaluations; however, benchmark specifications are not derived for such experiments (in some cases models are provided in an appendix). Approximately 770 experimental configurations are categorized as unacceptable for use as criticality safety benchmark experiments. Additional evaluations are in progress and will be

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

Dissipation Pattern, Processing Factors, and Safety Evaluation for Dimethoate and Its Metabolite (Omethoate in Tea (Camellia Sinensis.

Directory of Open Access Journals (Sweden)

Rong Pan

Full Text Available Residue levels of dimethoate and its oxon metabolite (omethoate during tea planting, manufacturing, and brewing were investigated using a modified QuEChERS sample preparation and gas chromatography. Dissipation of dimethoate and its metabolite in tea plantation followed the first-order kinetic with a half-life of 1.08-1.27 d. Tea manufacturing has positive effects on dimethoate dissipation. Processing factors of dimethoate are in the range of 2.11-2.41 and 1.41-1.70 during green tea and black tea manufacturing, respectively. Omethoate underwent generation as well as dissipation during tea manufacturing. Sum of dimethoate and omethoate led to a large portion of 80.5-84.9% transferring into tea infusion. Results of safety evaluation indicated that omethoate could bring higher human health risk than dimethoate due to its higher hazard quotient by drinking tea. These results would provide information for the establishment of maximum residue limit and instruction for the application of dimethoate formulation on tea crop.

Safety equipment and methods for evaluating its effectiveness

Energy Technology Data Exchange (ETDEWEB)

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

1993-05-01

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

Consumer and farmer safety evaluation of application of botanical pesticides in black pepper crop protection.

Science.gov (United States)

Hernández-Moreno, David; Soffers, Ans E M F; Wiratno; Falke, Hein E; Rietjens, Ivonne M C M; Murk, Albertinka J

2013-06-01

This study presents a consumer and farmer safety evaluation on the use of four botanical pesticides in pepper berry crop protection. The pesticides evaluated include preparations from clove, tuba root, sweet flag and pyrethrum. Their safety evaluation was based on their active ingredients being eugenol, rotenone, β-asarone and pyrethrins, respectively. Botanical pesticides from Acorus calamus are of possible concern because of the genotoxic and carcinogenic ingredient β-asarone although estimated margins of exposure (MOE) for consumers indicate a low priority for risk management. For the other three botanical pesticides the margin of safety (MOS) between established acute reference doses and/or acceptable daily intake values and intake estimates for the consumer, resulting from their use as a botanical pesticide are not of safety concern, with the exception for levels of rotenone upon use of tuba root extracts on stored berries. Used levels of clove and pyrethrum as botanical pesticides in pepper berry crop production is not of safety concern for consumers or farmers, whereas for use of tuba root and sweet flag some risk factors were defined requiring further evaluation and/or risk management. It seems prudent to look for alternatives for use of sweet flag extracts containing β-asarone. Copyright © 2013 Elsevier Ltd. All rights reserved.

Reliability Analysis and Calibration of Partial Safety Factors for Redundant Structures

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard

1998-01-01

Redundancy is important to include in the design and analysis of structural systems. In most codes of practice redundancy is not directly taken into account. In the paper various definitions of a deterministic and reliability based redundancy measure are reviewed. It is described how reundancy can...... be included in the safety system and how partial safety factors can be calibrated. An example is presented illustrating how redundancy is taken into account in the safety system in e.g. the Danish codes. The example shows how partial safety factors can be calibrated to comply with the safety level...

An Evaluation of the Effects of Human Factors and Ergonomics on Health Care and Patient Safety Practices: A Systematic Review

Science.gov (United States)

Zhang, Longhao; Zhao, Pujing; Chen, Ying; Zhang, Mingming

2015-01-01

Background From the viewpoint of human factors and ergonomics (HFE), errors often occur because of the mismatch between the system, technique and characteristics of the human body. HFE is a scientific discipline concerned with understanding interactions between human behavior, system design and safety. Objective To evaluate the effectiveness of HFE interventions in improving health care workers’ outcomes and patient safety and to assess the quality of the available evidence. Methods We searched databases, including MEDLINE, EMBASE, BIOSIS Previews and the CBM (Chinese BioMedical Literature Database), for articles published from 1996 to Mar.2015. The quality assessment tool was based on the risk of bias criteria developed by the Cochrane Effective Practice and Organization of Care (EPOC) Group. The interventions of the included studies were categorized into four relevant domains, as defined by the International Ergonomics Association. Results For this descriptive study, we identified 8, 949 studies based on our initial search. Finally, 28 studies with 3,227 participants were included. Among the 28 included studies, 20 studies were controlled studies, two of which were randomized controlled trials. The other eight studies were before/after surveys, without controls. Most of the studies were of moderate or low quality. Five broad categories of outcomes were identified in this study: 1) medical errors or patient safety, 2) health care workers’ quality of working life (e.g. reduced fatigue, discomfort, workload, pain and injury), 3) user performance (e.g., efficiency or accuracy), 4) health care workers’ attitudes towards the interventions(e.g., satisfaction and preference), and 5) economic evaluations. Conclusion The results showed that the interventions positively affected the outcomes of health care workers. Few studies considered the financial merits of these interventions. Most of the included studies were of moderate quality. This review highlights the need

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

Directory of Open Access Journals (Sweden)

B. V. Zubkov

2017-01-01

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

[Evaluating training programs on occupational health and safety: questionnaire development].

Science.gov (United States)

Zhou, Xiao-Yan; Wang, Zhi-Ming; Wang, Mian-Zhen

2006-03-01

To develop a questionnaire to evaluate the quality of training programs on occupational health and safety. A questionnaire comprising five subscales and 21 items was developed. The reliability and validity of the questionnaire was tested. Final validation of the questionnaire was undertaken in 700 workers in an oil refining company. The Cronbach's alpha coefficients of the five subscales ranged from 0.6194 to 0.6611. The subscale-scale Pearson correlation coefficients ranged from 0.568 to 0.834 . The theta coefficients of the five subscales were greater than 0.7. The factor loadings of the five subscales in the principal component analysis ranged from 0.731 to 0.855. Use of the questionnaire in the 700 workers produced a good discriminability, with excellent, good, fair and poor comprising 22.2%, 31.2%, 32.4% and 14.1 respectively. Given the fact that 18.7% of workers had never been trained and 29.7% of workers got one-off training only, the training program scored an average of 57.2. The questionnaire is suitable to be used in evaluating the quality of training programs on occupational health and safety. The oil refining company needs to improve training for their workers on occupational health and safety.

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

Automating the Human Factors Engineering and Evaluation Processes

International Nuclear Information System (INIS)

Mastromonico, C.

2002-01-01

The Westinghouse Savannah River Company (WSRC) has developed a software tool for automating the Human Factors Engineering (HFE) design review, analysis, and evaluation processes. The tool provides a consistent, cost effective, graded, user-friendly approach for evaluating process control system Human System Interface (HSI) specifications, designs, and existing implementations. The initial set of HFE design guidelines, used in the tool, was obtained from NUREG- 0700. Each guideline was analyzed and classified according to its significance (general concept vs. supporting detail), the HSI technology (computer based vs. non-computer based), and the HSI safety function (safety vs. non-safety). Approximately 10 percent of the guidelines were determined to be redundant or obsolete and were discarded. The remaining guidelines were arranged in a Microsoft Access relational database, and a Microsoft Visual Basic user interface was provided to facilitate the HFE design review. The tool also provides the capability to add new criteria to accommodate advances in HSI technology and incorporate lessons learned. Summary reports produced by the tool can be easily ported to Microsoft Word and other popular PC office applications. An IBM compatible PC with Microsoft Windows 95 or higher is required to run the application

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

Research on the Evaluation System for Rural Public Safety Planning

Institute of Scientific and Technical Information of China (English)

Ming; SUN; Jianxin; YAN

2014-01-01

The indicator evaluation system is introduced to the study of rural public safety planning in this article.By researching the current rural public safety planning and environmental carrying capacity,we select some carrying capacity indicators influencing the rural public safety,such as land,population,ecological environment,water resources,infrastructure,economy and society,to establish the environmental carrying capacity indicator system.We standardize the indicators,use gray correlation analysis method to determine the weight of indicators,and make DEA evaluation of the indicator system,to obtain the evaluation results as the basis for decision making in rural safety planning,and provide scientific and quantified technical support for rural public safety planning.

Patient safety - the role of human factors and systems engineering.

Science.gov (United States)

Carayon, Pascale; Wood, Kenneth E

2010-01-01

Patient safety is a global challenge that requires knowledge and skills in multiple areas, including human factors and systems engineering. In this chapter, numerous conceptual approaches and methods for analyzing, preventing and mitigating medical errors are described. Given the complexity of healthcare work systems and processes, we emphasize the need for increasing partnerships between the health sciences and human factors and systems engineering to improve patient safety. Those partnerships will be able to develop and implement the system redesigns that are necessary to improve healthcare work systems and processes for patient safety.

Patient Safety: The Role of Human Factors and Systems Engineering

Science.gov (United States)

Carayon, Pascale; Wood, Kenneth E.

2011-01-01

Patient safety is a global challenge that requires knowledge and skills in multiple areas, including human factors and systems engineering. In this chapter, numerous conceptual approaches and methods for analyzing, preventing and mitigating medical errors are described. Given the complexity of healthcare work systems and processes, we emphasize the need for increasing partnerships between the health sciences and human factors and systems engineering to improve patient safety. Those partnerships will be able to develop and implement the system redesigns that are necessary to improve healthcare work systems and processes for patient safety. PMID:20543237

Criticality safety benchmark evaluation project: Recovering the past

Energy Technology Data Exchange (ETDEWEB)

Trumble, E.F.

1997-06-01

A very brief summary of the Criticality Safety Benchmark Evaluation Project of the Westinghouse Savannah River Company is provided in this paper. The purpose of the project is to provide a source of evaluated criticality safety experiments in an easily usable format. Another project goal is to search for any experiments that may have been lost or contain discrepancies, and to determine if they can be used. Results of evaluated experiments are being published as US DOE handbooks.

Efficiency evaluation of a safety department in a construction company-A case study: A DEA approach

Directory of Open Access Journals (Sweden)

Solomon Odeyale

2015-01-01

Full Text Available Data Envelopment Analysis (DEA is a decision making tool based on linear programming for measuring the relative efficiency of a set of comparable units. DEA helps us identify the sources and level of inefficiency for each of the inputs and outputs. This approach has been used to evaluate the efficiency of the safety department in five construction companies. A three-input, safety workforce, safety training, and safety budget, and two-output, Perfect days and Uptime, constant returns-to-scale (CRS model was developed. The model indicated the necessary improvements required in the inefficient unit’s inputs and outputs to make it efficient, by identifying what factor is responsible for the low efficiency of performance, and also what factor should be improved in order to improve the efficiency of the safety department. The result shows that the safety department of firm A, B and D are efficient, but Firm C and Firm E can improve their efficiency by reducing inputs up to 3.34% and 6.05%, respectively. The inputs identified for reduction were; number of safety staffs and safety budget for Firm C and E respectively.

Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

International Nuclear Information System (INIS)

Mitake, Susumu

2003-01-01

Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions

Quality factors quantification/assurance for software related to safety in nuclear power plants

International Nuclear Information System (INIS)

Nunez McLeod, J.E.; Rivera, S.S.

1997-01-01

Quality assurance plan is needed to guarantee the software quality. The use of such a plan involves activities that should take place all along the life cycle, and which can be evaluated using the so called quality factors. This is due to the fact that the quality itself cannot be measured, but some of its manifestations can be used for this purpose. In the present work, a methodology to quantify a set of quality factors is proposed, for software based systems to be used in safety related areas in nuclear power plants. (author) [es

Partial safety factor calibration from stochastic finite element computation of welded joint with random geometries

International Nuclear Information System (INIS)

Schoefs, Franck; Chevreuil, Mathilde; Pasqualini, Olivier; Cazuguel, Mikaël

2016-01-01

Welded joints are used in various structures and infrastructures like bridges, ships and offshore structures, and are submitted to cyclic stresses. Their fatigue behaviour is an industrial key issue to deal with and still offers original research subjects. One of the available methods relies on the computing of the stress concentration factor. Even if some studies were previously driven to evaluate this factor onto some cases of welded structures, the shape of the weld joint is generally idealized through a deterministic parametric geometry. Previous experimental works however have shown that this shape plays a key role in the lifetime assessment. We propose in this paper a methodology for computing the stress concentration factor in presence of random geometries of welded joints. In view to make the results available by engineers, this method merges stochastic computation and semi-probabilistic analysis by computing partial safety factors with a dedicated method. - Highlights: • Numerical computation of stress concentration factor with random geometry of weld. • Real data are used for probabilistic modelling. • Identification of partial safety factor from SFEM computation in case of random geometries.

Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos

2009-01-01

Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: vasconv@cdtn.br, e-mail: silvaem@cdtn.br, e-mail: aclc@cdtn.br, e-mail: reissc@cdtn.br

2009-07-01

Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

Safety evaluation of synthetic β-carotene

NARCIS (Netherlands)

Woutersen, R.A.; Wolterbeek, A.P.M.; Appel, M.J.; Berg, H. van den; Goldbohm, R.A.; Feron, V.J.

1999-01-01

The safety of β-carotene was reassessed by evaluating the relevant literature on the beneficial and adverse effects of β-carotene on cancer and, in particular, by evaluating the results of toxicity studies. β- Carotene appeared neither genotoxic nor reprotoxic or teratogenic, and no signs of organ

CT fluoroscopy-guided renal tumour cutting needle biopsy. Retrospective evaluation of diagnostic yield, safety, and risk factors for diagnostic failure

International Nuclear Information System (INIS)

Iguchi, Toshihiro; Hiraki, Takao; Matsui, Yusuke; Fujiwara, Hiroyasu; Sakurai, Jun; Masaoka, Yoshihisa; Gobara, Hideo; Kanazawa, Susumu

2018-01-01

To evaluate retrospectively the diagnostic yield, safety, and risk factors for diagnostic failure of computed tomography (CT) fluoroscopy-guided renal tumour biopsy. Biopsies were performed for 208 tumours (mean diameter 2.3 cm; median diameter 2.1 cm; range 0.9-8.5 cm) in 199 patients. One hundred and ninety-nine tumours were ≤4 cm. All 208 initial procedures were divided into diagnostic success and failure groups. Multiple variables related to the patients, lesions, and procedures were assessed to determine the risk factors for diagnostic failure. After performing 208 initial and nine repeat biopsies, 180 malignancies and 15 benign tumours were pathologically diagnosed, whereas 13 were not diagnosed. In 117 procedures, 118 Grade I and one Grade IIIa adverse events (AEs) occurred. Neither Grade ≥IIIb AEs nor tumour seeding were observed within a median follow-up period of 13.7 months. Logistic regression analysis revealed only small tumour size (≤1.5 cm; odds ratio 3.750; 95% confidence interval 1.362-10.326; P = 0.011) to be a significant risk factor for diagnostic failure. CT fluoroscopy-guided renal tumour biopsy is a safe procedure with a high diagnostic yield. A small tumour size (≤1.5 cm) is a significant risk factor for diagnostic failure. (orig.)

CT fluoroscopy-guided renal tumour cutting needle biopsy. Retrospective evaluation of diagnostic yield, safety, and risk factors for diagnostic failure

Energy Technology Data Exchange (ETDEWEB)

Iguchi, Toshihiro; Hiraki, Takao; Matsui, Yusuke; Fujiwara, Hiroyasu; Sakurai, Jun; Masaoka, Yoshihisa; Gobara, Hideo; Kanazawa, Susumu [Okayama University Medical School, Department of Radiology, Okayama (Japan)

2018-01-15

To evaluate retrospectively the diagnostic yield, safety, and risk factors for diagnostic failure of computed tomography (CT) fluoroscopy-guided renal tumour biopsy. Biopsies were performed for 208 tumours (mean diameter 2.3 cm; median diameter 2.1 cm; range 0.9-8.5 cm) in 199 patients. One hundred and ninety-nine tumours were ≤4 cm. All 208 initial procedures were divided into diagnostic success and failure groups. Multiple variables related to the patients, lesions, and procedures were assessed to determine the risk factors for diagnostic failure. After performing 208 initial and nine repeat biopsies, 180 malignancies and 15 benign tumours were pathologically diagnosed, whereas 13 were not diagnosed. In 117 procedures, 118 Grade I and one Grade IIIa adverse events (AEs) occurred. Neither Grade ≥IIIb AEs nor tumour seeding were observed within a median follow-up period of 13.7 months. Logistic regression analysis revealed only small tumour size (≤1.5 cm; odds ratio 3.750; 95% confidence interval 1.362-10.326; P = 0.011) to be a significant risk factor for diagnostic failure. CT fluoroscopy-guided renal tumour biopsy is a safe procedure with a high diagnostic yield. A small tumour size (≤1.5 cm) is a significant risk factor for diagnostic failure. (orig.)

The use of human factors methods to identify and mitigate safety issues in radiation therapy

International Nuclear Information System (INIS)

Chan, Alvita J.; Islam, Mohammad K.; Rosewall, Tara; Jaffray, David A.; Easty, Anthony C.; Cafazzo, Joseph A.

2010-01-01

Background and purpose: New radiation therapy technologies can enhance the quality of treatment and reduce error. However, the treatment process has become more complex, and radiation dose is not always delivered as intended. Using human factors methods, a radiotherapy treatment delivery process was evaluated, and a redesign was undertaken to determine the effect on system safety. Material and methods: An ethnographic field study and workflow analysis was conducted to identify human factors issues of the treatment delivery process. To address specific issues, components of the user interface were redesigned through a user-centered approach. Sixteen radiation therapy students were then used to experimentally evaluate the redesigned system through a usability test to determine the effectiveness in mitigating use errors. Results: According to findings from the usability test, the redesigned system successfully reduced the error rates of two common errors (p < .04 and p < .01). It also improved the mean task completion time by 5.5% (p < .02) and achieved a higher level of user satisfaction. Conclusions: These findings demonstrated the importance and benefits of applying human factors methods in the design of radiation therapy systems. Many other opportunities still exist to improve patient safety in this area using human factors methods.

Patient safety risk factors in minimally invasive surgery : A validation study

NARCIS (Netherlands)

Rodrigues, S.P.; Ter Kuile, M.; Dankelman, J.; Jansen, F.W.

2012-01-01

This study was conducted to adapt and validate a patient safety (PS) framework for minimally invasive surgery (MIS) as a first step in understanding the clinical relevance of various PS risk factors in MIS. Eight patient safety risk factor domains were identified using frameworks from a systems

2005 dossier: clay. Tome: safety evaluation of the geologic disposal

International Nuclear Information System (INIS)

2005-01-01

This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of an argilite-type geologic disposal facility for high-level and long-lived (HLLL) radioactive wastes. Content: 1 - safety approach: context and general goals, general safety principles, specificity of the argilite repository safety approach, general approach; 2 - general description: HLLL wastes, geologic context of the Meuse/Haute-Marne site, repository architecture; 3 - safety functions and disposal design: time and space scales, safety approach by functions, functional analysis methodology, analysis of safety functions during the construction, exploitation and observation phases, safety functions analysis during post-closure phase; 4 - operational safety: dosimetric evaluation, risk analysis (explosible gases, fire hazards, lift cage drop, container drop); 5 - long-term efficiency of the disposal facility: normal evolution scenario, from conceptual models to the safety calculation model, description of the safety model, quantitative evaluation of the normal evolution scenario, main lessons learnt from the efficiency analysis; 6 - management of uncertainties: identification, building up of altered situations, mastery of uncertainties; 7 - evaluation of altered evolution scenarios: sealing defect scenario, container defect scenario, drilling scenario, strongly degraded operation scenario; 8 - conclusions: lessons learnt, possible improvements. (J.S.)

National plan to enhance aviation safety through human factors improvements

Science.gov (United States)

Foushee, Clay

1990-01-01

The purpose of this section of the plan is to establish a development and implementation strategy plan for improving safety and efficiency in the Air Traffic Control (ATC) system. These improvements will be achieved through the proper applications of human factors considerations to the present and future systems. The program will have four basic goals: (1) prepare for the future system through proper hiring and training; (2) develop a controller work station team concept (managing human errors); (3) understand and address the human factors implications of negative system results; and (4) define the proper division of responsibilities and interactions between the human and the machine in ATC systems. This plan addresses six program elements which together address the overall purpose. The six program elements are: (1) determine principles of human-centered automation that will enhance aviation safety and the efficiency of the air traffic controller; (2) provide new and/or enhanced methods and techniques to measure, assess, and improve human performance in the ATC environment; (3) determine system needs and methods for information transfer between and within controller teams and between controller teams and the cockpit; (4) determine how new controller work station technology can optimally be applied and integrated to enhance safety and efficiency; (5) assess training needs and develop improved techniques and strategies for selection, training, and evaluation of controllers; and (6) develop standards, methods, and procedures for the certification and validation of human engineering in the design, testing, and implementation of any hardware or software system element which affects information flow to or from the human.

[Role of some psycho-physiological factors on driving safety].

Science.gov (United States)

Bergomi, M; Vivoli, G; Rovesti, S; Bussetti, P; Ferrari, A; Vivoli, R

2010-01-01

Within a research project on the role played by human factors on road accidents, the aim of the present study is to evaluate, in young adults, the relationships between driver behaviour and personality factors as well as to assess the neuroendocrine correlates of psychological and behavioural factors investigated. Another aim is to estimate in what measure the performance levels are affected by demographic, psychological and chronobiological variables. It has been found a positive relation between highway code violations, extroversion trait of personality and Sensation Seeking scores, so confirming that this component of personality can affect risky behaviour. Furthermore the subjects more oriented to morningness chronotype were found to be prone to adopt safe driving behaviour. Regarding the relations of the neuroendocrine parameters and driving behaviour a positive correlation was observed between dopamine levels and frequency of driving violations while a negative relationship was found between adrenaline levels and frequency of driving errors. In conclusion the identification of psycho-physiological variables related to driving risky behaviour might be a useful instrument to design traffic safety programs tailored to high risk subjects.

Human factors and safety in emergency medicine

Science.gov (United States)

Schaefer, H. G.; Helmreich, R. L.; Scheidegger, D.

1994-01-01

A model based on an input process and outcome conceptualisation is suggested to address safety-relevant factors in emergency medicine. As shown in other dynamic and demanding environments, human factors play a decisive role in attaining high quality service. Attitudes held by health-care providers, organisational shells and work-cultural parameters determine communication, conflict resolution and workload distribution within and between teams. These factors should be taken into account to improve outcomes such as operational integrity, job satisfaction and morale.

Dissipation Pattern, Processing Factors, and Safety Evaluation for Dimethoate and Its Metabolite (Omethoate) in Tea (Camellia Sinensis)

Science.gov (United States)

Pan, Rong; Chen, Hong-Ping; Zhang, Ming-Lu; Wang, Qing-Hua; Jiang, Ying; Liu, Xin

2015-01-01

Residue levels of dimethoate and its oxon metabolite (omethoate) during tea planting, manufacturing, and brewing were investigated using a modified QuEChERS sample preparation and gas chromatography. Dissipation of dimethoate and its metabolite in tea plantation followed the first-order kinetic with a half-life of 1.08–1.27 d. Tea manufacturing has positive effects on dimethoate dissipation. Processing factors of dimethoate are in the range of 2.11–2.41 and 1.41–1.70 during green tea and black tea manufacturing, respectively. Omethoate underwent generation as well as dissipation during tea manufacturing. Sum of dimethoate and omethoate led to a large portion of 80.5–84.9% transferring into tea infusion. Results of safety evaluation indicated that omethoate could bring higher human health risk than dimethoate due to its higher hazard quotient by drinking tea. These results would provide information for the establishment of maximum residue limit and instruction for the application of dimethoate formulation on tea crop. PMID:26406463

Factors impacting on the microbiological quality and safety of ...

African Journals Online (AJOL)

Problems with the safety and shelf life of export hake have been raised by the Namibian fishing industry. This prompted an investigation into the factors that may have an impact on the microbiological quality and safety of processed hake. Samples were collected along the processing line; the general microbiological quality ...

The International Criticality Safety Benchmark Evaluation Project

International Nuclear Information System (INIS)

Briggs, B. J.; Dean, V. F.; Pesic, M. P.

2001-01-01

In order to properly manage the risk of a nuclear criticality accident, it is important to establish the conditions for which such an accident becomes possible for any activity involving fissile material. Only when this information is known is it possible to establish the likelihood of actually achieving such conditions. It is therefore important that criticality safety analysts have confidence in the accuracy of their calculations. Confidence in analytical results can only be gained through comparison of those results with experimental data. The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the US Department of Energy. The project was managed through the Idaho National Engineering and Environmental Laboratory (INEEL), but involved nationally known criticality safety experts from Los Alamos National Laboratory, Lawrence Livermore National Laboratory, Savannah River Technology Center, Oak Ridge National Laboratory and the Y-12 Plant, Hanford, Argonne National Laboratory, and the Rocky Flats Plant. An International Criticality Safety Data Exchange component was added to the project during 1994 and the project became what is currently known as the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Representatives from the United Kingdom, France, Japan, the Russian Federation, Hungary, Kazakhstan, Korea, Slovenia, Yugoslavia, Spain, and Israel are now participating on the project In December of 1994, the ICSBEP became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency's (OECD-NEA) Nuclear Science Committee. The United States currently remains the lead country, providing most of the administrative support. The purpose of the ICSBEP is to: (1) identify and evaluate a comprehensive set of critical benchmark data; (2) verify the data, to the extent possible, by reviewing original and subsequently revised documentation, and by talking with the

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

International Nuclear Information System (INIS)

Kim, Yun Goo; Oh, Eung Se

2016-01-01

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

Economic evaluation in patient safety: a literature review of methods.

Science.gov (United States)

de Rezende, Bruna Alves; Or, Zeynep; Com-Ruelle, Laure; Michel, Philippe

2012-06-01

Patient safety practices, targeting organisational changes for improving patient safety, are implemented worldwide but their costs are rarely evaluated. This paper provides a review of the methods used in economic evaluation of such practices. International medical and economics databases were searched for peer-reviewed publications on economic evaluations of patient safety between 2000 and 2010 in English and French. This was complemented by a manual search of the reference lists of relevant papers. Grey literature was excluded. Studies were described using a standardised template and assessed independently by two researchers according to six quality criteria. 33 articles were reviewed that were representative of different patient safety domains, data types and evaluation methods. 18 estimated the economic burden of adverse events, 3 measured the costs of patient safety practices and 12 provided complete economic evaluations. Healthcare-associated infections were the most common subject of evaluation, followed by medication-related errors and all types of adverse events. Of these, 10 were selected that had adequately fulfilled one or several key quality criteria for illustration. This review shows that full cost-benefit/utility evaluations are rarely completed as they are resource intensive and often require unavailable data; some overcome these difficulties by performing stochastic modelling and by using secondary sources. Low methodological transparency can be a problem for building evidence from available economic evaluations. Investing in the economic design and reporting of studies with more emphasis on defining study perspectives, data collection and methodological choices could be helpful for strengthening our knowledge base on practices for improving patient safety.

Safety evaluation of cation-exchange resins

International Nuclear Information System (INIS)

Kalkwarf, D.R.

1977-08-01

Results are presented of a study to evaluate whether sufficient information is available to establish conservative limits for the safe use of cation-exchange resins in separating radionuclides and, if not, to recommend what new data should be acquired. The study was also an attempt to identify in-line analytical techniques for the evaluation of resin degradation during radionuclide processing. The report is based upon a review of the published literature and upon discussions with many people engaged in the use of these resins. It was concluded that the chief hazard in the use of cation-exchange resins for separating radionuclides is a thermal explosion if nitric acid or other strong oxidants are present in the process solution. Thermal explosions can be avoided by limiting process parameters so that the rates of heat and gas generation in the system do not exceed the rates for their transfer to the surroundings. Such parameters include temperature, oxidant concentration, the amounts of possible catalysts, the radiation dose absorbed by the resin and the diameter of the resin column. Current information is not sufficient to define safe upper limits for these parameters. They can be evaluated, however, from equations derived from the Frank-Kamenetskii theory of thermal explosions provided the heat capacities, thermal conductivities and rates of heat evolution in the relevant resin-oxidant mixtures are known. It is recommended that such measurements be made and the appropriate limits be evaluated. A list of additional safety precautions are also presented to aid in the application of these limits and to provide additional margins of safety. In-line evaluation of resin degradation to assess its safety hazard is considered impractical. Rather, it is recommended that the resin be removed from use before it has received the limiting radiation dose, evaluated as described above

Nuclear criticality safety parameter evaluation for uranium metallic alloy

Energy Technology Data Exchange (ETDEWEB)

Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

2013-07-01

Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

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

Science.gov (United States)

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

2012-12-01

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

Evaluation and Customization of WHO Safety Checklist for Patient Safety in Otorhinolaryngology.

Science.gov (United States)

Dabholkar, Yogesh; Velankar, Haritosh; Suryanarayan, Sneha; Dabholkar, Twinkle Y; Saberwal, Akanksha A; Verma, Bhavika

2018-03-01

The WHO has designed a safe surgery checklist to enhance communication and awareness of patient safety during surgery and to minimise complications. WHO recommends that the check-list be evaluated and customised by end users as a tool to promote safe surgery. The aim of present study was to evaluate the impact of WHO safety checklist on patient safety awareness in otorhinolaryngology and to customise it for the speciality. A prospective structured questionnaire based study was done in ENT operating room for duration of 1 month each for cases, before and after implementation of safe surgery checklist. The feedback from respondents (surgeons, nurses and anaesthetists) was used to arrive at a customised checklist for otolaryngology as per WHO guidelines. The checklist significantly improved team member's awareness of patient's identity (from 17 to 86%) and each other's identity and roles (from 46 to 94%) and improved team communication (from 73 to 92%) in operation theatre. There was a significant improvement in preoperative check of equipment and critical events were discussed more frequently. The checklist could be effectively customised to suit otolaryngology needs as per WHO guidelines. The modified checklist needs to be validated by otolaryngology associations. We conclude from our study that the WHO Surgical safety check-list has a favourable impact on patient safety awareness, team-work and communication of operating team and can be customised for otolaryngology setting.

In vitro and in vivo evaluation of efficacy and safety of photoprotective formulations containing antioxidant extracts

Directory of Open Access Journals (Sweden)

Maria Cristina P.P. Reis Mansur

Full Text Available ABSTRACT Chronic exposure to solar radiation could contribute to premature skin aging and skin cancer. Skin presents its own antioxidant defense, however when defenses are out of balance, reactive oxygen species could damage biological structures. In the present work, an oil-in-water photoprotective emulsion was developed and Bauhinia microstachya var. massambabensis Vaz, Fabaceae, extracts at 1% (obtained by extraction with different solvents were added to this emulsion. In vitro and in vivo efficacy and safety of the formulations were evaluated. Spectrophotometric methods and in vivo Colipa test were performed to evaluated efficacy of the formulations, through sun protection factor (SPF determination and UVA protection factor assessment. To the in vitro safety assessment HET-CAM, CAM-TBS and Red Blood Cell tests were performed. Results showed that both extracts contributed to a higher in vivo photoprotection (SPF 18 when compared to the formulation without extract (SPF 13, this result could be attributed to the antioxidant activity of the plant extracts that act by capturing reactive oxygen species. Concerning safety, all formulations were considered non-irritant according to in vitro tests. Formulations containing extracts could be considered efficient and safe for cosmetic use since they presented higher sun protection factor and passed the toxicity tests.

Studying the Relationship between Individual and Organizational Factors and Nurses' Perception of Patient Safety Culture

Directory of Open Access Journals (Sweden)

Farahnaz Abdolahzadeh

2012-11-01

Full Text Available Introduction: Safety culture is considered as an important factor in improving patient safety. Therefore, identifying individual and organizational factors affecting safety culture is crucial. This study was carried out to determine individual and organizational factors associated with nurses' perception of patient safety culture. Methods: The present descriptive study included 940 nurses working in four training hospitals affiliated with Urmia University of Medical Sciences (Iran. Data was collected through the self-report questionnaire of patient safety culture. Descriptive (number, percent, mean, and standard deviation and inferential (t-test and analysis of variance statistics were used to analyze the data in SPSS. Results: Nurses' perception of patient safety culture was significantly correlated with marital status, workplace, and overtime hours. Conclusion: The results of this study revealed that some individual and organizational factors can impact on nurses' perception of patient safety culture. Nursing authorities should thus pay more attention to factors which promote patient safety culture and ultimately the safety of provided services.

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

Human factors and systems engineering approach to patient safety for radiotherapy.

Science.gov (United States)

Rivera, A Joy; Karsh, Ben-Tzion

2008-01-01

The traditional approach to solving patient safety problems in healthcare is to blame the last person to touch the patient. But since the publication of To Err is Human, the call has been instead to use human factors and systems engineering methods and principles to solve patient safety problems. However, an understanding of the human factors and systems engineering is lacking, and confusion remains about what it means to apply their principles. This paper provides a primer on them and their applications to patient safety.

Human Factors and Systems Engineering Approach to Patient Safety for Radiotherapy

International Nuclear Information System (INIS)

Rivera, A. Joy; Karsh, Ben-Tzion

2008-01-01

The traditional approach to solving patient safety problems in healthcare is to blame the last person to touch the patient. But since the publication of To Err is Human, the call has been instead to use human factors and systems engineering methods and principles to solve patient safety problems. However, an understanding of the human factors and systems engineering is lacking, and confusion remains about what it means to apply their principles. This paper provides a primer on them and their applications to patient safety

Packaging Evaluation Approach to Improve Cosmetic Product Safety

OpenAIRE

Benedetta Briasco; Priscilla Capra; Arianna Cecilia Cozzi; Barbara Mannucci; Paola Perugini

2016-01-01

In the Regulation 1223/2009, evaluation of packaging has become mandatory to assure cosmetic product safety. In fact, the safety assessment of a cosmetic product can be successfully carried out only if the hazard deriving from the use of the designed packaging for the specific product is correctly evaluated. Despite the law requirement, there is too little information about the chemical-physical characteristics of finished packaging and the possible interactions between formulation and packag...

Systematic safety evaluation of old nuclear power plants

International Nuclear Information System (INIS)

Dredemis, G.; Fourest, B.

1984-01-01

The French safety authorities have undertaken a systematic evaluation of the safety of old nuclear power plants. Apart from a complete revision of safety documents (safety analysis report, general operating rules, incident and accident procedures, internal emergency plan, quality organisation manual), this examination consisted of analysing the operating experience of systems frequently challenged and a systematic examination of the safety-related systems. This paper is based on an exercise at the Ardennes Nuclear Power Plant which has been in operation for 15 years. This paper also summarizes the main surveys and modifications relating to this power plant. (orig.)

The Interagency Nuclear Safety Review Panel's Galileo safety evaluation report

International Nuclear Information System (INIS)

Nelson, R.C.; Gray, L.B.; Huff, D.A.

1989-01-01

The safety evaluation report (SER) for Galileo was prepared by the Interagency Nuclear Safety Review Panel (INSRP) coordinators in accordance with Presidential directive/National Security Council memorandum 25. The INSRP consists of three coordinators appointed by their respective agencies, the Department of Defense, the Department of Energy (DOE), and the National Aeronautics and Space Administration (NASA). These individuals are independent of the program being evaluated and depend on independent experts drawn from the national technical community to serve on the five INSRP subpanels. The Galileo SER is based on input provided by the NASA Galileo Program Office, review and assessment of the final safety analysis report prepared by the Office of Special Applications of the DOE under a memorandum of understanding between NASA and the DOE, as well as other related data and analyses. The SER was prepared for use by the agencies and the Office of Science and Technology Policy, Executive Office of the Present for use in their launch decision-making process. Although more than 20 nuclear-powered space missions have been previously reviewed via the INSRP process, the Galileo review constituted the first review of a nuclear power source associated with launch aboard the Space Transportation System

Identification and assessment of organisational factors related to the safety of NPPs - State-of-the-Art Report

International Nuclear Information System (INIS)

Baumont, Genevieve; Bourrier, Mathilde; Frischknecht, Albert; Schoenfeld, Isabelle; Weber, Mike J.

1999-01-01

The initiation of this State-of-the-Art Report (SOAR) on Organisational Factors Identification and Assessment comes from operating experience associated with a number of major events world-wide which caused power plants to be shutdown for a significant period of time. Root cause assessments of these events identified weaknesses in organisational factors as contributing to these events. There is general recognition that organisational factors need to be evaluated for their contribution to plant safety performance and risk to prevent their recurrence in events. There is a need to collect and analyse operational and event data from the nuclear environment to determine the safety and risk significance of organisational factors, to identify assessment methods for those factors, and to gain peer review of the results to ensure credibility and acceptability of these methods and possibly their measures. The SOAR presents a representative view of developments in this field and addresses the following topics: - identification of organisational factors; - identification of methods for the evaluation of organisational factors; - identification of methods for the evaluation of whole organisations; - identification of gaps in knowledge and needed research to evaluate adequately the influence of organisation and management on safety and risk. The workshop participants identified 12 organisational factors as important to assess in determining organisational safety performance. They are: external influences; goals and strategies; management functions and overview; resource allocation; human resource management; training; co-ordination of work; organisational knowledge; proceduralization; organisational culture; organisational learning; and communication. Different cultural backgrounds of participants using their own terminology sometimes made it difficult to have a common definition for certain factors. Some factors could be defined by consensus; other factors such as

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

Directory of Open Access Journals (Sweden)

E Mohammadfam

2015-11-01

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

Modelling of Safety Factors in the Design of GRP Composite Products

DEFF Research Database (Denmark)

Babu, B.J.C.; Prabhakaran, R.T. Durai; Lystrup, Aage

2010-01-01

as independent, while in real applications these factors may interact/influence each other. Following the concept developed by the authors, a simple graph theoretic model has been used to determine overall factor of safety. This is described with the help of an example and it has been demonstrated......An attempt has been made in this paper to arrive at the safety factor design of glass fibre reinforced polymer (GRP) composite products using graph theoretic model. In the conventional design and recommendations of the standards, these design factors affecting properties have been considered...

29 CFR 1960.80 - Secretary's evaluations of agency occupational safety and health programs.

Science.gov (United States)

2010-07-01

... EMPLOYEE OCCUPATIONAL SAFETY AND HEALTH PROGRAMS AND RELATED MATTERS Evaluation of Federal Occupational Safety and Health Programs § 1960.80 Secretary's evaluations of agency occupational safety and health... evaluating an agency's occupational safety and health program. To accomplish this, the Secretary shall...

29 CFR 1960.11 - Evaluation of occupational safety and health performance.

Science.gov (United States)

2010-07-01

... 29 Labor 9 2010-07-01 2010-07-01 false Evaluation of occupational safety and health performance. 1960.11 Section 1960.11 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH... AND HEALTH PROGRAMS AND RELATED MATTERS Administration § 1960.11 Evaluation of occupational safety and...

A reliability evaluation method for NPP safety DCS application software

International Nuclear Information System (INIS)

Li Yunjian; Zhang Lei; Liu Yuan

2014-01-01

In the field of nuclear power plant (NPP) digital i and c application, reliability evaluation for safety DCS application software is a key obstacle to be removed. In order to quantitatively evaluate reliability of NPP safety DCS application software, this paper propose a reliability evaluating method based on software development life cycle every stage's v and v defects density characteristics, by which the operating reliability level of the software can be predicted before its delivery, and helps to improve the reliability of NPP safety important software. (authors)

The International Criticality Safety Benchmark Evaluation Project (ICSBEP)

International Nuclear Information System (INIS)

Briggs, J.B.

2003-01-01

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the United States Department of Energy. The ICSBEP became an official activity of the Organisation for Economic Cooperation and Development (OECD) - Nuclear Energy Agency (NEA) in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Israel, Spain, and Brazil are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled 'International Handbook of Evaluated Criticality Safety Benchmark Experiments.' The 2003 Edition of the Handbook contains benchmark model specifications for 3070 critical or subcritical configurations that are intended for validating computer codes that calculate effective neutron multiplication and for testing basic nuclear data. (author)

Operational safety evaluation for minor reactor accidents

International Nuclear Information System (INIS)

Wang, O.S.

1981-01-01

The purpose of this paper is to address a concern of applying conservatism in analysing minor reactor incidents. A so-called ''conservative'' safety analysis may exaggerate the system responses and result in a reactor scram tripped by the reactor protective system (RPS). In reality, a minor incident may lead the reactor to a new thermal hydraulic steady-state without scram, and the mitigation or termination of the incident may entirely depend on operator actions. An example on a small steamline break evaluation for a pressurized water reactor recently investigated by the staff at the Washington Public Power Supply System is presented to illustrate this point. A safety evaluation using mainly the safety-related systems to be consistent with the conservative assumptions used in the Safety Analysis Report was conducted. For comparison, a realistic analysis was also performed using both the safety- and control-related systems. The analyses were performed using the RETRAN plant simulation computer code. The ''conservative'' safety analysis predicts that the incident can be turned over by the RPS scram trips without operator intervention. However, the realistic analysis concludes that the reactor will reach a new steady-state at a different plant thermal hydraulic condition. As a result, the termination of the incident at this stage depends entirely on proper operator action. On the basis of this investigation it is concluded that, for minor incidents, ''conservative'' assumptions are not necessary, sometimes not justifiable. A realistic investigation from the operational safety point of view is more appropriate. It is essential to highlight the key transient indications for specific incident recognition in the operator training program

Analysis of human factors effects on the safety of transporting radioactive waste materials: Technical report

Energy Technology Data Exchange (ETDEWEB)

Abkowitz, M.D.; Abkowitz, S.B.; Lepofsky, M.

1989-04-01

This report examines the extent of human factors effects on the safety of transporting radioactive waste materials. It is seen principally as a scoping effort, to establish whether there is a need for DOE to undertake a more formal approach to studying human factors in radioactive waste transport, and if so, logical directions for that program to follow. Human factors effects are evaluated on driving and loading/transfer operations only. Particular emphasis is placed on the driving function, examining the relationship between human error and safety as it relates to the impairment of driver performance. Although multi-modal in focus, the widespread availability of data and previous literature on truck operations resulted in a primary study focus on the trucking mode from the standpoint of policy development. In addition to the analysis of human factors accident statistics, the report provides relevant background material on several policies that have been instituted or are under consideration, directed at improving human reliability in the transport sector. On the basis of reported findings, preliminary policy areas are identified. 71 refs., 26 figs., 5 tabs.

A cross-cultural study of organizational factors on safety: Japanese vs. Taiwanese oil refinery plants.

Science.gov (United States)

Hsu, Shang Hwa; Lee, Chun-Chia; Wu, Muh-Cherng; Takano, Kenichi

2008-01-01

This study attempts to identify idiosyncrasies of organizational factors on safety and their influence mechanisms in Taiwan and Japan. Data were collected from employees of Taiwanese and Japanese oil refinery plants. Results show that organizational factors on safety differ in the two countries. Organizational characteristics in Taiwanese plants are highlighted as: higher level of management commitment to safety, harmonious interpersonal relationship, more emphasis on safety activities, higher devotion to supervision, and higher safety self-efficacy, as well as high quality of safety performance. Organizational characteristics in Japanese plants are highlighted as: higher level of employee empowerment and attitude towards continuous improvement, more emphasis on systematic safety management approach, efficient reporting system and teamwork, and high quality of safety performance. The casual relationships between organizational factors and workers' safety performance were investigated using structural equation modeling (SEM). Results indicate that the influence mechanisms of organizational factors in Taiwan and Japan are different. These findings provide insights into areas of safety improvement in emerging countries and developed countries respectively.

Evaluation of the food safety training for food handlers in restaurant operations

OpenAIRE

Park, Sung-Hee; Kwak, Tong-Kyung; Chang, Hye-Ja

2010-01-01

This study examined the extent of improvement of food safety knowledge and practices of employee through food safety training. Employee knowledge and practice for food safety were evaluated before and after the food safety training program. The training program and questionnaires for evaluating employee knowledge and practices concerning food safety, and a checklist for determining food safety performance of restaurants were developed. Data were analyzed using the SPSS program. Twelve restaur...

Factor Analysis and Framework Development for Incorporating Public Trust on Nuclear Safety issues

Energy Technology Data Exchange (ETDEWEB)

Cho, Seongkyung; Lee, Gyebong [The Myongji Univ., Seoul (Korea, Republic of); Lee, Gihyung; Lee, Gyehwi; Jeong, Jina [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-05-15

The Korea Institute of Nuclear Safety (KINS), a regulatory expert organization in charge of nuclear safety in Korea, realized that a more fundamental and systematic analysis of activities is needed to actively meet the greater variety of concerns people have and increase the reliability of the results of regulation. Nuclear safety, a highly specialized field, has previously been discussed primarily from the viewpoint of the engineers who deal with the technology, but now 'public trust in nuclear safety' has to be viewed from the standpoint of the general public and from the socio-cultural perspective. Specific measures must be taken to examine which factors affect public trust and how we can secure and reproduce those factors to gain it. Also, an efficient system for incorporating public trust in nuclear safety must be established. In this study, various case studies were examined to identify the factors that affect public trust in nuclear safety. First, nuclear safety laws and information disclosure systems of major countries were examined by investigating data and conducting in-depth interviews. To explore a public framework concerning nuclear safety, big data of social media were analyzed. Also, Q methodology was used to analyze the risk schemata of the opinion leaders living in areas near nuclear power plants. Several surveys were conducted to analyze the amount of trust the public had in nuclear safety as well as their awareness of nuclear safety issues. Based on these analyses, factors affecting public trust in nuclear safety were extracted, and measures to build systems incorporating public trust in nuclear safety were proposed. This study addresses the public trust in nuclear safety on condition that the safety is ensured technically and mechanically.

Factor Analysis and Framework Development for Incorporating Public Trust on Nuclear Safety issues

International Nuclear Information System (INIS)

Cho, Seongkyung; Lee, Gyebong; Lee, Gihyung; Lee, Gyehwi; Jeong, Jina

2014-01-01

The Korea Institute of Nuclear Safety (KINS), a regulatory expert organization in charge of nuclear safety in Korea, realized that a more fundamental and systematic analysis of activities is needed to actively meet the greater variety of concerns people have and increase the reliability of the results of regulation. Nuclear safety, a highly specialized field, has previously been discussed primarily from the viewpoint of the engineers who deal with the technology, but now 'public trust in nuclear safety' has to be viewed from the standpoint of the general public and from the socio-cultural perspective. Specific measures must be taken to examine which factors affect public trust and how we can secure and reproduce those factors to gain it. Also, an efficient system for incorporating public trust in nuclear safety must be established. In this study, various case studies were examined to identify the factors that affect public trust in nuclear safety. First, nuclear safety laws and information disclosure systems of major countries were examined by investigating data and conducting in-depth interviews. To explore a public framework concerning nuclear safety, big data of social media were analyzed. Also, Q methodology was used to analyze the risk schemata of the opinion leaders living in areas near nuclear power plants. Several surveys were conducted to analyze the amount of trust the public had in nuclear safety as well as their awareness of nuclear safety issues. Based on these analyses, factors affecting public trust in nuclear safety were extracted, and measures to build systems incorporating public trust in nuclear safety were proposed. This study addresses the public trust in nuclear safety on condition that the safety is ensured technically and mechanically

GOLIMUMAB — A NEW TNF α-BLOCKER. THE REVIEW OF THE EFFICACY AND SAFETY EVALUATION RESULTS

Directory of Open Access Journals (Sweden)

R. V. Denisova

2012-01-01

Full Text Available The article represents the results of efficacy and safety evaluation of the human monoclonal antibodies — golimumab, according to the data of international multicenter randomized double-blind placebo-controlled trials, including patients with active stage of rheumatoid arthritis. It was shown, that golimumab was reliably more effective than placebo both when administered hypodermic and intravenous. The safety profile of golimumab is comparable to that of the other tumor necrosis factor alpha blockers. The review also contains information on the 3d phase of golimumab efficacy and safety research in patients with juvenile idiopathic arthritis.

Methods for addressing "innocent bystanders" when evaluating safety of concomitant vaccines.

Science.gov (United States)

Wang, Shirley V; Abdurrob, Abdurrahman; Spoendlin, Julia; Lewis, Edwin; Newcomer, Sophia R; Fireman, Bruce; Daley, Matthew F; Glanz, Jason M; Duffy, Jonathan; Weintraub, Eric S; Kulldorff, Martin

2018-04-01

The need to develop methods for studying the safety of childhood immunization schedules has been recognized by the Institute of Medicine and Department of Health and Human Services. The recommended childhood immunization schedule includes multiple vaccines in a visit. A key concern is safety of concomitant (same day) versus separate day vaccination. This paper addresses a methodological challenge for observational studies using a self-controlled design to investigate the safety of concomitant vaccination. We propose a process for distinguishing which of several concomitantly administered vaccines is responsible for increased risk of an adverse event while adjusting for confounding due to relationships between effect modifying risk factors and concomitant vaccine combinations. We illustrate the approach by re-examining the known increase in risk of seizure 7 to 10 days after measles-mumps-rubella (MMR) vaccination and evaluating potential independent or modifying effects of other vaccines. Initial analyses suggested that DTaP had both an independent and potentiating effect on seizure. After accounting for the relationship between age at vaccination and vaccine combination, there was little evidence for increased risk of seizure with same day administration of DTaP and MMR; incidence rate ratio, 95% confidence interval 1.2 (0.9-1.6), P value = θ.226. We have shown that when using a self-controlled design to investigate safety of concomitant vaccination, it can be critically important to adjust for time-invariant effect modifying risk factors, such as age at time of vaccination, which are structurally related to vaccination patterns due to recommended immunization schedules. Copyright © 2018 John Wiley & Sons, Ltd.

Factors associated with the enactment of safety belt and motorcycle helmet laws.

Science.gov (United States)

Law, Teik Hua; Noland, Robert B; Evans, Andrew W

2013-07-01

It has been shown that road safety laws, such as motorcycle helmet and safety belt laws, have a significant effect in reducing road fatalities. Although an expanding body of literature has documented the effects of these laws on road safety, it remains unclear which factors influence the likelihood that these laws are enacted. This study attempts to identify the factors that influence the decision to enact safety belt and motorcycle helmet laws. Using panel data from 31 countries between 1963 and 2002, our results reveal that increased democracy, education level, per capita income, political stability, and more equitable income distribution within a country are associated with the enactment of road safety laws. © 2012 Society for Risk Analysis.

Construction of Earthquake-Proof Safety Evaluation Methods for Pipes with Wall Thinning

International Nuclear Information System (INIS)

Miyano, Hiroshi; Sekimura, Naoto; Takizawa, Masayuki; Matsumoto, Masaaki

2012-01-01

After the accident at the Fukushima Daiichi Nuclear Power Plant, the extreme importance of 'system safety' evaluation has been recognized. In this study, some fundamental ways of thinking about the concept of 'system safety' for operating plants is shown, and concrete evaluation structures of system safety are proposed. System safety for nuclear power plants and safety assessment for aging plants are constructed. (author)

FFTF railroad tank car Safety Evaluation for Packaging

International Nuclear Information System (INIS)

Carlstrom, R.F.

1995-01-01

This Safety Evaluation for Packaging (SEP) provides evaluations considered necessary to approve transfer of the 8,000 gallon Liquid Waste Tank Car (LWTC) from Fast Flux Test Facility (FFTF) to the 200 Areas. This SEP will demonstrate that the transfer of the LWTC will provide an equivalent degree of safety as would be provided by packages meeting U.S. Department of Transportation (DOT) requirements. This fulfills onsite transportation requirements implemented in the Hazardous Material Packaging and Shipping, WHC-CM-2-14

Application of classification algorithms for analysis of road safety risk factor dependencies.

Science.gov (United States)

Kwon, Oh Hoon; Rhee, Wonjong; Yoon, Yoonjin

2015-02-01

Transportation continues to be an integral part of modern life, and the importance of road traffic safety cannot be overstated. Consequently, recent road traffic safety studies have focused on analysis of risk factors that impact fatality and injury level (severity) of traffic accidents. While some of the risk factors, such as drug use and drinking, are widely known to affect severity, an accurate modeling of their influences is still an open research topic. Furthermore, there are innumerable risk factors that are waiting to be discovered or analyzed. A promising approach is to investigate historical traffic accident data that have been collected in the past decades. This study inspects traffic accident reports that have been accumulated by the California Highway Patrol (CHP) since 1973 for which each accident report contains around 100 data fields. Among them, we investigate 25 fields between 2004 and 2010 that are most relevant to car accidents. Using two classification methods, the Naive Bayes classifier and the decision tree classifier, the relative importance of the data fields, i.e., risk factors, is revealed with respect to the resulting severity level. Performances of the classifiers are compared to each other and a binary logistic regression model is used as the basis for the comparisons. Some of the high-ranking risk factors are found to be strongly dependent on each other, and their incremental gains on estimating or modeling severity level are evaluated quantitatively. The analysis shows that only a handful of the risk factors in the data dominate the severity level and that dependency among the top risk factors is an imperative trait to consider for an accurate analysis. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

International Nuclear Information System (INIS)

Bess, John D.; Briggs, J. Blair; Nigg, David W.

2009-01-01

One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

Reactor safety; Description and evaluation of safety activities in Nordic countries

International Nuclear Information System (INIS)

Wahlstroem, B.; Gunsell, L.

1998-03-01

The report gives a description of safety activities in the nuclear power industry. The study has been carried out as a part of the four year programme in Nordic Safety Research (NKS) which was completed in 1997. The objective of the NKS/RAK-1.1 project 'A survey and an evaluation of safety activities in nuclear power' was to make a broad description of various activities important for safety and to make an assessment of their efficiency. A special consideration was placed on a comparison of practices in Finland and Sweden, and between their nuclear utilities. The study has been divided into two parts, one theoretical part in which a model of the relationships between various activities important for safety has been constructed and one practical part where a total of 62 persons have been interviewed at the authorities, the nuclear utilities and one reactor vendor. To restrict the amount of work two activities, safety analysis and experience feedback, were selected. A few cases connected to incidents at nuclear power plants were discussed in more detail. The report has been structured around a simple model of nuclear safety consisting of the concepts of goals, means and outcomes. This model illustrates the importance of goal formulation, systematic planning and feedback of operational experience as major components in nuclear safety. In assessing organisation and management at authorities and the power utilities there is a clear trend of decentralisation and delegation of authority. The general impression from the study is that the safety activities in Finland and Sweden are efficient and well targeted. The experience from the methodology is favourable and the comparison of practices gives a good ground for a discussion of contents and targeting of safety activities. (EG) activities. (EG)

Evaluation Standard for Safety Coefficient of Roller Compacted Concrete Dam Based on Finite Element Method

Directory of Open Access Journals (Sweden)

Bo Li

2014-01-01

Full Text Available The lack of evaluation standard for safety coefficient based on finite element method (FEM limits the wide application of FEM in roller compacted concrete dam (RCCD. In this paper, the strength reserve factor (SRF method is adopted to simulate gradual failure and possible unstable modes of RCCD system. The entropy theory and catastrophe theory are used to obtain the ultimate bearing resistance and failure criterion of the RCCD. The most dangerous sliding plane for RCCD failure is found using the Latin hypercube sampling (LHS and auxiliary analysis of partial least squares regression (PLSR. Finally a method for determining the evaluation standard of RCCD safety coefficient based on FEM is put forward using least squares support vector machines (LSSVM and particle swarm optimization (PSO. The proposed method is applied to safety coefficient analysis of the Longtan RCCD in China. The calculation shows that RCCD failure is closely related to RCCD interface strength, and the Longtan RCCD is safe in the design condition. Considering RCCD failure characteristic and combining the advantages of several excellent algorithms, the proposed method determines the evaluation standard for safety coefficient of RCCD based on FEM for the first time and can be popularized to any RCCD.

Evaluation of SPT energy for Donut and Safety hammers using CPT measurements in Egypt

Directory of Open Access Journals (Sweden)

Rami M. El-Sherbiny

2013-12-01

Full Text Available Standard Penetration Test (SPT blow counts require correction prior to utilization in soil characterization and determination of properties and behavior. Among the most important corrections is the energy correction required to adjust the blow counts to 60% energy efficiency. However, there are no published data supporting commonly used value in Egypt. This paper presents an evaluation of the energy efficiency of the Donut and Safety hammers commonly used in Egypt and the associated energy correction factor. The energy efficiency is estimated by comparing N-values from the SPT to back-calculated N60 values from the Cone Penetration Test (CPT using well established correlations. Results indicate that the energy efficiency of the Donut hammer based on current practice in Egypt is approximately 50%. Thus, the back-calculated energy correction factor is approximately 0.82. For the Safety hammer, results indicate that the energy efficiency is approximately 60%, and the energy correction factor is approximately 1.0.

Evaluating fuel cycle safety for CITa

International Nuclear Information System (INIS)

Longhurst, G.R.; Reilly, H.J.; Piet, S.J.

1987-01-01

A safety concern in the design of the Compact Ignition Tokamak (CIT) currently being designed in the U. S. is the accidental release of tritium. To evaluate the basis for that concern, an assessment of the risk to the public posed by CIT was conducted that made use of probabilistic risk assessment (PRA) techniques. These include both frequency and consequence elements of risk. This analysis concluded that the tritium systems on the CIT could be designed and operated as planned with negligible safety impact, well within the established guidelines. (author)

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

International Nuclear Information System (INIS)

Richei, A.

1998-01-01

As international studies have shown, accidents in plants are increasingly caused by combinations of technical failures and human errors. Therefore careful investigations of man-machine-interactions to determine human reliability are gaining importance worldwide. Regarding nuclear power plants such investigations are usually carried out within the scope of probabilistic safety assessments. A great number of procedures to evaluate human factors has been developed up to now. However, none of them is able to take into account the whole spectrum of requirements - as for instance transferability of date to other plants, analysis of weak points, and evaluation of cognitive tasks - for a complete and reliable probabilistic safety assessment. Based on an advanced model for a man-machine-system, the Human Error Rate Assessment and Optimizing System (HEROS) and a corresponding expert system of the same name are introduced. This expert system enables the quantification of human error probabilities for plant operator actions on the one hand and is also capable of providing quantitative statements regarding the optimization of man-machine-system in terms of human error probability minimization on the other one. Three relevant evaluation levels, i.e. 'Management Structure', 'Working Environment' and 'Man-Machine-Interface', are derived from a model of the man-machine-system. Linguistic variables are assigned to all performance shaping factors at these levels. These variables are used to establish a rule-based expert system. The knowledge bases of this system are represented by rules. Processing of these rules is carried out by means of the fuzzy set theory, after provision of relevant data for a particular personal action to be evaluated. This procedure enables a simple and effective use of ergonomic studies as the relevant database, which is also transferable to other plants with any design. The expert system consist in total of 16 rule bases in which all ascertainable and

Consensus achievement of leadership, organisational and individual factors that influence safety climate: Implications for nursing management.

Science.gov (United States)

Fischer, Shelly A; Jones, Jacqueline; Verran, Joyce A

2018-01-01

To validate a framework of factors that influence the relationship of transformational leadership and safety climate, and to enable testing of safety chain factors by generating hypotheses regarding their mediating and moderating effects. Understanding the patient safety chain and mechanisms by which leaders affect a strong climate of safety is essential to transformational leadership practice, education, and research. A systematic review of leadership and safety literature was used to develop an organising framework of factors proposed to influence the climate of safety. A panel of 25 international experts in leadership and safety engaged a three-round modified Delphi study with Likert-scored surveys. Eighty per cent of participating experts from six countries were retained to the final survey round. Consensus (>66% agreement) was achieved on 40 factors believed to influence safety climate in the acute care setting. Consensus regarding specific factors that play important roles in an organisation's climate of safety can be reached. Generally, the demonstration of leadership commitment to safety is key to cultivating a culture of patient safety. Transformational nurse leaders should consider and employ all three categories of factors in daily leadership activities and decision-making to drive a strong climate of patient safety. © 2017 John Wiley & Sons Ltd.

Optimized Evaluation System to Athletic Food Safety

OpenAIRE

Shanshan Li

2015-01-01

This study presented a new method of optimizing evaluation function in athletic food safety information programming by particle swarm optimization. The process of food information evaluation function is to automatically adjust these parameters in the evaluation function by self-optimizing method accomplished through competition, which is a food information system plays against itself with different evaluation functions. The results show that the particle swarm optimization is successfully app...

Safety and cost evaluation of nuclear waste management

International Nuclear Information System (INIS)

Vieno, T.; Hautojaervi, A.; Korhonen, R.

1989-11-01

The report introduces the results of the nuclear waste management safety and cost evaluation research carried out in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1984-1988. The emphasis is on the description of the state-of-art of performance and cost evaluation methods. The report describes VTT's most important assessment models. Development, verification and validation of the models has largely taken place within international projects, including the Stripa, HYDROCOIN, INTRACOIN, INTRAVAL, PSACOIN and BIOMOVS projects. Furthermore, VTT's other laboratories are participating in the Natural Analogue Working Group,k the CHEMVAL project and the CoCo group. Resent safety analyses carried out in the Nuclear Engineering Laboratory include a concept feasibility study of spent fuel disposal, safety analyses for the Preliminary Safety Analysis Reports (PSAR's) of the repositories to be constructed for low and medium level operational reactor waste at the Olkiluoto and Loviisa power plants as well as safety analyses of disposal of decommissioning wastes. Appendix 1 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail

Evaluation for nuclear safety-critical software reliability of DCS

International Nuclear Information System (INIS)

Liu Ying

2015-01-01

With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I and C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make besement for evaluating the reliability and safety of DCS. (author)

Effects of the safety factor on ion temperature gradient modes

International Nuclear Information System (INIS)

Wang, A.K.; Dong, J.Q.; Sanuki, H.; Itoh, K.

2003-01-01

A model for the ion temperature gradient (ITG) driven instability is derived from Braginskii magnetohydrodynamic equations of ions. The safety factor q in a toroidal plasma is introduced into the model through the current density J parallel . The effects of q or J parallel on both the ITG instability in k perpendicular and k parallel spectra and the critical stability thresholds are studied. It is shown that the current density // J or the safety factor q plays an important role in stabilizing the ITG instability. (author)

Evaluating the effectiveness of active vehicle safety systems.

Science.gov (United States)

Jeong, Eunbi; Oh, Cheol

2017-03-01

Advanced vehicle safety systems have been widely introduced in transportation systems and are expected to enhance traffic safety. However, these technologies mainly focus on assisting individual vehicles that are equipped with them, and less effort has been made to identify the effect of vehicular technologies on the traffic stream. This study proposed a methodology to assess the effectiveness of active vehicle safety systems (AVSSs), which represent a promising technology to prevent traffic crashes and mitigate injury severity. The proposed AVSS consists of longitudinal and lateral vehicle control systems, which corresponds to the Level 2 vehicle automation presented by the National Highway Safety Administration (NHTSA). The effectiveness evaluation for the proposed technology was conducted in terms of crash potential reduction and congestion mitigation. A microscopic traffic simulator, VISSIM, was used to simulate freeway traffic stream and collect vehicle-maneuvering data. In addition, an external application program interface, VISSIM's COM-interface, was used to implement the AVSS. A surrogate safety assessment model (SSAM) was used to derive indirect safety measures to evaluate the effectiveness of the AVSS. A 16.7-km freeway stretch between the Nakdong and Seonsan interchanges on Korean freeway 45 was selected for the simulation experiments to evaluate the effectiveness of AVSS. A total of five simulation runs for each evaluation scenario were conducted. For the non-incident conditions, the rear-end and lane-change conflicts were reduced by 78.8% and 17.3%, respectively, under the level of service (LOS) D traffic conditions. In addition, the average delay was reduced by 55.5%. However, the system's effectiveness was weakened in the LOS A-C categories. Under incident traffic conditions, the number of rear-end conflicts was reduced by approximately 9.7%. Vehicle delays were reduced by approximately 43.9% with 100% of market penetration rate (MPR). These results

CT fluoroscopy-guided renal tumour cutting needle biopsy: retrospective evaluation of diagnostic yield, safety, and risk factors for diagnostic failure.

Science.gov (United States)

Iguchi, Toshihiro; Hiraki, Takao; Matsui, Yusuke; Fujiwara, Hiroyasu; Sakurai, Jun; Masaoka, Yoshihisa; Gobara, Hideo; Kanazawa, Susumu

2018-01-01

To evaluate retrospectively the diagnostic yield, safety, and risk factors for diagnostic failure of computed tomography (CT) fluoroscopy-guided renal tumour biopsy. Biopsies were performed for 208 tumours (mean diameter 2.3 cm; median diameter 2.1 cm; range 0.9-8.5 cm) in 199 patients. One hundred and ninety-nine tumours were ≤4 cm. All 208 initial procedures were divided into diagnostic success and failure groups. Multiple variables related to the patients, lesions, and procedures were assessed to determine the risk factors for diagnostic failure. After performing 208 initial and nine repeat biopsies, 180 malignancies and 15 benign tumours were pathologically diagnosed, whereas 13 were not diagnosed. In 117 procedures, 118 Grade I and one Grade IIIa adverse events (AEs) occurred. Neither Grade ≥IIIb AEs nor tumour seeding were observed within a median follow-up period of 13.7 months. Logistic regression analysis revealed only small tumour size (≤1.5 cm; odds ratio 3.750; 95% confidence interval 1.362-10.326; P = 0.011) to be a significant risk factor for diagnostic failure. CT fluoroscopy-guided renal tumour biopsy is a safe procedure with a high diagnostic yield. A small tumour size (≤1.5 cm) is a significant risk factor for diagnostic failure. • CT fluoroscopy-guided renal tumour biopsy has a high diagnostic yield. • CT fluoroscopy-guided renal tumour biopsy is safe. • Small tumour size (≤1.5 cm) is a risk factor for diagnostic failure.

International Criticality Safety Benchmark Evaluation Project (ICSBEP) - ICSBEP 2015 Handbook

International Nuclear Information System (INIS)

Bess, John D.

2015-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy (DOE). The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirements and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross-section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span approximately 69000 pages and contain 567 evaluations with benchmark specifications for 4874 critical, near-critical or subcritical configurations, 31 criticality alarm placement/shielding configurations with multiple dose points for each, and 207 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the handbook are benchmark specifications for neutron activation foil and thermoluminescent dosimeter measurements performed at the SILENE critical assembly in Valduc, France as part of a joint venture in 2010 between the US DOE and the French Alternative Energies and Atomic Energy Commission (CEA). A photograph of this experiment is shown on the front cover. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these

Safety climate in university and college laboratories: impact of organizational and individual factors.

Science.gov (United States)

Wu, Tsung-Chih; Liu, Chi-Wei; Lu, Mu-Chen

2007-01-01

Universities and colleges serve to be institutions of education excellence; however, problems in the areas of occupational safety may undermine such goals. Occupational safety must be the concern of every employee in the organization, regardless of job position. Safety climate surveys have been suggested as important tools for measuring the effectiveness and improvement direction of safety programs. Thus, this study aims to investigate the influence of organizational and individual factors on safety climate in university and college laboratories. Employees at 100 universities and colleges in Taiwan were mailed a self-administered questionnaire survey; the response rate was 78%. Multivariate analysis of variance revealed that organizational category of ownership, the presence of a safety manager and safety committee, gender, age, title, accident experience, and safety training significantly affected the climate. Among them, accident experience and safety training affected the climate with practical significance. The authors recommend that managers should address important factors affecting safety issues and then create a positive climate by enforcing continuous improvements.

The impact of WASH-1400 on reactor safety evaluation

International Nuclear Information System (INIS)

Tanguy, P.Y.

1976-01-01

Trends in reactor safety evaluation in France following the publication of WASH-1400 (the Rasmussen Report) are presented. What is called 'the meteorite case' is first schematically presented as follows: WASH-1400 shows nuclear risk equivalent to meteorite risk and reasonable corrections cannot make many orders of magnitude, consequently present safety rules are adequate. The very impact of WASH-1400 on safety approach is then discussed as for: assistance to deterministic safety analysis, introduction of probabilistic safety criteria, acceptable level of risk, and the use of results in research and reactor operating experience

Nursing Student Experiences Regarding Safe Use of Electronic Health Records: A Pilot Study of the Safety and Assurance Factors for EHR Resilience Guides.

Science.gov (United States)

Whitt, Karen J; Eden, Lacey; Merrill, Katreena Collette; Hughes, Mckenna

2017-01-01

Previous research has linked improper electronic health record configuration and use with adverse patient events. In response to this problem, the US Office of the National Coordinator for Health Information Technology developed the Safety and Assurance Factors for EHR Resilience guides to evaluate electronic health records for optimal use and safety features. During the course of their education, nursing students are exposed to a variety of clinical practice settings and electronic health records. This descriptive study evaluated 108 undergraduate and 51 graduate nursing students' ratings of electronic health record features and safe practices, as well as what they learned from utilizing the computerized provider order entry and clinician communication Safety and Assurance Factors for EHR Resilience guide checklists. More than 80% of the undergraduate and 70% of the graduate students reported that they experienced user problems with electronic health records in the past. More than 50% of the students felt that electronic health records contribute to adverse patient outcomes. Students reported that many of the features assessed were not fully implemented in their electronic health record. These findings highlight areas where electronic health records can be improved to optimize patient safety. The majority of students reported that utilizing the Safety and Assurance Factors for EHR Resilience guides increased their understanding of electronic health record features.

A report on human factors in nuclear safety

International Nuclear Information System (INIS)

1983-03-01

Following the Three Mile Island incident of 1979, studies were undertaken by the Atomic Energy Control Board (AECB), in-house and through outside consultants, to address the role of human factors in the regulatory process. This report by the Advisory Committee on Nuclear Safety (ACNS) comments briefly on these studies and offers suggestions which would promote a more formal treatment of human factors by the AECB

Classification analysis of organization factors related to system safety

International Nuclear Information System (INIS)

Liu Huizhen; Zhang Li; Zhang Yuling; Guan Shihua

2009-01-01

This paper analyzes the different types of organization factors which influence the system safety. The organization factor can be divided into the interior organization factor and exterior organization factor. The latter includes the factors of political, economical, technical, law, social culture and geographical, and the relationships among different interest groups. The former includes organization culture, communication, decision, training, process, supervision and management and organization structure. This paper focuses on the description of the organization factors. The classification analysis of the organization factors is the early work of quantitative analysis. (authors)

Evaluation of the reliability concerning the identification of human factors as contributing factors by a computer supported event analysis (CEA)

International Nuclear Information System (INIS)

Wilpert, B.; Maimer, H.; Loroff, C.

2000-01-01

The project's objectives are the evaluation of the reliability concerning the identification of Human Factors as contributing factors by a computer supported event analysis (CEA). CEA is a computer version of SOL (Safety through Organizational Learning). Parts of the first step were interviews with experts from the nuclear power industry and the evaluation of existing computer supported event analysis methods. This information was combined to a requirement profile for the CEA software. The next step contained the implementation of the software in an iterative process of evaluation. The completion of this project was the testing of the CEA software. As a result the testing demonstrated that it is possible to identify contributing factors with CEA validly. In addition, CEA received a very positive feedback from the experts. (orig.) [de

The safety evaluation guide for laboratories and plants a tool for enhancing safety

International Nuclear Information System (INIS)

Lhomme, Veronique; Daubard, Jean-Paul

2013-01-01

The Institute for Radioprotection and Nuclear Safety (IRSN) acts as technical support for the French government Authorities competent in nuclear safety and radiation protection for civil and defence activities. In this frame, the Institute's performs safety assessments of the safety cases submitted by operators to these Authorities for each stage in the life cycle of a nuclear facility, including dismantling operations, which is subjected to a licensing procedure. In the fuel cycle field, this concerns a large variety of facilities. Very often, depending on facilities and on safety cases, safety assessment to be performed is multidisciplinary and involves the supervisor in charge of the facility and several safety experts, particularly to cover the whole set of risks (criticality, exposure to radiation, fire, handling, containment, human and organisational factors...) encountered during facility's operations. Taking these into account, and in order to formalize the assessment process of the fuel cycle facilities, laboratories, irradiators, particle accelerators, under-decommissioning reactors and radioactive waste management, the 'Plants, Laboratories, Transports and Waste Safety' Division of IRSN has developed an internal guide, as a tool: - To present the methodological framework, and possible specificities, for the assessment according to the 'Defence in Depth Concept' (Part 1); - To provide key questions associated to the necessary contradictory technical review of the safety cases (Part 2); - To capitalise on experience on the basis of technical examples (coming from incident reports, previous safety assessments...) demonstrating the questioning (Part 3). The guide is divided in chapters, each dedicated to a type of risk (dissemination of radioactive material, external or internal exposure from ionising radiation, criticality, radiolysis mechanisms, handling operations, earthquake, human or organisational factors...) or to a type

Mathematical models for prediction of safety factors for a simply ...

African Journals Online (AJOL)

From the results obtained, mathematical prediction models were developed using a least square regression analysis for bending, shear and deflection modes of failure considered in the study. The results showed that the safety factors for material, dead and live load are not unique, but they are influenced by safety index ...

Criticality safety evaluations - a open-quotes stalking horseclose quotes for integrated safety assessment

International Nuclear Information System (INIS)

Williams, R.A.

1995-01-01

The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility's criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE

Evaluating and Predicting Patient Safety for Medical Devices With Integral Information Technology

Science.gov (United States)

2005-01-01

323 Evaluating and Predicting Patient Safety for Medical Devices with Integral Information Technology Jiajie Zhang, Vimla L. Patel, Todd R...errors are due to inappropriate designs for user interactions, rather than mechanical failures. Evaluating and predicting patient safety in medical ...the users on the identified trouble spots in the devices. We developed two methods for evaluating and predicting patient safety in medical devices

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

Energy Technology Data Exchange (ETDEWEB)

Yoo, J. K. [Systemix Company, Seoul (Korea, Republic of); Yoon, T. S. [Korea Electric Power Research Institute (Korea, Republic of)

2003-07-01

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.

Influenza vaccines: Evaluation of the safety profile

Science.gov (United States)

Trombetta, Claudia Maria; Gianchecchi, Elena; Montomoli, Emanuele

2018-01-01

ABSTRACT The safety of vaccines is a critical factor in maintaining public trust in national vaccination programs. Vaccines are recommended for children, adults and elderly subjects and have to meet higher safety standards, since they are administered to healthy subjects, mainly healthy children. Although vaccines are strictly monitored before authorization, the possibility of adverse events and/or rare adverse events cannot be totally eliminated. Two main types of influenza vaccines are currently available: parenteral inactivated influenza vaccines and intranasal live attenuated vaccines. Both display a good safety profile in adults and children. However, they can cause adverse events and/or rare adverse events, some of which are more prevalent in children, while others with a higher prevalence in adults. The aim of this review is to provide an overview of influenza vaccine safety according to target groups, vaccine types and production methods. PMID:29297746

The discussion on the qualitative and quantitative evaluation methods for safety culture

International Nuclear Information System (INIS)

Gao Kefu

2005-01-01

The fundamental methods for safely culture evaluation are described. Combining with the practice of the quantitative evaluation of safety culture in Daya Bay NPP, the quantitative evaluation method for safety culture are discussed. (author)

Safety review for human factors engineering and control rooms of nuclear power plants

International Nuclear Information System (INIS)

Yang Mengzhuo

1998-01-01

Safety review for human factors engineering and control rooms of nuclear power plants (NPP) is in a forward position of science and technology, which began at American TMI severe accident and had been implemented in China. The importance and the significance of the safety review are expounded, the requirements of its scope and profundity are explained in detail. In addition, the situation of the technical document system for nuclear safety regulation on human factors engineering and control rooms of NPP in China is introduced briefly, on which the safety review is based

Packaging Evaluation Approach to Improve Cosmetic Product Safety

Directory of Open Access Journals (Sweden)

Benedetta Briasco

2016-09-01

Full Text Available In the Regulation 1223/2009, evaluation of packaging has become mandatory to assure cosmetic product safety. In fact, the safety assessment of a cosmetic product can be successfully carried out only if the hazard deriving from the use of the designed packaging for the specific product is correctly evaluated. Despite the law requirement, there is too little information about the chemical-physical characteristics of finished packaging and the possible interactions between formulation and packaging; furthermore, different from food packaging, the cosmetic packaging is not regulated and, to date, appropriate guidelines are still missing. The aim of this work was to propose a practical approach to investigate commercial polymeric containers used in cosmetic field, especially through mechanical properties’ evaluation, from a safety point of view. First of all, it is essential to obtain complete information about raw materials. Subsequently, using an appropriate full factorial experimental design, it is possible to investigate the variables, like polymeric density, treatment, or type of formulation involved in changes to packaging properties or in formulation-packaging interaction. The variation of these properties can greatly affect cosmetic safety. In particular, mechanical properties can be used as an indicator of pack performances and safety. As an example, containers made of two types of polyethylene with different density, low-density polyethylene (LDPE and high-density polyethylene (HDPE, are investigated. Regarding the substances potentially extractable from the packaging, in this work the headspace solid-phase microextraction method (HSSPME was used because this technique was reported in the literature as suitable to detect extractables from the polymeric material here employed.

A Study on the Allowable Safety Factor of Cut-Slopes for Nuclear Facilities

Energy Technology Data Exchange (ETDEWEB)

Kim, Myung Soo; Yee, Eric [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

In this study, the issues of allowable safety factor design criteria for cut-slopes in nuclear facilities is derived through case analysis, a proposed construction work slope design criteria that provides relatively detailed conditions can be applied in case of the dry season and some unclear parts of slope design criteria be modified in case of the rainy season. This safety factor can be further subdivided into two; normal and earthquake factors, a factor of 1.5 is applied for normal conditions and a factor of 1.2 is applied for seismic conditions. This safety factor takes into consideration the effect of ground water and rainfall conditions. However, no criteria for the case of cut-slope in nuclear facilities and its response to seismic conditions is clearly defined, this can cause uncertainty in design. Therefore, this paper investigates the allowable safety factor for cut-slopes in nuclear facilities, reviews conditions of both local and international cut-slope models and finally suggests an alternative method of analysis. It is expected that the new design criteria adequately ensures the stability of the cut-slope to reflect clear conditions for both the supervising and design engineers.

Squale: evaluation criteria of functioning safety; Squale: criteres d`evaluation de la surete de fonctionnement

Energy Technology Data Exchange (ETDEWEB)

Deswarte, Y; Kaaniche, M [Centre National de la Recherche Scientifique (CNRS), 31 - Toulouse (France). Laboratoire d` Analyse et d` Architecture des Systemes; Corneillie, P [CE2A-DI, 92 - Courbevoie (France); Benoit, P [Matra Transport International, 92 - Montrouge (France)

1998-05-01

The SQUALE (security, safety and quality evaluation for dependable systems) project is part of the ACTS (advanced communications, technologies and services) European program. Its aim is to develop confidence evaluation criteria to test the functioning safety of systems. All industrial sectors that use critical applications (nuclear, railway, aerospace..) are concerned. SQUALE evaluation criteria differ from the classical evaluation methods: they are independent of the application domains and industrial sectors, they take into account the overall functioning safety attributes, and they can progressively change according to the level of severity required. In order to validate the approach and to refine the criteria, a first experiment is in progress with the METEOR automatic underground railway and another will be carried out on a telecommunication system developed by Bouygues company. (J.S.) 15 refs.

75 FR 5536 - Pipeline Safety: Control Room Management/Human Factors, Correction

Science.gov (United States)

2010-02-03

... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Parts...: Control Room Management/Human Factors, Correction AGENCY: Pipeline and Hazardous Materials Safety... following correcting amendments: PART 192--TRANSPORTATION OF NATURAL AND OTHER GAS BY PIPELINE: MINIMUM...

An evaluation of sharp safety blood evacuation devices.

Science.gov (United States)

Ford, Joanna; Phillips, Peter

This article describes an evaluation of three sharp safety blood evacuation devices in seven Welsh NHS boards and the Welsh Blood Service. Products consisted of two phlebotomy needles possessing safety shields and one phlebotomy device with wings, tubing and a retractable needle. The device companies provided the devices and appropriate training. Participating healthcare workers used the safety device instead of the conventional device to sample blood during the evaluation period and each type of device was evaluated in random order. Participants filled in a questionnaire for each type of device and then a further questionnaire comparing the two shielded evacuation needles with each other Results showed that responses to all three products were fairly positive, although each device was not liked by everyone who used it. When the two shielded evacuation devices were compared with each other, most users preferred the device with the shield positioned directly above the needle to the device with the shield at the side. However, in laboratory tests, the preferred device produced more fluid splatter than the other shielded device on activation.

A Web-based Alternative Non-animal Method Database for Safety Cosmetic Evaluations.

Science.gov (United States)

Kim, Seung Won; Kim, Bae-Hwan

2016-07-01

Animal testing was used traditionally in the cosmetics industry to confirm product safety, but has begun to be banned; alternative methods to replace animal experiments are either in development, or are being validated, worldwide. Research data related to test substances are critical for developing novel alternative tests. Moreover, safety information on cosmetic materials has neither been collected in a database nor shared among researchers. Therefore, it is imperative to build and share a database of safety information on toxicological mechanisms and pathways collected through in vivo, in vitro, and in silico methods. We developed the CAMSEC database (named after the research team; the Consortium of Alternative Methods for Safety Evaluation of Cosmetics) to fulfill this purpose. On the same website, our aim is to provide updates on current alternative research methods in Korea. The database will not be used directly to conduct safety evaluations, but researchers or regulatory individuals can use it to facilitate their work in formulating safety evaluations for cosmetic materials. We hope this database will help establish new alternative research methods to conduct efficient safety evaluations of cosmetic materials.

Functionality of road safety devices – identification and analysis of factors

Directory of Open Access Journals (Sweden)

Jeliński Łukasz

2017-01-01

Full Text Available Road safety devices are designed to protect road users from the risk of injury or death. The principal type of restraint is the safety barrier. Deployed on sites with the highest risk of run-off-road accidents, safety barriers are mostly found on bridges, flyovers, central reservations, and on road edges which have fixed obstacles next to them. If properly designed and installed, safety barriers just as other road safety devices, should meet a number of functional features. This report analyses factors which may deteriorate functionality, ways to prevent this from happening and the thresholds for loss of road safety device functionality.

29 CFR 1960.79 - Self-evaluations of occupational safety and health programs.

Science.gov (United States)

2010-07-01

... 29 Labor 9 2010-07-01 2010-07-01 false Self-evaluations of occupational safety and health programs. 1960.79 Section 1960.79 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH... AND HEALTH PROGRAMS AND RELATED MATTERS Evaluation of Federal Occupational Safety and Health Programs...

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

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.

2015-01-01

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

Development of evaluation method for software safety analysis techniques

International Nuclear Information System (INIS)

Huang, H.; Tu, W.; Shih, C.; Chen, C.; Yang, W.; Yih, S.; Kuo, C.; Chen, M.

2006-01-01

Full text: Full text: Following the massive adoption of digital Instrumentation and Control (I and C) system for nuclear power plant (NPP), various Software Safety Analysis (SSA) techniques are used to evaluate the NPP safety for adopting appropriate digital I and C system, and then to reduce risk to acceptable level. However, each technique has its specific advantage and disadvantage. If the two or more techniques can be complementarily incorporated, the SSA combination would be more acceptable. As a result, if proper evaluation criteria are available, the analyst can then choose appropriate technique combination to perform analysis on the basis of resources. This research evaluated the applicable software safety analysis techniques nowadays, such as, Preliminary Hazard Analysis (PHA), Failure Modes and Effects Analysis (FMEA), Fault Tree Analysis (FTA), Markov chain modeling, Dynamic Flowgraph Methodology (DFM), and simulation-based model analysis; and then determined indexes in view of their characteristics, which include dynamic capability, completeness, achievability, detail, signal/ noise ratio, complexity, and implementation cost. These indexes may help the decision makers and the software safety analysts to choose the best SSA combination arrange their own software safety plan. By this proposed method, the analysts can evaluate various SSA combinations for specific purpose. According to the case study results, the traditional PHA + FMEA + FTA (with failure rate) + Markov chain modeling (without transfer rate) combination is not competitive due to the dilemma for obtaining acceptable software failure rates. However, the systematic architecture of FTA and Markov chain modeling is still valuable for realizing the software fault structure. The system centric techniques, such as DFM and Simulation-based model analysis, show the advantage on dynamic capability, achievability, detail, signal/noise ratio. However, their disadvantage are the completeness complexity

Identification and assessment of organisational factors related to the safety of NPPs - State-of-the-Art Report

International Nuclear Information System (INIS)

Baumont, Genevieve; Bourrier, Mathilde; Frischknecht, Albert; Schoenfeld, Isabelle; Weber, Mike J.

1999-09-01

The initiation of this State-of-the-Art Report (SOAR) on Organisational Factors Identification and Assessment comes from operating experience associated with a number of major events world-wide which caused power plants to be shutdown for a significant period of time. Root cause assessments of these events identified weaknesses in organisational factors as contributing to these events. There is general recognition that organisational factors need to be evaluated for their contribution to plant safety performance and risk to prevent their recurrence in events. A special recommendation to create a SOAR was presented in the NEA report on Research Strategies for Human Performance [NEA/CSNI/R(97)24]. Based on this recommendation the Principle Working Group 1 (PWG1) requested, as a top priority, that the Expanded Task Force (ETF) on Human Factors develop a SOAR for the September 1998 meeting. The ETF members were aware that it was a challenging topic. The field of organisational behaviour is not yet fully developed for the nuclear organisation. There is a need to collect and analyse operational and event data from the nuclear environment to determine the safety and risk significance of organisational factors, to identify assessment methods for those factors, and to gain peer review of the results to ensure credibility and acceptability of these methods and possibly their measures. This SOAR reports on the results of the workshop on Organisational Factors Identification and Assessment held in Boettstein Castle, Switzerland, on 14-19 June 1998. Twenty-eight participants from twelve Member countries and Russia represented three different environments: nuclear utilities; regulatory bodies and inspectorates; and the research and academic community. The various approaches discussed in the SOAR reflect the perspective of these entities. The SOAR addresses the following topics: - identification of organisational factors; - identification of methods for the evaluation of

Safety evaluation model of urban cross-river tunnel based on driving simulation.

Science.gov (United States)

Ma, Yingqi; Lu, Linjun; Lu, Jian John

2017-09-01

Currently, Shanghai urban cross-river tunnels have three principal characteristics: increased traffic, a high accident rate and rapidly developing construction. Because of their complex geographic and hydrological characteristics, the alignment conditions in urban cross-river tunnels are more complicated than in highway tunnels, so a safety evaluation of urban cross-river tunnels is necessary to suggest follow-up construction and changes in operational management. A driving risk index (DRI) for urban cross-river tunnels was proposed in this study. An index system was also constructed, combining eight factors derived from the output of a driving simulator regarding three aspects of risk due to following, lateral accidents and driver workload. Analytic hierarchy process methods and expert marking and normalization processing were applied to construct a mathematical model for the DRI. The driving simulator was used to simulate 12 Shanghai urban cross-river tunnels and a relationship was obtained between the DRI for the tunnels and the corresponding accident rate (AR) via a regression analysis. The regression analysis results showed that the relationship between the DRI and the AR mapped to an exponential function with a high degree of fit. In the absence of detailed accident data, a safety evaluation model based on factors derived from a driving simulation can effectively assess the driving risk in urban cross-river tunnels constructed or in design.

The use of living PSA in safety management, a procedure developed in the nordic project ''safety evaluation, NKS/SIK-1''

International Nuclear Information System (INIS)

Johanson, G.; Holmberg, J.

1994-01-01

The essential objective with the development of a living PSA concept is to bring the use of the plant specific PSA model out to the daily safety work to allow operational risk experience feedback and to increase the risk awareness of the intended users. This paper will present results of the Nordic project ''Safety Evaluation, NKS/SIK-1''. The SIK-1 project has defined and demonstrated the practical use of living PSA for safety evaluation and for identification of possible improvements in operational safety. Subjects discussed in this paper are dealing with the practical implementation and use of PSA to make proper safety related decisions and evaluation. (author). 24 refs, 1 fig., 1 tab

Research on review technology for three key safety factors of periodic safety review (PSR) and its application to Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Xu Shoulv; Yao Weida; Dou Yikang; Lin Shaoxuan; Cao Yenan; Zhou Quanfu; Zheng Jiong; Zhang Ming

2009-04-01

In 2001, after 10 years' operation, Qinshan Nuclear Power Plant (Q1) started to carry out periodic safety review (PSR) based on a nuclear safety guideline, Periodic Safety Review for Operational Nuclear Power Plants (HAF0312), issued by National Nuclear Safety Administration of China (NNSA). Entrusted by the owner of Q1, Shanghai Nuclear Engineering Research and Design Institute (SNERDI) implemented reviews of three key safety factors including safety analysis, equipment qualification and ageing. PSR was a challenging work in China at that time and through three years' research and practice, SNERDI summarized a systematic achievement for the review including review methodology, scoping, review contents and implementation steps, etc.. During the process of review for the three safety factors, totally 148 review reports and 341 recommendations for corrections were submitted to Q1. These reports and recommendations have provided guidance for correction actions as follow-up of PSR. This paper focuses on technical aspects to carry out PSR for the above-mentioned three safety factors, including review scoping, contents, methodology and main steps. The review technology and relevant experience can be taken for reference for other NPPs to carry out PSR. (authors)

Modeling the factors affecting unsafe behavior in the construction industry from safety supervisors' perspective.

Science.gov (United States)

Khosravi, Yahya; Asilian-Mahabadi, Hassan; Hajizadeh, Ebrahim; Hassanzadeh-Rangi, Narmin; Bastani, Hamid; Khavanin, Ali; Mortazavi, Seyed Bagher

2014-01-01

There can be little doubt that the construction is the most hazardous industry in the worldwide. This study was designed to modeling the factors affecting unsafe behavior from the perspective of safety supervisors. The qualitative research was conducted to extract a conceptual model. A structural model was then developed based on a questionnaire survey (n=266) by two stage Structural Equation Model (SEM) approach. An excellent confirmed 12-factors structure explained about 62% of variances unsafe behavior in the construction industry. A good fit structural model indicated that safety climate factors were positively correlated with safety individual factors (Pconstruction workers' engagement in safe or unsafe behavior. In order to improve construction safety performance, more focus on the workplace condition is required.

Human and organization factors: engineering operating safety into offshore structures

International Nuclear Information System (INIS)

Bea, Robert G.

1998-01-01

History indicates clearly that the safety of offshore structures is determined primarily by the humans and organizations responsible for these structures during their design, construction, operation, maintenance, and decommissioning. If the safety of offshore structures is to be preserved and improved, then attention of engineers should focus on to how to improve the reliability of the offshore structure 'system,' including the people that come into contact with the structure during its life-cycle. This article reviews and discusss concepts and engineering approaches that can be used in such efforts. Two specific human factor issues are addressed: (1) real-time management of safety during operations, and (2) development of a Safety Management Assessment System to help improve the safety of offshore structures

Evaluation of reliability assurance approaches to operational nuclear safety

International Nuclear Information System (INIS)

Mueller, C.J.; Bezella, W.A.

1984-01-01

This report discusses the results of research to evaluate existing and/or recommended safety/reliability assurance activities among nuclear and other high technology industries for potential nuclear industry implementation. Since the Three Mile Island (TMI) accident, there has been increased interest in the use of reliability programs (RP) to assure the performance of nuclear safety systems throughout the plant's lifetime. Recently, several Nuclear Regulatory Commission (NRC) task forces or safety issue review groups have recommended RPs for assuring the continuing safety of nuclear reactor plants. 18 references

The Relationship Between Antecedent and Consequence Factors with Safety Behaviour in PT.X

OpenAIRE

Fitriani, Apris; Nawawiwetu, Erwin Dyah

2017-01-01

Background : Safety behaviour is an act worker to minimize the possibility of accidents in workplace. Based on the Antecedents-Behaviour-Consequence (ABC) theory, safety behaviour of worker related with the antecedent and consequence factors. Purpose : The purpose of this research was to study the association between antecedent and consequence factors with safety behaviour of workers in Ring Frame Unit Spinning II PT. X. Methods : This was an observational descriptive research with cross sect...

A research framework of organizational factors on safety in the Republic of Korea

International Nuclear Information System (INIS)

Kwang Seok Lee

1997-01-01

Korean nuclear society is yet unfamiliar with the topic, 'organizational factors on safety', while having shown lots of accomplishments in the area of physical and human factors on safety. However, recent large-scale accidents in other technological areas illustrate the importance of managing organization factors on safety. Recently Korea Atomic Energy Research Institute (KAERI) started paying attention to this topic and is trying to establish a future research framework of organizational factors on safety. This paper tries to explain overall direction of the framework. Our framework, as managing organizational factors on safety, considers two kinds of areas: design of management systems, which implies a feed-forward system including organizational models; and operation of those systems, which implies a feedback system including management information and implementation systems. Our framework also considers the evolution stage of a management system. Management systems evolve from visibility stage to optimization stage. To optimize a management system, we should be able to control the system. To control the system, we should be able to see how the system is going. In addition, this paper tries to share some experience of KAERI on how organizational structure and culture affects organizational performance in R and D perspective. (author). 2 refs, 1 fig

Decommissioning: Regulatory activities and identification of key organizational and human factors safety issues

International Nuclear Information System (INIS)

Durbin, N.E.; Melber, B.D.; Lekberg, A.

2001-12-01

In the late 1990's the Swedish government decided to shut down Unit 1 of the Barsebaeck nuclear power plant. This report documents some of the efforts made by the Swedish Nuclear Power Inspectorate (SKI) to address human factors and organizational issues in nuclear safety during decommissioning of a nuclear facility. This report gives a brief review of the background to the decommissioning of Barsebaeck 1 and points out key safety issues that can arise during decommissioning. The main regulatory activities that were undertaken were requirements that the plant provide special safety reports on decommissioning focusing on first, the operation of both units until closure of Unit 1 and second, the operation of Unit 2 when Unit 1 was closed. In addition, SKI identified areas that might be affected by decommissioning and called these areas out for special attention. With regard to these areas of special attention, SKI required that the plant provide monthly reports on changing and emerging issues as well as self-assessments of the areas to be addressed in the special safety reports. Ten key safety issues were identified and evaluated with regard to different stages of decommissioning and with regard to the actions taken by Barsebaeck. Some key conclusions from SKI's experience in regulating a decommissioning nuclear power plant conclude the report

Human and organizational factors in nuclear safety

International Nuclear Information System (INIS)

Garcia, A.; Barrientos, M.; Gil, B.

2015-01-01

Nuclear installations are socio technical systems where human and organizational factors, in both utilities and regulators, have a significant impact on safety. Three Mile Island (TMI) accident, original of several initiatives in the human factors field, nevertheless became a lost opportunity to timely acquire lessons related to the upper tiers of the system. Nowadays, Spanish nuclear installations have integrated in their processes specialists and activities in human and organizational factors, promoted by the licensees After many years of hard work, Spanish installations have achieved a better position to face new challenges, such as those posed by Fukushima. With this experience, only technology-centered action plan would not be acceptable, turning this accident in yet another lost opportunity. (Author)

Evaluation of atmospheric dispersion/consequence models supporting safety analysis

International Nuclear Information System (INIS)

O'Kula, K.R.; Lazaro, M.A.; Woodard, K.

1996-01-01

Two DOE Working Groups have completed evaluation of accident phenomenology and consequence methodologies used to support DOE facility safety documentation. The independent evaluations each concluded that no one computer model adequately addresses all accident and atmospheric release conditions. MACCS2, MATHEW/ADPIC, TRAC RA/HA, and COSYMA are adequate for most radiological dispersion and consequence needs. ALOHA, DEGADIS, HGSYSTEM, TSCREEN, and SLAB are recommended for chemical dispersion and consequence applications. Additional work is suggested, principally in evaluation of new models, targeting certain models for continued development, training, and establishing a Web page for guidance to safety analysts

Evaluating the safety impact of adaptive cruise control in traffic oscillations on freeways.

Science.gov (United States)

Li, Ye; Li, Zhibin; Wang, Hao; Wang, Wei; Xing, Lu

2017-07-01

Adaptive cruise control (ACC) has been considered one of the critical components of automated driving. ACC adjusts vehicle speeds automatically by measuring the status of the ego-vehicle and leading vehicle. Current commercial ACCs are designed to be comfortable and convenient driving systems. Little attention is paid to the safety impacts of ACC, especially in traffic oscillations when crash risks are the highest. The primary objective of this study was to evaluate the impacts of ACC parameter settings on rear-end collisions on freeways. First, the occurrence of a rear-end collision in a stop-and-go wave was analyzed. A car-following model in an integrated ACC was developed for a simulation analysis. The time-to-collision based factors were calculated as surrogate safety measures of the collision risk. We also evaluated different market penetration rates considering that the application of ACC will be a gradual process. The results showed that the safety impacts of ACC were largely affected by the parameters. Smaller time delays and larger time gaps improved safety performance, but inappropriate parameter settings increased the collision risks and caused traffic disturbances. A higher reduction of the collision risk was achieved as the ACC vehicle penetration rate increased, especially in the initial stage with penetration rates of less than 30%. This study also showed that in the initial stage, the combination of ACC and a variable speed limit achieved better safety improvements on congested freeways than each single technique. Copyright © 2017 Elsevier Ltd. All rights reserved.

Periodic safety re-evaluations in NPPs in EC member states, Finland and Sweden

International Nuclear Information System (INIS)

1990-01-01

The work on periodic safety re-evaluations summarized in this report was performed by a Task Force of the CEC Working Group on the Safety of Thermal Reactors. The periodic safety re-evaluations under review in this study were those that are carried out in addition to other reviews which represent the primary means of safety assurance. The periodic safety re-evaluation is broader and more comprehensive in nature. The cumulative effects of experience (national and international), advances in knowledge and analysis techniques, improvements in safety standards and operating practices, overall effects of plant ageing, and the totality of all modifications over the period in question need to be taken into account. All countries have recognized the value of such periodic reviews, and licensees, either as a regulatory requirement or as a voluntary action, are carrying them out. The scope and contents of each country's review showed many similarities of approach, any differences being explained by the age and type of reactor in operation. Many similarities emerged in the topics selected for re-evaluation and in the approach to re-evaluation itself. The overall conclusion was that while approaches may differ in some respects, for practical purposes comparable levels of safety are achieved in the periodic safety re-evaluation of nuclear power plants

Organisational factors. Their definition and influence on nuclear safety. Final report

International Nuclear Information System (INIS)

Baumont, G.; Wahlstroem, B.; Sola, R.; Williams, J.; Frischknecht, A.; Wilpert, B.; Rollenhagen, C.

2000-12-01

The importance of organisational factors in the operational safety and efficiency of nuclear power plants (NPP) has been recognised by many organisations around the world. Despite this recognition, however, there are as yet very few methods by which organisational factors can be systematically assessed and improved. The majority of research efforts applied so far have tended to be modest and scattered. The ORFA project was created as a remedy to these problems. The objective of the project is to create a better understanding of how organisation and management factors influence nuclear safety. A key scientific objective of the project is to identify components of a theoretical framework, which would help in understanding the relationships between organisational factors and nuclear safety. Three work packages were planned. First, a review of literature listed out the identified factors and methods for assessing them. Then, a draft version of the present report was prepared to clarify the environment context and the main issues of the topics. This draft was discussed at the ORFA seminar in Madrid 21-22 October 1999. During the seminar views and comments were collected on preliminary results of the project. Finally, this information has been integrated in the present and other reports and will be used to give further guidance to the European Commission in the development of forthcoming research programmes in the field. The project has addressed nuclear safety taking a broad perspective, which reflected and took into account the views of senior NPP management and regulators. The questions discussed during the project have been: how can organisational factors be included in safety assessments, how can good and bad operational practices be identified, which methods can be used for detecting weak signals of deteriorating performance, how should incidents be analysed with respect to organisational factors to give the largest learning benefit, how can data on organisational

Periodic safety review of the HTR-10 safety analysis

International Nuclear Information System (INIS)

Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

2015-01-01

Designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first modular High Temperature Gas-cooled Reactor (HTGR) in China. According to the nuclear safety regulations of China, the periodic safety review (PSR) of the HTR-10 was initiated by INET after approved by the National Nuclear Safety Administration (NNSA) of China. Safety analysis of the HTR-10 is one of the key safety factors of the PSR. In this paper, the main contents in the review of safety analysis are summarized; meanwhile, the internal evaluation on the review results is presented by INET. (authors)

Organizational factors influencing improvements in safety

International Nuclear Information System (INIS)

Marcus, A.; Nichols, M.L.; Olson, J.; Osborn, R.; Thurber, J.

1991-01-01

Results of conceptual and empirical research conducted by this research team, and published in NUREG-CR 5437, suggested that processes of organizational problem solving and learning provide a promising area for understanding improvement in safety-related performance in nuclear power plants. In this paper the authors describe the way in which they have built upon that work and gone much further in empirically examining a range of potentially important organizational factors related to safety. The paper describes (1) overall trends in plant performance over time on the Nuclear Regulatory Commission performance indicators, (2) the major elements in the conceptual framework guiding the current work, which seeks among other things to explain those trends, (3) the specific variables used as measures of the central concepts, (4) the results to date of the quantitative empirical work and qualitative work in progress, and (5) conclusions from the research

Longitudinal safety evaluation of electric vehicles with the partial wireless charging lane on freeways.

Science.gov (United States)

Li, Ye; Wang, Wei; Xing, Lu; Fan, Qi; Wang, Hao

2018-02-01

As an environment friendly transportation mode, the electric vehicle (EV) has drawn an increasing amount of attention from governments, vehicle manufactories and researchers recently. One of the biggest issue impeding EV's popularization associates with the charging process. The wireless charging lane (WCL) has been proposed as a convenient charging facility for EVs. Due to the high costs, the application of WCL on the entire freeways is impractical in the near future, while the partial WCL (PWCL) may be a feasible solution. This study aims to evaluate longitudinal safety of EVs with PWCL on freeways based on simulations. The simulation experiments are firstly designed, including deployment of PWCL on freeways and distribution of state of charge (SOC) of EVs. Then, a vehicle behavior model for EVs is proposed based on the intelligent driver model (IDM). Two surrogate safety measures, derived from time-to-collision (TTC), are utilized as indicators for safety evaluations. Sensitivity analysis is also conducted for related factors. Results show that the distribution of EVs' SOC significantly affect longitudinal safety when the PWCL is utilized. The low SOC in traffic consisting of EVs has the negative effect on longitudinal safety. The randomness and incompliance of EV drivers worsens the safety performance. The sensitivity analysis indicates that the larger maximum deceleration rate results in the higher longitudinal crash risks of EVs, while the length of PWCL has no monotonous effect. Different TTC thresholds also show no impact on results. A case study shows the consistent results. Based on the findings, several suggestions are discussed for EVs' safety improvement. Results of this study provide useful information for freeway safety when EVs are applied in the future. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

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

1985-02-01

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

Identifying the Critical Factors Affecting Safety Program Performance for Construction Projects within Pakistan Construction Industry

Directory of Open Access Journals (Sweden)

Zubair Ahmed Memon

2013-04-01

Full Text Available Many studies have shown that the construction industry one of the most hazardous industries with its high rates of fatalities and injuries and high financial losses incurred through work related accident. To reduce or overcome the safety issues on construction sites, different safety programs are introduced by construction firms. A questionnaire survey study was conducted to highlight the influence of the Construction Safety Factors on safety program implementation. The input from the questionnaire survey was analyzed by using AIM (Average Index Method and rank correlation test was conducted between different groups of respondents to measure the association between different groups of respondent. The finding of this study highlighted that management support is the critical factor for implementing the safety program on projects. From statistical test, it is concluded that all respondent groups were strongly in the favor of management support factor as CSF (Critical Success Factor. The findings of this study were validated on selected case studies. Results of the case studies will help to know the effect of the factors on implementing safety programs during the execution stage.

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

Energy Technology Data Exchange (ETDEWEB)

Yu, Jae Kook; Ahn, Nam Sung [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jae, Moo Sung [Hanyang Univ., Seoul (Korea, Republic of)

2004-02-01

The purpose of this study is to develop a system dynamics model for the assessment of the organizational and human factors in a nuclear power plant which contribute to nuclear safety. Previous studies can be classified into two major approaches. One is the engineering approach using tools such as ergonomics and Probability Safety Assessment (PSA). The other is the socio-psychology approach. Both have contributed to find organizational and human factors and to present guidelines to lessen human error in plants. However, since these approaches assume that the relationship among factors is independent they do not explain the interactions among the factors or variables in nuclear power plants. To overcome these restrictions, a system dynamics model, which can show cause and effect relationships among factors and quantify the organizational and human factors, has been developed. Handling variables such as the degree of leadership, the number of employees, and workload in each department, users can simulate various situations in nuclear power plant organization. Through simulation, users can get insights to improve safety in plants and to find managerial tools in both organizational and human factors.

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

International Nuclear Information System (INIS)

Yu, Jae Kook; Ahn, Nam Sung; Jae, Moo Sung

2004-01-01

The purpose of this study is to develop a system dynamics model for the assessment of the organizational and human factors in a nuclear power plant which contribute to nuclear safety. Previous studies can be classified into two major approaches. One is the engineering approach using tools such as ergonomics and Probability Safety Assessment (PSA). The other is the socio-psychology approach. Both have contributed to find organizational and human factors and to present guidelines to lessen human error in plants. However, since these approaches assume that the relationship among factors is independent they do not explain the interactions among the factors or variables in nuclear power plants. To overcome these restrictions, a system dynamics model, which can show cause and effect relationships among factors and quantify the organizational and human factors, has been developed. Handling variables such as the degree of leadership, the number of employees, and workload in each department, users can simulate various situations in nuclear power plant organization. Through simulation, users can get insights to improve safety in plants and to find managerial tools in both organizational and human factors

Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

International Nuclear Information System (INIS)

Jeong, Hae Yong; Kwon, Young Min; Kim, Tae Woon; Park, Soo Yong; Suk, Soo Dong; Lee, Kwi Lim; Lee, Yong Bum; Chang, Won Pyo; Ha, Kwi Seok; Hahn, Sang Hoon

2010-07-01

Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

Criticality Safety Evaluation for the TACS at DAF

Energy Technology Data Exchange (ETDEWEB)

Percher, C. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2011-06-10

Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, Guidance for Nuclear Criticality Safety Engineer Training and Qualification. This document is a criticality safety evaluation of the training activities and operations associated with HS-3201-P, Nuclear Criticality 4-Day Training Course (Practical). This course was designed to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program1. The hands-on, or laboratory, portion of the course will utilize the Training Assembly for Criticality Safety (TACS) and will be conducted in the Device Assembly Facility (DAF) at the Nevada Nuclear Security Site (NNSS). The training activities will be conducted by Lawrence Livermore National Laboratory following the requirements of an Integrated Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of an LLNL Certified Fissile Material Handler.

Key Factors Affecting Construction Safety Performance in Developing Countries: Evidence from Cambodia

Directory of Open Access Journals (Sweden)

Serdar Durdyev

2017-12-01

Full Text Available Although proper safety management in construction is of utmost importance; anecdotal evidence suggests that safety is not adequately considered in many developing countries. This paper considers the key variables affecting construction safety performance in Cambodia. Using an empirical questionnaire survey targeting local construction professionals, respondents were invited to rate the level of importance of 30 variables identified from the seminal literature. The data set was subjected to factor analysis. Correlations between the variables show that five key factors underlie the challenges facing the local industry; management and organisation, resources, site management, cosmetic and workforce. It is found that the forefront construction professionals (top management and government authorities should take more responsibilities for further improvements in safety performance on project sites. Findings and recommendations of this study may be useful to construction professional who are seeking ways to improve safety records in developing countries.

Safety culture management: The importance of organizational factors

International Nuclear Information System (INIS)

Haber, S.B.; Shurberg, D.A.; Jacobs, R.; Hofmann, D.

1995-01-01

The concept of safety culture has been used extensively to explain the underlying causes of performance based events, both positive and negative, across the nuclear industry. The work described in this paper represents several years of effort to identify, define and assess the organizational factors important to safe performance in nuclear power plants (NPPs). The research discussed in this paper is primarily conducted in support of the US Nuclear Regulatory Commission's (NRC) efforts in understanding the impact of organizational performance on safety. As a result of a series of research activities undertaken by numerous NRC contractors, a collection of organizational dimensions has been identified and defined. These dimensions represent what is believed to be a comprehensive taxonomy of organizational elements that relate to the safe operation of nuclear power plants. Techniques were also developed by which to measure these organizational dimensions, and include structured interview protocols, behavioral checklists, and behavioral anchored rating scales (BARS). Recent efforts have focused on devising a methodology for the extraction of information related to the identified organizational dimensions from existing NRC documentation. This type of effort would assess the applicability of the organizational dimensions to existing NRC inspection and evaluation reports, refine the organizational dimensions previously developed so they are more relevant to the task of retrospective analysis, and attempt to rate plants based on the review of existing NRC documentation using the techniques previously developed for the assessment of organizational dimensions

Nuclear safety and human factors: the French factory of expertise

International Nuclear Information System (INIS)

Rolina, G.

2009-01-01

The French regulation of the nuclear safety is based on the maintaining of a deep technical dialogue between the nuclear safety authority, the I.R.S.N. (Institute of radiation protection and nuclear safety) and the nuclear operators. This type of risk management is called 'french coking' by the Anglo-Saxons, followers of stricter regulatory approach, more readable by the civil society. This technical dialogue is not without quality, especially in the field of human and organizational factors where it allows to improve the know how situation that stays incomplete. (N.C.)

Workers’ Age and the Impact of Psychological Factors on the Perception of Safety at Construction Sites

Directory of Open Access Journals (Sweden)

Muhammad Dawood Idrees

2017-05-01

Full Text Available The safety of construction workers is always a major concern at construction sites as the construction industry is inherently dangerous with many factors influencing worker safety. Several studies concluded that psychological factors such as workload, organizational relationships, mental stress, job security, and job satisfaction have significant effects on workers’ safety. However, research on psychological factors that are characteristic of different age groups have been limited. The aim of this study was to examine the impact of psychological factors on the perception of worker safety for two different age groups. After an extensive literature review, different psychological factors were identified, and a hypothetical research model was developed based on psychological factors that could affect workers’ perception of safety. A survey instrument was developed, and data were collected from seven different construction sites in Pakistan. Structural equation modeling (SEM was employed to test the hypothetical model for both age groups. The results revealed that workload and job satisfaction are significantly dominant factors on workers’ perception of safety in older workers, whereas organizational relationships, mental stress, and job security are dominant factors for younger workers at construction sites.

Development of 4S and related technologies. (3) Statistical evaluation of safety performance of 4S on ULOF event

International Nuclear Information System (INIS)

Ishii, Kyoko; Matsumiya, Hisato; Horie, Hideki; Miyagi, Kazumi

2009-01-01

The purpose of this work is to evaluate quantitatively and statistically the safety performance of Super-Safe, Small, and Simple reactor (4S) by analyzing with ARGO code, a plant dynamics code for a sodium-cooled fast reactor. In this evaluation, an Anticipated Transient Without Scram (ATWS) is assumed, and an Unprotected Loss of Flow (ULOF) event is selected as a typical ATWS case. After a metric concerned with safety design is defined as performance factor a Phenomena Identification Ranking Table (PIRT) is produced in order to select the plausible phenomena that affect the metric. Then a sensitivity analysis is performed for the parameters related to the selected plausible phenomena. Finally the metric is evaluated with statistical methods whether it satisfies the given safety acceptance criteria. The result is as follows: The Cumulative Damage Fraction (CDF) for the cladding is defined as a metric, and the statistical estimation of the one-sided upper tolerance limit of 95 percent probability at a 95 percent confidence level in CDF is within the safety acceptance criterion; CDF < 0.1. The result shows that the 4S safety performance is acceptable in the ULOF event. (author)

An Evaluation Tool for Agricultural Health and Safety Mobile Applications.

Science.gov (United States)

Reyes, Iris; Ellis, Tammy; Yoder, Aaron; Keifer, Matthew C

2016-01-01

As the use of mobile devices and their software applications, or apps, becomes ubiquitous, use amongst agricultural working populations is expanding as well. The smart device paired with a well-designed app has potential for improving workplace health and safety in the hands of those who can act upon the information provided. Many apps designed to assess workplace hazards and implementation of worker protections already exist. However, the abundance and diversity of such applications also presents challenges regarding evaluation practices and assignation of value. This is particularly true in the agricultural workspace, as there is currently little information on the value of these apps for agricultural safety and health. This project proposes a framework for developing and evaluating apps that have potential usefulness in agricultural health and safety. The evaluation framework is easily transferable, with little modification for evaluation of apps in several agriculture-specific areas.

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

Science.gov (United States)

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

2017-06-01

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

Evaluation of pedestrian safety at intersections: A theoretical framework based on pedestrian-vehicle interaction patterns.

Science.gov (United States)

Ni, Ying; Wang, Menglong; Sun, Jian; Li, Keping

2016-11-01

Pedestrians are the most vulnerable road users, and pedestrian safety has become a major research focus in recent years. Regarding the quality and quantity issues with collision data, conflict analysis using surrogate safety measures has become a useful method to study pedestrian safety. However, given the inequality between pedestrians and vehicles in encounters and the multiple interactions between pedestrians and vehicles, it is insufficient to simply use the same indicator(s) or the same way to aggregate indicators for all conditions. In addition, behavioral factors cannot be neglected. To better use information extracted from trajectories for safety evaluation and pay more attention on effects of behavioral factors, this paper develops a more sophisticated framework for pedestrian conflict analysis that takes pedestrian-vehicle interactions into consideration. A concept of three interaction patterns has been proposed for the first time, namely "hard interaction," "no interaction," and "soft-interaction." Interactions have been categorized under one of these patterns by analyzing profiles of speed and conflict indicators during the whole interactive processes. In this paper, a support vector machine (SVM) approach has been adopted to classify severity levels for a dataset including 1144 events extracted from three intersections in Shanghai, China, followed by an analysis of variable importance. The results revealed that different conflict indicators have different contributions to indicating the severity level under various interaction patterns. Therefore, it is recommended either to use specific conflict indicators or to use weighted indicator aggregation for each interaction pattern when evaluating pedestrian safety. The implementation has been carried out at the fourth crosswalk, and the results indicate that the proposed method can achieve a higher accuracy and better robustness than conventional methods. Furthermore, the method is helpful for better

Criticality safety evaluations - a {open_quotes}stalking horse{close_quotes} for integrated safety assessment

Energy Technology Data Exchange (ETDEWEB)

Williams, R.A. [Westinghouse Electric Corp., Columbia, SC (United States)

1995-12-31

The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility`s criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE.

Fuel Receiving and Storage Station. Nuclear Regulatory Commission's safety evaluation report

International Nuclear Information System (INIS)

1976-01-01

The safety evaluation report covers design of structures, components, equipment, and systems; nuclear criticality safety; radiological safety; accident analysis; conduct of operations; quality assurance; common defense and security; financial qualifications; financial protection and indemnity requirements; and technical specifications

Development of safety evaluation guidelines for base-isolated buildings in Japan

International Nuclear Information System (INIS)

Aoyama, Hiroyuki

1989-01-01

This paper describes the safety evaluation guidelines and the review process for non-nuclear base-isolated buildings proposed for construction in Japan. The paper discusses the guidelines application for two types of soil: hard soil and intermediate soil (soft soil was excluded.); safety evaluation items included in the level C design review; and safety margin of base isolation. Lessons learned through these design review efforts have potential applicability to design of seismic base isolation for nuclear power plants

Review of studies on criticality safety evaluation and criticality experiment methods

International Nuclear Information System (INIS)

Naito, Yoshitaka; Yamamoto, Toshihiro; Misawa, Tsuyoshi; Yamane, Yuichi

2013-01-01

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS. (author)

Relationships between road safety, safety measures and external factors : a scan of the literature in view of model development and topics for further research.

NARCIS (Netherlands)

Churchill, T. & Norden, Y. van

2010-01-01

The purpose of this literature scan is to examine where literature on the effect of external factors and road safety measures on road safety exists and where it is lacking. This scan will help us to decide which factors to include in a comprehensive road safety model as SWOV is working on, and at

Addressing the fundamental issues in reliability evaluation of passive safety of AP1000 for a comparison with active safety of PWR

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Yang Ming

2013-01-01

Passive safety systems adopted in advanced Pressurized Water Reactor (PWR), such as AP1000 and EPR, should attain higher reliability than the existing active safety systems of the conventional PWR. The objective of this study is to discuss the fundamental issues relating to the reliability evaluation of AP1000 passive safety systems for a comparison with the active safety systems of conventional PWR, based on several aspects. First, comparisons between conventional PWR and AP1000 are made from the both aspects of safety design and cost reduction. The main differences between these PWR plants exist in the configurations of safety systems: AP1000 employs the passive safety system while reducing the number of active systems. Second, the safety of AP1000 is discussed from the aspect of severe accident prevention in the event of large break loss of coolant accidents (LOCA). Third, detailed fundamental issues on reliability evaluation of AP1000 passive safety systems are discussed qualitatively by using single loop models of safety systems of both PWRs plants. Lastly, methodology to conduct quantitative estimation of dynamic reliability for AP1000 passive safety systems in LOCA condition is discussed, in order to evaluate the reliability of AP1000 in future by a success-path-based reliability analysis method (i.e., GO-FLOW). (author)

Safety Criteria and Standards for Bearing Capacity of Foundation

Directory of Open Access Journals (Sweden)

Yanlong Li

2017-01-01

Full Text Available This paper focuses on the evaluation standards of factor of safety for foundation stability analysis. The problem of foundation stability is analyzed via the methods of risk analysis of engineering structures and reliability-based design, and the factor of safety for foundation stability is determined by using bearing capacity safety-factor method (BSFM and strength safety-factor method (SSFM. Based on a typical example, the admissible factors of safety were calibrated with a target reliability index specified in relevant standards. Two safety criteria and their standards of bearing capacity of foundation for these two methods (BSFM and SSFM were established. The universality of the safety criteria and their standards for foundation reliability was verified based on the concept of the ratio of safety margin (RSM.

Evaluation of seismic hazards for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2002-01-01

The main objective of this Safety Guide is to provide recommendations on how to determine the ground motion hazards for a plant at a particular site and the potential for surface faulting, which could affect the feasibility of construction and safe operation of a plant at that site. The guidelines and procedures presented in this Safety Guide can appropriately be used in evaluations of site suitability and seismic hazards for nuclear power plants in any seismotectonic environment. The probabilistic seismic hazard analysis recommended in this Safety Guide also addresses the needs for seismic hazard analysis of external event PSAs conducted for nuclear power plants. Many of the methods and processes described may also be applicable to nuclear facilities other than power plants. Other phenomena of permanent ground displacement (liquefaction, slope instability, subsidence and collapse) as well as the topic of seismically induced flooding are treated in Safety Guides relating to foundation safety and coastal flooding. Recommendations of a general nature are given in Section 2. Section 3 discusses the acquisition of a database containing the information needed to evaluate and address all hazards associated with earthquakes. Section 4 covers the use of this database for construction of a seismotectonic model. Sections 5 and 6 review ground motion hazards and evaluations of the potential for surface faulting, respectively. Section 7 addresses quality assurance in the evaluation of seismic hazards for nuclear power plants

The role of engineering judgement, safety culture, and organizational factors in risk assessment

International Nuclear Information System (INIS)

Muzumdar, Ajit; Professor, Visiting

1996-01-01

This paper reviews the role of engineering judgement, safety culture, and organizational factors in risk assessment by examining the reasons for human-based error. The need for more emphasis on producing engineers with good engineering judgement is described. The progress in quantifying the role of safety culture and organizational factors in risk assessment studies is summarized

IRSN-ANCCLI partnership. Organizational and human factors in nuclear safety - April 2014

International Nuclear Information System (INIS)

Jeffroy, Francois; Garron, Joel; Mercel, Philippe; Compagnat, Gilles; Gaucher, Eric; Gaillard, Pierre; Fanchini, Henri; Jacquemont, Vincent

2013-06-01

The contributions (Power Point presentations) of this seminar first address the history of the taking into account of organizational and human factors until the Fukushima accident (history of their taking into account in nuclear safety expertise in France, history of the development of policy of organizational and human factors by an operator). The next contributions discuss the main issues regarding these factors after Fukushima: report by a work-group, work performed by the the Comite d'Orientation sur les Facteurs Sociaux, Organisationnels et Humains (Committee of orientation on social, organizational and human factors). The third session addresses the implication of stakeholders in expertise on these factors: analysis of organizational and human factors by a local information commission or by a CHSCT (committee of hygiene, safety and working conditions)

Investigation of occupational health and safety application using the internal and external factor assessment matrix: SWOT

Directory of Open Access Journals (Sweden)

2012-09-01

Material and Method: IIn this study, the threats, opportunities, weaknesses and strengths were evaluated by one of the tools named SWOT, in one of the assembly industries company in Iran, in order to controlling the operations in this company considering to safety and health standard (OHSAS18001. A comparison of the company’s performance in implementing the safety and health standards was done between years 1387 and 1388 contain in the Company considered, and weighted scoring weaknesses, strengths, threats and opportunities using the matrix of internal factors (strengths and weaknesses and external factors (treats and opportunities then, the importance of each factor were determined in the company’s implementation and enforcement of those standards. . Result: Focusing on the strengths and weaknesses, opportunities and threats, some strategies to improve the implementation were presented. Any points were weighted based on the most important weaknesses identified as the lack of monitoring contractors, lack of management commitment for implementation of OHSAS18001, no attempt to identify the risks of change, lack of training needs assessment, main strengths identified in the context of adequate budget health and safety, environmental efforts, identify risk for abnormal conditions, the most important threats for immediate delivery customer orders and the opportunity to support the safety and health plans, were determined. . Conclusion: Sum of the weighted scores in year 87 were obtained for the external factors (opportunities and threats, 2.16 and internal factors (strengths and weaknesses 1.66. Both of these scores were less than 2.5 (minimum amount of the acceptable rate so, the company has been poor performance in the implementation of this standard for the year 87 and a weak reaction in the use of opportunities and the minimize threats has. In case of internal factors, it was worse than external one and the situation was more bold of the weaknesses companies to

Patient and carer identified factors which contribute to safety incidents in primary care: a qualitative study.

Science.gov (United States)

Hernan, Andrea L; Giles, Sally J; Fuller, Jeffrey; Johnson, Julie K; Walker, Christine; Dunbar, James A

2015-09-01

Patients can have an important role in reducing harm in primary-care settings. Learning from patient experience and feedback could improve patient safety. Evidence that captures patients' views of the various contributory factors to creating safe primary care is largely absent. The aim of this study was to address this evidence gap. Four focus groups and eight semistructured interviews were conducted with 34 patients and carers from south-east Australia. Participants were asked to describe their experiences of primary care. Audio recordings were transcribed verbatim and specific factors that contribute to safety incidents were identified in the analysis using the Yorkshire Contributory Factors Framework (YCFF). Other factors emerging from the data were also ascertained and added to the analytical framework. Thirteen factors that contribute to safety incidents in primary care were ascertained. Five unique factors for the primary-care setting were discovered in conjunction with eight factors present in the YCFF from hospital settings. The five unique primary care contributing factors to safety incidents represented a range of levels within the primary-care system from local working conditions to the upstream organisational level and the external policy context. The 13 factors included communication, access, patient factors, external policy context, dignity and respect, primary-secondary interface, continuity of care, task performance, task characteristics, time in the consultation, safety culture, team factors and the physical environment. Patient and carer feedback of this type could help primary-care professionals better understand and identify potential safety concerns and make appropriate service improvements. The comprehensive range of factors identified provides the groundwork for developing tools that systematically capture the multiple contributory factors to patient safety. Published by the BMJ Publishing Group Limited. For permission to use (where not

Evaluating the impact of child safety seat check-up events on parental knowledge.

Science.gov (United States)

Herring, Ashley B; Jones, Ches; Nunez, Casandra

2002-12-01

Riding unrestrained is the greatest risk factor for death and injury among children in motor vehicles. Restraining a child can reduce the risk of death for that child by up to 71%. However, despite increased awareness, child safety seat usage rates are still disturbingly low. The purpose of this study was to evaluate the impact that child safety seat check-up events have on parental knowledge on child safety seats and installation. The subjects for this study were 101 parents/caregivers who attended child safety seat check-up events in northwest Arkansas from May 2000 through June 2001. A 20-item survey was conducted via the telephone. Results showed that the check-up events in northwest Arkansas have had an impact on self-efficacy. The participants of the events were primarily Caucasian and females in the 30-34 age group. Nine of 10 subjects scored in the high knowledge category. Conclusions are that check-up events do have an impact on parental knowledge and are accepted by the target group. Additionally, participants believed that car seats are of great importance and do protect their children in the event of a crash.

Probabilistic safety assessment technology for commercial nuclear power plant security evaluation

International Nuclear Information System (INIS)

Liming, J.K.; Johnson, D.H.; Dykes, A.A.

2004-01-01

Commercial nuclear power plant physical security has received much more intensive treatment and regulatory attention since September 11, 2001. In light of advancements made by the nuclear power industry in the field of probabilistic safety assessment (PSA) for its power plants over that last 30 years, and given the many examples of successful applications of risk-informed regulation at U. S. nuclear power plants during recent years, it may well be advisable to apply a 'risk-informed' approach to security management at nuclear power plants from now into the future. In fact, plant PSAs developed in response to NRC Generic Letter 88-20 and related requirements are used to help define target sets of critical plant safety equipment in our current security exercises for the industry. With reasonable refinements, plant PSAs can be used to identify, analyze, and evaluate reasonable and prudent approaches to address security issues and associated defensive strategies at nuclear power plants. PSA is the ultimate scenario-based approach to risk assessment, and thus provides a most powerful tool in identifying and evaluating potential risk management decisions. This paper provides a summary of observations of factors that are influencing or could influence cost-effective or 'cost-reasonable' security management decision-making in the current political environment, and provides recommendations for the application of PSA tools and techniques to the nuclear power plant operational safety response exercise process. The paper presents a proposed framework for nuclear power plant probabilistic terrorist risk assessment that applies these tools and techniques. (authors)

Impact of demographic factors on employees perceptions on health and safety management in the Greek Ministries

Directory of Open Access Journals (Sweden)

Pavlopoulou Georgia

2016-10-01

Full Text Available The purpose of the present study was to investigate the impact of selected demographic factors on perceptions of office workers regarding the management of health and safety in the office work place. For the data collection it was used a scale validated with a sample of 155 office employees. The final sample of the study was 301 subjects from three large Ministries in the Athens region of Greece, selected randomly. Exploratory factor analysis revealed four factors. A further comparison of the health and safety scale factors toward gender, marital status, working hours, monitoring or not seminars related to workplace safety and involvement or not in accidents in the office revealed that: (a The male employees had more positive perceptions than their female counterparts (t = 2.62, p <0.010. (b Positive perceptions showed and those who had attended seminars on safety and those who were not involved in office accidents (t = 2.16, p <0.032 and t = -2.19, p <0.033, respectively. (c It was also founded that men had more positive perceptions than women in the factor workplace environmental conditions (t = 2.40, p <0.018, while employees who had attended seminars on safety had a higher score on the factor health and safety issues in the office in comparison with their colleagues who did not, (t = 2.17, p <0.031. (d Employees who were involved in office accidents rated higher the questions of the factor health and safety issues in the office (t = -2.52, p <0. 015 and lower the factor workplace environmental conditions (t = -2.07, p = .043. It is concluded that despite the differences in the rating health and safety scale, in relation to selected variables, perceptions of employees regarding the management health and safety in the office work are positive.

Development of the safety evaluation system in the respects of organizational factors and workers' consciousness. Pt. 3. On know-how of its applying to an engineering company

International Nuclear Information System (INIS)

Sasou, Kunihide; Hasegawa, Naoko; Hirose, Ayako; Tsuge, Tadashi; Hayase, Kenichi; Takano, Kenichi

2003-01-01

'Safety Culture' has been paid attentions since Chernobyl accident in 1986. The criticality accident in 1999 and other kinds of scandals involving big name companies in Japan make them realize the importance of safety culture. CRIEPI is developing a safety evaluation system. The evaluation is based on the answers to the questionnaire and their statistical analysis such as t-test principal component analysis. This report discusses know-how when applying this evaluation technique to an engineering company whose jobs are ranging from production of products to engineering services to customers. About 15% engineers of the company answered the questionnaire and the answers were statistically analyzed. The results show the followings. First, the evaluation technique is not suitable to evaluations between departments with different kinds of jobs in each. That is because risk on the business of each department differs from each other due to the differences in the kinds of jobs. This indicates that the evaluation technique should be applied to groups whose jobs and risks on their business are equal. Second, the technique is applicable to branches with some kinds of jobs. A branch consists of small groups with different jobs but the ratios of the groups in a branch are nearly equal to those in other branches. Therefore, risks in each branch are equal. Finally, the technique should consider the frequency in which risks of a group to be tested realize. The larger the frequency in which workers face them is, the more the workers pay attention to safety issues. These findings indicate that the safety evaluation system needs several kinds of the standards of comparisons to be applied to evaluate safety levels in wide range of industrial companies. (author)

Seismic Hazards in Site Evaluation for Nuclear Installations. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-08-15

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear installations. It supplements the Safety Requirements publication on Site Evaluation for Nuclear Installations. The present publication provides guidance and recommends procedures for the evaluation of seismic hazards for nuclear power plants and other nuclear installations. It supersedes Evaluation of Seismic Hazards for Nuclear Power Plants, IAEA Safety Standards Series No. NS-G-3.3 (2002). In this publication, the following was taken into account: the need for seismic hazard curves and ground motion spectra for the probabilistic safety assessment of external events for new and existing nuclear installations; feedback of information from IAEA reviews of seismic safety studies for nuclear installations performed over the previous decade; collective knowledge gained from recent significant earthquakes; and new approaches in methods of analysis, particularly in the areas of probabilistic seismic hazard analysis and strong motion simulation. In the evaluation of a site for a nuclear installation, engineering solutions will generally be available to mitigate, by means of certain design features, the potential vibratory effects of earthquakes. However, such solutions cannot always be demonstrated to be adequate for mitigating the effects of phenomena of significant permanent ground displacement such as surface faulting, subsidence, ground collapse or fault creep. The objective of this Safety Guide is to provide recommendations and guidance on evaluating seismic hazards at a nuclear installation site and, in particular, on how to determine: (a) the vibratory ground motion hazards, in order to establish the design basis ground motions and other relevant parameters for both new and existing nuclear installations; and (b) the potential for fault displacement and the rate of fault displacement that could affect the feasibility of the site or the safe operation of the installation at

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)

10CFR50.59 safety evaluation training and expert system development

International Nuclear Information System (INIS)

Kline, S.W.; Dickinson, D.B.

1988-01-01

10CFR50.59 permits utilities to make changes to and conduct tests or experiments on operating nuclear power plants without prior US Nuclear Regulatory Commission (NCR) approval unless the proposed change, test, or experiment (i.e, the proposed activity) involves a change to the plant technical specifications or an unreviewed safety question (USQ). To provide guidance to their engineers for making the determination of whether a proposed activity involves a USQ. Bechtel has developed a safety evaluation training program. This training program incorporates the guidance in and NRC comments to the November 1987 draft Nuclear Management and Resources Council safety evaluation guidance document, NRC statements contained in inspection reports and other documents, and the experience of senior Bechtel engineers. To further develop the question and concerns that need to be addressed in a safety evaluation in a systematic manner, Bechtel is incorporating the training program guidance and other information into an IBM PC-AT-based working model of an expert system using the NEXPERT expert system development tool. The development and use of this expert system working model are being undertaken to provide consistency and completeness to the thought process used and the output provided by Bechtel engineers when performing a safety evaluation

Evaluation procedure of software safety plan for digital I and C of KNGR

International Nuclear Information System (INIS)

Lee, Jang Soo; Park, Jong Kyun; Lee, Ki Young; Kwon, Ki Choon; Kim, Jang Yeol; Cheon, Se Woo

2000-05-01

The development, use, and regulation of computer systems in nuclear reactor instrumentation and control (I and C) systems to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Korean next generation reactor (KNGR) software safety verification and validation (SSVV) task, Korea Atomic Energy Research Institute, which investigates different aspects of computer software in reactor I and C systems, and describes the engineering procedures for developing such a software. The purpose of this guideline is to give the software safety evaluator the trail map between the code and standards layer and the design methodology and documents layer for the software important to safety in nuclear power plants. Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organizations. The requirements for software important to safety of nuclear reactor are described in such positions and standards, for example, the new standard review plan (SRP), IEC 880 supplements, IEEE standard 1228-1994, IEEE standard 7-4.3.2-1993, and IAEA safety series No. 50-SG-D3 and D8. We presented the guidance for evaluating the safety plan of the software in the KNGR protection systems. The guideline consists of the regulatory requirements for software safety in chapter 2, the evaluation checklist of software safety plan in chapter3, and the evaluation results of KNGR software safety plan in chapter 4

Evaluation procedure of software safety plan for digital I and C of KNGR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Soo; Park, Jong Kyun; Lee, Ki Young; Kwon, Ki Choon; Kim, Jang Yeol; Cheon, Se Woo

2000-05-01

The development, use, and regulation of computer systems in nuclear reactor instrumentation and control (I and C) systems to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Korean next generation reactor (KNGR) software safety verification and validation (SSVV) task, Korea Atomic Energy Research Institute, which investigates different aspects of computer software in reactor I and C systems, and describes the engineering procedures for developing such a software. The purpose of this guideline is to give the software safety evaluator the trail map between the code and standards layer and the design methodology and documents layer for the software important to safety in nuclear power plants. Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organizations. The requirements for software important to safety of nuclear reactor are described in such positions and standards, for example, the new standard review plan (SRP), IEC 880 supplements, IEEE standard 1228-1994, IEEE standard 7-4.3.2-1993, and IAEA safety series No. 50-SG-D3 and D8. We presented the guidance for evaluating the safety plan of the software in the KNGR protection systems. The guideline consists of the regulatory requirements for software safety in chapter 2, the evaluation checklist of software safety plan in chapter3, and the evaluation results of KNGR software safety plan in chapter 4.

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

International Nuclear Information System (INIS)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-01

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments that lie outside the

Using partial safety factors in wind turbine design and testing

Energy Technology Data Exchange (ETDEWEB)

Musial, W.D. [National Renewable Energy Lab., Golden, CO (United States)

1997-12-31

This paper describes the relationship between wind turbine design and testing in terms of the certification process. An overview of the current status of international certification is given along with a description of limit-state design basics. Wind turbine rotor blades are used to illustrate the principles discussed. These concepts are related to both International Electrotechnical Commission and Germanischer Lloyd design standards, and are covered using schematic representations of statistical load and material strength distributions. Wherever possible, interpretations of the partial safety factors are given with descriptions of their intended meaning. Under some circumstances, the authors` interpretations may be subjective. Next, the test-load factors are described in concept and then related to the design factors. Using technical arguments, it is shown that some of the design factors for both load and materials must be used in the test loading, but some should not be used. In addition, some test factors not used in the design may be necessary for an accurate test of the design. The results show that if the design assumptions do not clearly state the effects and uncertainties that are covered by the design`s partial safety factors, outside parties such as test labs or certification agencies could impose their own meaning on these factors.

Case Study on Influence Factor Trend Analysis of the Accidents and Events of Nuclear Power Plants by applying Nuclear Safety Culture Framework

International Nuclear Information System (INIS)

Park, J. Y.; Park, Y. W.; Park, H.G.

2016-01-01

This study 1) established the standard based on frameworks of safety culture principles that show safety culture promotion goals, 2) analyzed the linkages with the frameworks that were established by analyzing each incident cause and weak point from selected 268 cases(rating over INES grade 1) among 4,088 cases (as of April 1, 2015). The 4,088 cases were selected as a result of database analysis from 702 accidents recorded in accident and rating evaluation reports that were published in the National Nuclear Safety Commission and overseas IRS (International Reporting System for operating Experience), and 3) finally conducted a trend analysis studies with these comprehensive results. From the investigations, followings were concluded. 1) In order to analyze the safety culture, analysis methodology is required. 2) Analytical methodology for building sustainable safety culture promoting a virtuous cycle system was developed 3) Among variety of process input data, 970 domestic and overseas incidents were selected as targets and 502 accidents were classified as safety culture related events by utilizing screen filter of IAEA GS-G-3.5 Appendix I and Framework (Nuclear Safety Culture Base Frame) developed by BEES, Inc. for safety culture analysis method. 4) As a result, complex safety culture influence factors for the one reason which was difficult to separate by conventional methods was able to be analyzed. 5) The cumulative data through the system was results of virtuous trend analysis rather than temporary results. Thus, it could be unique cultural factors of the domestic industry and could derive trend differences for domestic safety culture factors accordingly

Case Study on Influence Factor Trend Analysis of the Accidents and Events of Nuclear Power Plants by applying Nuclear Safety Culture Framework

Energy Technology Data Exchange (ETDEWEB)

Park, J. Y.; Park, Y. W.; Park, H.G. [KAIST, Daejeon (Korea, Republic of)

2016-05-15

This study 1) established the standard based on frameworks of safety culture principles that show safety culture promotion goals, 2) analyzed the linkages with the frameworks that were established by analyzing each incident cause and weak point from selected 268 cases(rating over INES grade 1) among 4,088 cases (as of April 1, 2015). The 4,088 cases were selected as a result of database analysis from 702 accidents recorded in accident and rating evaluation reports that were published in the National Nuclear Safety Commission and overseas IRS (International Reporting System for operating Experience), and 3) finally conducted a trend analysis studies with these comprehensive results. From the investigations, followings were concluded. 1) In order to analyze the safety culture, analysis methodology is required. 2) Analytical methodology for building sustainable safety culture promoting a virtuous cycle system was developed 3) Among variety of process input data, 970 domestic and overseas incidents were selected as targets and 502 accidents were classified as safety culture related events by utilizing screen filter of IAEA GS-G-3.5 Appendix I and Framework (Nuclear Safety Culture Base Frame) developed by BEES, Inc. for safety culture analysis method. 4) As a result, complex safety culture influence factors for the one reason which was difficult to separate by conventional methods was able to be analyzed. 5) The cumulative data through the system was results of virtuous trend analysis rather than temporary results. Thus, it could be unique cultural factors of the domestic industry and could derive trend differences for domestic safety culture factors accordingly.

Prescribing safety, negotiating expertise. Building of nuclear safety human factors expertise

International Nuclear Information System (INIS)

Rolina, Gregory

2008-01-01

This Ph.D thesis is dedicated to a specific type of expertise, the safety of nuclear installations in the field of human and organisational factors. Empirical work is at the foundation of this thesis: the monitoring of experts 'in action', allowed a detailed reconstruction of three cases they were examining. The analysis, at the core of which lies the definition of what an efficient expertise can be, emphasizes the incompleteness of the knowledge that links together the nuclear facilities' organisational characteristics and their safety. This leads us to identify the expert's three ranges of actions (rhetorical, cognitive, operative). Defined from objectives and constraints likely to influence the expert's behaviour, those three ranges each require specific skills. A conception of expertise based on these ranges seems adaptable to other sectors and allows an enrichment of models of expertise cited in literature. Historical elements from French institutions of nuclear safety are also called upon to take into consideration some of the determinants of the expertise; its efficiency relies on the upholding of a continuous dialogue between the regulators (the experts and the control authority) and the regulated (the operators). This type of historically inherited regulation makes up a specificity of the French system of external control of nuclear risks. (author) [fr

Implication of human factors in terms of safety

International Nuclear Information System (INIS)

Furuta, Kazuo

2001-01-01

A critical accident of JCO occurred on September 30, 1999 gave a large impact not only to common society but also to nuclear energy field. This accident occurred by direct reason perfectly out of forecasting of the participants of nuclear energy, where a company made up a guideline violating from business allowance and safety rule and workmen also operated under a procedure out of the guideline. After the accident, a number of countermeasures on equipments, rules, and regulations were carried out, but discussion on software such as their operating methods, concrete regulation on business and authority of operators, and training of specialists seems to be much late. Safety is a problem on a complex system, containing not only hardware but also software such as human, organization, society, and so on. Then, here was discussed on a problem directly faced by conventional safety, engineering centering at hardware through thinking of a problem on human factors. (G.K.)

Report of the evaluation by the Ad Hoc Review Committee on Nuclear Safety Research. Result evaluation in fiscal year 2000

International Nuclear Information System (INIS)

2001-06-01

The Research Evaluation Committee, which consisted of 14 members from outside of the Japan Atomic Energy Research Institute (JAERI), set up an Ad Hoc Review Committee on Nuclear Safety Research in accordance with the Fundamental Guideline for the Evaluation of Research and Development (R and D) at JAERI' and its subsidiary regulations in order to evaluate the R and D accomplishments achieved for five years from Fiscal Year 1995 to Fiscal Year 1999 at Department of Reactor Safety Research, Department of Fuel Cycle Safety Research, Department of Environmental Safety Research and Department of Safety Research Technical Support in Tokai Research Establishment at JAERI. The Ad Hoc Review Committee consisted of 11 specialists from outside of JAERI. The Ad Hoc Review Committee conducted its activities from December 2000 to February 2001. The evaluation was performed on the basis of the materials submitted in advance and of the oral presentations made at the Ad Hoc Review Committee meeting which was held on December 11, 2000, in line with the items, viewpoints, and criteria for the evaluation specified by the Research Evaluation Committee. The result of the evaluation by the Ad Hoc Review Committee was submitted to the Research Evaluation Committee, and was judged to be appropriate at its meeting held on March 16, 2001. This report describes the result of the evaluation by the Ad Hoc Review Committee on Nuclear Safety Research. (author)

Organizational safety factors research lessons learned

International Nuclear Information System (INIS)

Ryan, T.G.

1995-01-01

This Paper reports lessons learned and state of knowledge gained from an organizational factors research activity involving commercial nuclear power plants in the United States, through the end of 1991, as seen by the scientists immediately involved in the research. Lessons learned information was gathered from the research teams and individuals using a question and answer format. The following five questions were submitted to each team and individual: (1) What organizational factors appear to influence safety performance in some systematic way, (2) Should organizational factors research focus at the plant level, or should it extend beyond the plant level to the parent company, rate setting commissions, regulatory agencies, (3) How important is having direct access to plants for doing organizational factors research, (4) What lessons have been learned to date as the result of doing organizational factors research in a nuclear regulatory setting, and (5) What organizational research topics and issues should be pursued in the future? Conclusions based on the responses provided for this report are that organizational factors research can be conducted in a regulatory setting and produce useful results. Technologies pioneered in other academic, commercial, and military settings can be adopted for use in a nuclear regulatory setting. The future success of such research depends upon the cooperation of regulators, contractors, and the nuclear industry

Second Meeting for Evaluation of the Nuclear Safety Convention

International Nuclear Information System (INIS)

2002-01-01

This report presents the results of the Second Meeting for Evaluation of the Nuclear Safety Convention. the CSN. as the only competent Government organism on nuclear safety, represented Spain in the preparation of the national report and at the Review Meeting, acquiring a set of obligations for the next three years, until the holding of third meeting. (Author)

Quality factors in the life cycle of software oriented to safety systems in nuclear power plants

International Nuclear Information System (INIS)

Nunez McLeod, J.E.; Rivera, S.S.

1997-01-01

The inclusion of software in safety related systems for nuclear power plants, makes it necessary to include the software quality assurance concept. The software quality can be defined as the adjustment degree between the software and the specified requirements and user expectations. To guarantee a certain software quality level it is necessary to make a systematic and planned set of tasks, that constitute a software quality guaranty plan. The application of such a plan involves activities that should be performed all along the software life cycle, and that can be evaluated through the so called quality factors, due to the fact that the quality itself cannot be directly measured, but indirectly as some of it manifestations. In this work, a software life cycle model is proposed, for nuclear power plant safety related systems. A set os software quality factors is also proposed , with its corresponding classification according to the proposed model. (author) [es

Safety update on the use of recombinant activated factor VII in approved indications.

Science.gov (United States)

Neufeld, Ellis J; Négrier, Claude; Arkhammar, Per; Benchikh el Fegoun, Soraya; Simonsen, Mette Duelund; Rosholm, Anders; Seremetis, Stephanie

2015-06-01

This updated safety review summarises the large body of safety data available on the use of recombinant activated factor VII (rFVIIa) in approved indications: haemophilia with inhibitors, congenital factor VII (FVII) deficiency, acquired haemophilia and Glanzmann's thrombasthenia. Accumulated data up to 31 December 2013 from clinical trials as well as post-marketing data (registries, literature reports and spontaneous reports) were included. Overall, rFVIIa has shown a consistently favourable safety profile, with no unexpected safety concerns, in all approved indications. No confirmed cases of neutralising antibodies against rFVIIa have been reported in patients with congenital haemophilia, acquired haemophilia or Glanzmann's thrombasthenia. The favourable safety profile of rFVIIa can be attributed to the recombinant nature of rFVIIa and its localised mechanism of action at the site of vascular injury. Recombinant FVIIa activates factor X directly on the surface of activated platelets, which are present only at the site of injury, meaning that systemic activation of coagulation is avoided and the risk of thrombotic events (TEs) thus reduced. Nonetheless, close monitoring for signs and symptoms of TE is warranted in all patients treated with any pro-haemostatic agent, including rFVIIa, especially the elderly and any other patients with concomitant conditions and/or predisposing risk factors to thrombosis. Copyright © 2015 Elsevier Ltd. All rights reserved.

Safety assessment of human and organizational factors in French fuel cycle facilities

International Nuclear Information System (INIS)

Menuet, Lise; Beauquier, Sophie

2013-01-01

According to the French law, each nuclear facility has to provide a safety demonstration every ten years. The assessment of this demonstration supports the decision of the French Safety Authority regarding the authorisation of operating for the ten years to come. In addition, transversal topics, which are linked with safety performance, such as safety management, management of competencies, maintenance's policy are periodically evaluated. One aspect of these assessments relates to Human and Organizational Factors (HOF) and their contribution to safety. Our communication will describe the assessment of the HOF-related part, performed by the Institute for Radioprotection and Nuclear Safety Institute (IRSN) the Technical Support Organisation of the French Safety Authority). It will focus on the methodological framework, the tools which are developed and used for assessing the integration of HOF in safety demonstration, and the main difficulties of this kind of assessment. Each situation will be illustrated by concrete examples coming from safety assessments concerning fuel cycle's plants: Areva's plants dedicated to uranium conversion, uranium enrichment, fuel manufacturing, spent fuel reprocessing, treatment facilities and CEA's laboratories dedicated to research and development and to interim spent fuel storage. The methodological framework for assessing HOF currently implements three main steps which will be precisely described: - checking that the nuclear plant has made an exhaustive analysis of the risks linked with HOF. Regarding to HOF, the Licensee safety demonstration is based on the description of the main human activities which are considered as hazardous regarding safety. These activities are accomplished with a human contribution and they require a safe realisation. - assessing the human, organisational and technical barriers that the nuclear plant have planed in order to make the operations safe, to avoid, prevent or detect an

Evaluating Safety Culture Under the Socio-Technical Complex Systems Perspective

International Nuclear Information System (INIS)

Lemos, F. L. de

2016-01-01

Since the term “safety culture” was coined, it has gained more and more attention as an effort to achieve higher levels of system safety. A good deal of effort has been done in order to better define, evaluate and implement safety culture programs in organizations throughout all industries, and especially in the Nuclear Industry. Unfortunately, despite all those efforts, we continue to witness accidents that are, in great part, attributed to flaws in the safety culture of the organization. Fukushima nuclear accident is one example of a serious accident in which flaws in the safety culture has been pointed to as one of the main contributors. In general, the definitions of safety culture emphasise the social aspect of the system. While the definitions also include the relations with the technical aspects, it does so in a general sense. For example, the International Nuclear Safety Advisory Group (INSAG) defines safety culture as: “The assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receives the attention warranted by their significance.” By the way safety culture is defined we can infer that it represents a property of a social system, or a property of the social aspect of the system. In this sense, the social system is a component of the whole system. Where, “system” is understood to be comprised of a social (humans) and technical (equipment) aspects, as a Nuclear Power Plant, for example. Therefore, treating safety culture as an identity on its own right, finding and fixing flaws in the safety culture may not be enough to improve safety of the system. We also needed to evaluate all the interactions between the components that comprise all the aspects of the system. In some cases a flaw in the safety culture can easily be detected, such as an employee not wearing appropriate individual protection equipment, e.g., dosimeter, or when basic safety

Preliminary safety evaluation, based on initial site investigation data. Planning document

International Nuclear Information System (INIS)

Hedin, Allan

2002-12-01

This report is a planning document for the preliminary safety evaluations (PSE) to be carried out at the end of the initial stage of SKBs ongoing site investigations for a deep repository for spent nuclear fuel. The main purposes of the evaluations are to determine whether earlier judgements of the suitability of the candidate area for a deep repository with respect to long-term safety holds up in the light of borehole data and to provide feed-back to continued site investigations and site specific repository design. The preliminary safety evaluations will be carried out by a safety assessment group, based on a site model, being part of a site description, provided by a site modelling group and a repository layout within that model suggested by a repository engineering group. The site model contains the geometric features of the site as well as properties of the host rock. Several alternative interpretations of the site data will likely be suggested. Also the biosphere is included in the site model. A first task for the PSE will be to compare the rock properties described in the site model to previously established criteria for a suitable host rock. This report gives an example of such a comparison. In order to provide more detailed feedback, a number of thermal, hydrological, mechanical and chemical analyses of the site will also be included in the evaluation. The selection of analyses is derived from the set of geosphere and biosphere analyses preliminarily planned for the comprehensive safety assessment named SR-SITE, which will be based on a complete site investigation. The selection is dictated primarily by the expected feedback to continued site investigations and by the availability of data after the PSE. The repository engineering group will consider several safety related factors in suggesting a repository layout: Thermal calculations will be made to determine a minimum distance between canisters avoiding canister surface temperatures above 100 deg C

Safety design and evaluation policy for future FBRs in Japan

International Nuclear Information System (INIS)

Aizawa, Kiyoto

1991-01-01

The safety policy for fast breeder reactors (FBRs) has gradually matured in accordance with the development of FBRs. The safety assessment of the Japanese prototype FBR, Monju during the licensing process accelerated the maturity and the integration of knowledge and databases. Results are expected to be reflected in the establishment of the safety design and evaluation policy for FBRs. Although the methodologies and safety policies developed for LWRs are applicable in principle to future FBRs, it is neither rational nor realistic to treat safety only with these policies. It is recommended that one should develop the methodologies and safety policies starting from understanding of the inherent safety characteristics of FBR's through safety research, plant operating experience and design work. In the last few years, some technical committees were organized in Japan and have discussed key safety issues which are specific to FBRs in order to provide preparatory reports and to establish safety standards and guidelines for future commercial FBRs. (author)

Evaluating the effectiveness of Behavior-Based Safety education methods for commercial vehicle drivers.

Science.gov (United States)

Wang, Xuesong; Xing, Yilun; Luo, Lian; Yu, Rongjie

2018-08-01

Risky driving behavior is one of the main causes of commercial vehicle related crashes. In order to achieve safer vehicle operation, safety education for drivers is often provided. However, the education programs vary in quality and may not always be successful in reducing crash rates. Behavior-Based Safety (BBS) education is a popular approach found effective by numerous studies, but even this approach varies as to the combination of frequency, mode and content used by different education providers. This study therefore evaluates and compares the effectiveness of BBS education methods. Thirty-five drivers in Shanghai, China, were coached with one of three different BBS education methods for 13 weeks following a 13-week baseline phase with no education. A random-effects negative binomial (NB) model was built and calibrated to investigate the relationship between BBS education and the driver at-fault safety-related event rate. Based on the results of the random-effects NB model, event modification factors (EMF) were calculated to evaluate and compare the effectiveness of the methods. Results show that (1) BBS education was confirmed to be effective in safety-related event reduction; (2) the most effective method among the three applied monthly face-to-face coaching, including feedback with video and statistical data, and training on strategies to avoid driver-specific unsafe behaviors; (3) weekly telephone coaching using statistics and strategies was rated by drivers as the most convenient delivery mode, and was also significantly effective. Copyright © 2018 Elsevier Ltd. All rights reserved.

Review on the Evaluation System of Public Safety Carrying Capacity about Small Town Community

Institute of Scientific and Technical Information of China (English)

Ming; SUN; Tianyu; ZHU

2014-01-01

Recently,small town community public safety problem has been increasingly highlighted,but its research is short on public safety carrying capacity. Through the investigation and study of community public safety carrying capacity,this paper analyzes the problem of community public safety in our country,to construct index evaluation system of public safety carrying capacity in small town community. DEA method is used to evaluate public safety carrying capacity in small town community,to provide scientific basis for the design of support and standardization theory about small town community in public safety planning.

Application of factor analysis in psychological diagnostics (sample: study of students’ social safety

Directory of Open Access Journals (Sweden)

Pavel Aleksandrovich Kislyakov

2015-10-01

Our recommendations for the use of factor analysis, with necessary restrictions and clear reasons of a possible ambiguity of solutions, will be useful to everyone interested in mastering an adequate mathematical tool for solving problems pertaining to the humanities, in particular, those of practical psychology. As a practical example is presented the research of the psychological factors which provide students’ social safety. With the help of the factor analysis relevant personal and professional qualities of a teacher were revealed which are the subjective factors of students’ social safety, namely: social anticipation, socio-psychological stress resistance, social tolerance, professional orientation, responsibility, communication skills.

DEVELOPMENT OF HUMAN FACTORS ENGINEERING GUIDANCE FOR SAFETY EVALUATIONS OF ADVANCED REACTORS

International Nuclear Information System (INIS)

O'HARA, J.; PERSENSKY, J.; SZABO, A.

2006-01-01

Advanced reactors are expected to be based on a concept of operations that is different from what is currently used in today's reactors. Therefore, regulatory staff may need new tools, developed from the best available technical bases, to support licensing evaluations. The areas in which new review guidance may be needed and the efforts underway to address the needs will be discussed. Our preliminary results focus on some of the technical issues to be addressed in three areas for which new guidance may be developed: automation and control, operations under degraded conditions, and new human factors engineering methods and tools

Safety of Abiraterone Acetate in Castration-resistant Prostate Cancer Patients With Concomitant Cardiovascular Risk Factors.

Science.gov (United States)

Procopio, Giuseppe; Grassi, Paolo; Testa, Isabella; Verzoni, Elena; Torri, Valter; Salvioni, Roberto; Valdagni, Riccardo; de Braud, Filippo

2015-10-01

The aim of this study was to evaluate the safety profile of abiraterone acetate (AA) in metastatic castration-resistant prostate cancer (mCRPC) men with cardiovascular comorbidity, as little conclusive safety data are available in this patient subset. A retrospective analysis of mCRPC patients with controlled cardiovascular comorbidities, receiving AA 1000 mg administered orally once daily and prednisone 5 mg twice daily, between April 2011 and July 2012, was performed. All clinical and instrumental variables and toxicity data were analyzed by descriptive statistics: mean, standard deviation, minimum and maximum values for continuous variables, and absolute and relative frequencies for categorical variables. A total of 51 mCRPC patients were evaluated. Metastatic sites included the bone (74%), lungs, and liver (26%). All patients were previously treated with at least 2 lines of hormone and 1 docetaxel-based chemotherapy. Preexisting cardiac risk factors included hypertension (41%), cardiac ischemia (12%), arrhythmias (6%), dislipidemia (18%), and hyperglycemia (30%). No grade 3-4 adverse events were observed. Grade 1-2 adverse events included fluid retention (18%), asthenia (15%), and hypertension (16%). Median progression-free survival was 5.1 months (95% confidence interval, 0.5-12). Prostate specific antigen assessment revealed a good overall disease control rate (64%). AA appears to be safe and well tolerated even in patients with cardiovascular comorbidities or with increased risk factors for cardiovascular diseases.

Expert evaluation in NPP safety important systems licensing process

International Nuclear Information System (INIS)

Mikhail, A Yastrebenetsky; Vasilchenko, V.N.

2001-01-01

Expert evaluation of nuclear power plant safety important systems modernization is an integral part of these systems licensing process. The paper contains some aspects of this evaluation which are based on Ukrainian experience of VVER-1000 and VVER-440 modernization. (authors)

Expert evaluation in NPP safety important systems licensing process

Energy Technology Data Exchange (ETDEWEB)

Mikhail, A Yastrebenetsky; Vasilchenko, V.N. [Ukrainian State Scientific Technical Center of Nuclear and Radiation Safety (Ukraine)

2001-07-01

Expert evaluation of nuclear power plant safety important systems modernization is an integral part of these systems licensing process. The paper contains some aspects of this evaluation which are based on Ukrainian experience of VVER-1000 and VVER-440 modernization. (authors)

The balance between safety and productivity and its relationship with human factors and safety awareness and communication in aircraft manufacturing

NARCIS (Netherlands)

Karanikas, N.; Melis, Damien Jose; Kourousis, Kyriakos

2017-01-01

Background: This paper presents the findings of a pilot research survey which assessed the degree of balance between safety and productivity, and its relationship with awareness and communication of human factors and safety rules in the aircraft manufacturing environment. Methods: The study was

Partial safety factors for berthing velocity and loads on marine structures

NARCIS (Netherlands)

Roubos, A.A.; Peters, D.J.; Groenewegen, Leon; Steenbergen, R.

2018-01-01

Design methods for marine structures have evolved into load and resistance factor design, however existing partial safety factors related to berthing velocity and loads have not been verified and validated by measurement campaigns. In this study, field observations of modern seagoing vessels

Meteorological events in site evaluation for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

This Safety Guide provides recommendations and guidance on conducting hazard assessments of extreme and rare meteorological phenomena. It is of interest to safety assessors and regulators involved in the licensing process as well as to designers of nuclear power plants. This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It supplements the IAEA Safety Requirements publication on Site Evaluation for Nuclear Facilities which is to supersede the Code on the Safety of Nuclear Power Plants: Siting, Safety Series No. 50-C-S (Rev. 1), IAEA, Vienna (1988). The present Safety Guide supersedes two earlier Safety Guides: Safety Series No. 50-SG-S11A (1981) on Extreme Meteorological Events in Nuclear Power Plant Siting, Excluding Tropical Cyclones and Safety Series No. 50-SG-S11B (1984) on Design Basis Tropical Cyclone for Nuclear Power Plants. The purpose of this Safety Guide is to provide recommendations and guidance on conducting hazard assessments of extreme and rare meteorological phenomena. This Safety Guide provides interpretation of the Safety Requirements publication on Site Evaluation for Nuclear Facilities and guidance on how to fulfil these requirements. It is aimed at safety assessors or regulators involved in the licensing process as well as designers of nuclear power plants, and provides them with guidance on the methods and procedures for analyses that support the assessment of the hazards associated with extreme and rare meteorological events. This Safety Guide discusses the extreme values of meteorological variables and rare meteorological phenomena, as well as their rates of occurrence, according to the following definitions: (a) Extreme values of meteorological variables such as air temperature and wind speed characterize the meteorological or climatological environment. And (b) Rare meteorological phenomena

The 'PROCESO' index: a new methodology for the evaluation of operational safety in the chemical industry

International Nuclear Information System (INIS)

Marono, M.; Pena, J.A.; Santamaria, J.

2006-01-01

The acknowledgement of industrial installations as complex systems in the early 1980s outstands as a milestone in the path to operational safety. Process plants are social-technical complex systems of a dynamic nature, whose properties depend not only on their components, but also on the inter-relations among them. A comprehensive assessment of operational safety requires a systemic approach, i.e. an integrated framework that includes all the relevant factors influencing safety. Risk analysis methodologies and safety management systems head the list of methods that point in this direction, but they normally require important plant resources. As a consequence, their use is frequently restricted to especially dangerous processes often driven by compliance with legal requirements. In this work a new safety index for the chemical industry, termed the 'Proceso' Index (standing for the Spanish terms for PROCedure for the Evaluation of Operational Safety), has been developed. PROCESO is based on the principles of systems theory, has a tree-like structure and considers 25 areas to guide the review of plant safety. The method uses indicators whose respective weight values have been obtained via an expert judgement technique. This paper describes the steps followed to develop this new Operational Safety Index, explains its structure and illustrates its application to process plants

Report of the evaluation by the Ad Hoc Review Committee on Nuclear Safety Research. Result evaluation in fiscal year 2000

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-06-01

The Research Evaluation Committee, which consisted of 14 members from outside of the Japan Atomic Energy Research Institute (JAERI), set up an Ad Hoc Review Committee on Nuclear Safety Research in accordance with the Fundamental Guideline for the Evaluation of Research and Development (R and D) at JAERI' and its subsidiary regulations in order to evaluate the R and D accomplishments achieved for five years from Fiscal Year 1995 to Fiscal Year 1999 at Department of Reactor Safety Research, Department of Fuel Cycle Safety Research, Department of Environmental Safety Research and Department of Safety Research Technical Support in Tokai Research Establishment at JAERI. The Ad Hoc Review Committee consisted of 11 specialists from outside of JAERI. The Ad Hoc Review Committee conducted its activities from December 2000 to February 2001. The evaluation was performed on the basis of the materials submitted in advance and of the oral presentations made at the Ad Hoc Review Committee meeting which was held on December 11, 2000, in line with the items, viewpoints, and criteria for the evaluation specified by the Research Evaluation Committee. The result of the evaluation by the Ad Hoc Review Committee was submitted to the Research Evaluation Committee, and was judged to be appropriate at its meeting held on March 16, 2001. This report describes the result of the evaluation by the Ad Hoc Review Committee on Nuclear Safety Research. (author)

Expert opinions on the acceptance of alternative methods in food safety evaluations

NARCIS (Netherlands)

Punt, Ans; Bouwmeester, Hans; Schiffelers, Marie Jeanne W.A.; Peijnenburg, Ad A.C.M.

2018-01-01

Inclusion of alternative methods that replace, reduce, or refine (3R) animal testing within regulatory safety evaluations of chemicals generally faces many hurdles. The goal of the current work is to i) collect responses from key stakeholders involved in food safety evaluations on what they consider

A new cyber security risk evaluation method for oil and gas SCADA based on factor state space

International Nuclear Information System (INIS)

Yang, Li; Cao, Xiedong; Li, Jie

2016-01-01

Based on comprehensive analysis of the structure and the potential safety problem of oil and gas SCADA(Supervisor control and data acquisition) network, aiming at the shortcomings of traditional evaluation methods, combining factor state space and fuzzy comprehensive evaluation method, a new network security risk evaluation method of oil and gas SCADA is proposed. First of all, formal description of factor state space and its complete mathematical definition were presented; secondly, factor fuzzy evaluation steps were discussed; then, using analytic hierarchy method, evaluation index system for oil and gas SCADA system was established, the index weights of all factors were determined by two-two comparisons; structure design of three layers in reasoning machine was completed. Experiments and tests show that the proposed method is accurate, reliable and practical. Research results provide the template and the new method for the other industries.

Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

Energy Technology Data Exchange (ETDEWEB)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-15

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments

Evaluation of the nuclear installations safety of the CEA in 1998

International Nuclear Information System (INIS)

Laverie, M.

1999-09-01

Michel Laverie, Director of the nuclear safety and quality at the Cea, took stoke of the CEA nuclear installations in 1998. After a recall of the nuclear safety policy and organization, the author presents the risks factors bound to the CEA activities as the dismantling, the wastes and the human factors. A last part is devoted to the list of the accidents occurred during 1998 in the nuclear installations. Tables and statistics illustrate this analysis. (A.L.B.)

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

OpenAIRE

B. V. Zubkov; H. E. Fourar

2017-01-01

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

Development of an evaluation framework for African-European hospital patient safety partnerships.

Science.gov (United States)

Rutter, Paul; Syed, Shamsuzzoha B; Storr, Julie; Hightower, Joyce D; Bagheri-Nejad, Sepideh; Kelley, Edward; Pittet, Didier

2014-04-01

Patient safety is recognised as a significant healthcare problem worldwide, and healthcare-associated infections are an important aspect. African Partnerships for Patient Safety is a WHO programme that pairs hospitals in Africa with hospitals in Europe with the objective to work together to improve patient safety. To describe the development of an evaluation framework for hospital-to-hospital partnerships participating in the programme. The framework was structured around the programme's three core objectives: facilitate strong interhospital partnerships, improve in-hospital patient safety and spread best practices nationally. Africa-based clinicians, their European partners and experts in patient safety were closely involved in developing the evaluation framework in an iterative process. The process defined six domains of partnership strength, each with measurable subdomains. We developed a questionnaire to measure these subdomains. Participants selected six indicators of hospital patient safety improvement from a short-list of 22 based on their relevance, sensitivity to intervention and measurement feasibility. Participants proposed 20 measures of spread, which were refined into a two-part conceptual framework, and a data capture tool created. Taking a highly participatory approach that closely involved its end users, we developed an evaluation framework and tools to measure partnership strength, patient safety improvements and the spread of best practice.