Examining the influence of urban definition when assessing relative safety of drinking-water in Nigeria

Energy Technology Data Exchange (ETDEWEB)

Christenson, Elizabeth; Bain, Robert [The Water Institute at UNC, University of North Carolina at Chapel Hill, NC (United States); Wright, Jim [Geography and Environment, University of Southampton, Southampton (United Kingdom); Aondoakaa, Stephen [Geography and Environmental Management, University of Abuja, Abuja (Nigeria); School of Geography, University of Nottingham, Nottingham (United Kingdom); Hossain, Rifat [World Health Organization, Geneva (Switzerland); Bartram, Jamie, E-mail: jbartram@unc.edu [The Water Institute at UNC, University of North Carolina at Chapel Hill, NC (United States)

2014-08-15

Reducing inequalities is a priority from a human rights perspective and in water and public health initiatives. There are periodic calls for differential national and global standards for rural and urban areas, often justified by the suggestion that, for a given water source type, safety is worse in urban areas. For instance, initially proposed post-2015 water targets included classifying urban but not rural protected dug wells as unimproved. The objectives of this study were to: (i) examine the influence of urban extent definition on water safety in Nigeria, (ii) compare the frequency of thermotolerant coliform (TTC) contamination and prevalence of sanitary risks between rural and urban water sources of a given type and (iii) investigate differences in exposure to contaminated drinking-water in rural and urban areas. We use spatially referenced data from a Nigerian national randomized sample survey of five improved water source types to assess the extent of any disparities in urban–rural safety. We combined the survey data on TTC and sanitary risk with map layers depicting urban versus rural areas according to eight urban definitions. When examining water safety separately for each improved source type, we found no significant urban–rural differences in TTC contamination and sanitary risk for groundwater sources (boreholes and protected dug wells) and inconclusive findings for piped water and stored water. However, when improved and unimproved source types were combined, TTC contamination was 1.6 to 2.3 times more likely in rural compared to urban water sources depending on the urban definition. Our results suggest that different targets for urban and rural water safety are not justified and that rural dwellers are more exposed to unsafe water than urban dwellers. Additionally, urban–rural analyses should assess multiple definitions or indicators of urban to assess robustness of findings and to characterize a gradient that disaggregates the urban�

Examining the influence of urban definition when assessing relative safety of drinking-water in Nigeria

International Nuclear Information System (INIS)

Christenson, Elizabeth; Bain, Robert; Wright, Jim; Aondoakaa, Stephen; Hossain, Rifat; Bartram, Jamie

2014-01-01

Reducing inequalities is a priority from a human rights perspective and in water and public health initiatives. There are periodic calls for differential national and global standards for rural and urban areas, often justified by the suggestion that, for a given water source type, safety is worse in urban areas. For instance, initially proposed post-2015 water targets included classifying urban but not rural protected dug wells as unimproved. The objectives of this study were to: (i) examine the influence of urban extent definition on water safety in Nigeria, (ii) compare the frequency of thermotolerant coliform (TTC) contamination and prevalence of sanitary risks between rural and urban water sources of a given type and (iii) investigate differences in exposure to contaminated drinking-water in rural and urban areas. We use spatially referenced data from a Nigerian national randomized sample survey of five improved water source types to assess the extent of any disparities in urban–rural safety. We combined the survey data on TTC and sanitary risk with map layers depicting urban versus rural areas according to eight urban definitions. When examining water safety separately for each improved source type, we found no significant urban–rural differences in TTC contamination and sanitary risk for groundwater sources (boreholes and protected dug wells) and inconclusive findings for piped water and stored water. However, when improved and unimproved source types were combined, TTC contamination was 1.6 to 2.3 times more likely in rural compared to urban water sources depending on the urban definition. Our results suggest that different targets for urban and rural water safety are not justified and that rural dwellers are more exposed to unsafe water than urban dwellers. Additionally, urban–rural analyses should assess multiple definitions or indicators of urban to assess robustness of findings and to characterize a gradient that disaggregates the urban�

Examining the Relationship Between Safety Management System Implementation and Safety Culture in Collegiate Flight Schools

OpenAIRE

Robertson, Michael F

2018-01-01

Safety management systems (SMS) are becoming the industry standard for safety management throughout the aviation industry. As the Federal Aviation Administration continues to mandate SMS for different segments, the assessment of an organization’s safety culture becomes more important. An SMS can facilitate the development of a strong aviation safety culture. This study describes how safety culture and SMS are integrated. The purpose of this study was to examine the relationship between an ...

Integrated plant safety assessment systematic evaluation program. R.E. Ginna Nuclear Power Plant, Rochester Gas and Electric Corporation, Docket No. 50-244

International Nuclear Information System (INIS)

1982-05-01

The Systematic Evaluation Program was initiated in February 1978 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to reconfirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the R.E. Ginna Nuclear Power Plant (located in Wayne County near Rochester, NY), one of ten plants reviewed under Phase II of this program, and indicates how 137 topics selected for review under Phase I of the program were addressed. Equipment and procedural changes have been identified as a result of the review. It is expected that this report will be one of the bases in considering the issuance of a full-term operating license in place of the existing provisional operating license

General re-evaluation of the safety on the nuclear ship 'Mutsu' and its repair work

International Nuclear Information System (INIS)

1980-01-01

According to the proposition by the Committee for Investigation Radiation Leak on Mutsu, the works of the general re-evaluation of safety were started after the approval by the Committee for Investigating General Re-evaluation and Repair Techniques for Mutsu. The contents of the general re-evaluation of safety are the inspection of the machines and equipments in the nuclear reactor plant, the review of the design of the nuclear reactor plant, the analysis of the nuclear reactor plant behavior in accidents, and the related experimental researches. These works have been carried out for five years, and problem did not arise at all regarding the nuclear reactor so far, but from the viewpoint of improving the safety and reliability further, it was decided to carry out the repair work based on the general re-evaluation of safety. The contents of the repair work are the improvement of the emergency core-cooling system, the improvement of the safety protection system, the improvement of the radiation monitoring equipments, the improvement of the containment vessel boundary, the improvement of the actuators for technological safety facilities, the improvement of the method controlling secondary water quality, and other repair works. The progress of the general re-evaluation of safety is reported. (Kako, I.)

The ISRN has assessed the safety and radioprotection of the solid radioactive waste management area at CEA/Saclay; L'IRSN a expertise la surete et la radioprotection de la zone de gestion des dechets radioactifs solides du CEA/Saclay

Energy Technology Data Exchange (ETDEWEB)

NONE

2009-07-01

Completed by a report of a Permanent Group of Experts on the safety re-examination for the solid radioactive waste management area of the CEA/Saclay, this paper is a statement of the assessment by the IRSN (the French Institute of Radioprotection and Nuclear Safety) of this safety re-examination. It first recalls what a safety re-examination is for a nuclear installation and indicates the documents which have been transmitted by the CEA and analysed. Then, it briefly describes the activities which are performed in the concerned area: collecting, warehousing, conditioning, control and shipping of solid radioactive wastes or useless fuels. The IRSN comments the works done and to be done for the different operations on different types of waste (low or middle activity), and then the different aspects of the installation safety (radiations, confining, fire, climatic events, civil engineering)

Training courses on integrated safety assessment modelling for waste repositories

International Nuclear Information System (INIS)

Mallants, D.

2007-01-01

Near-surface or deep repositories of radioactive waste are being developed and evaluated all over the world. Also, existing repositories for low- and intermediate-level waste often need to be re-evaluated to extend their license or to obtain permission for final closure. The evaluation encompasses both a technical feasibility as well as a safety analysis. The long term safety is usually demonstrated by means of performance or safety assessment. For this purpose computer models are used that calculate the migration of radionuclides from the conditioned radioactive waste, through engineered barriers to the environment (groundwater, surface water, and biosphere). Integrated safety assessment modelling addresses all relevant radionuclide pathways from source to receptor (man), using in combination various computer codes in which the most relevant physical, chemical, mechanical, or even microbiological processes are mathematically described. SCK-CEN organizes training courses in Integrated safety assessment modelling that are intended for individuals who have either a controlling or supervising role within the national radwaste agencies or regulating authorities, or for technical experts that carry out the actual post-closure safety assessment for an existing or new repository. Courses are organised by the Department of Waste and Disposal

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

International Nuclear Information System (INIS)

2013-01-01

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

Re-assessment of dose from the Vinten extremity dosemeter

International Nuclear Information System (INIS)

O'Hagan, J.B.; Pearson, A.J.; Dutt, J.C.

1989-01-01

A procedure is described for re-assessing the dose from the Vinten extremity dosemeter using phototransferred thermoluminescence. The technique produces a linear response up to 50 Sv. The re-assessment efficiency is approximately 10% for the first re-assessment and 3-5% for the second re-assessment. The detection threshold values (at the 95% confidence level) are 3 mSv and 9 mSv respectively. (author)

Towards understanding work-as-done in air traffic management safety assessment and design

International Nuclear Information System (INIS)

Woltjer, Rogier; Pinska-Chauvin, Ella; Laursen, Tom; Josefsson, Billy

2015-01-01

This paper describes the approach taken and the results to develop guidance, to include Resilience Engineering principles in methodology for safety assessment of functional changes, in Air Traffic Management (ATM). It summarizes the process of deriving resilience principles for ATM, originating from Resilience Engineering concepts and transposed into ATM operations. These principles are the foundation for guidance material incorporating Resilience Engineering (RE) concepts into safety assessment methodology. The guidance material provides a method using workshops generating qualitative descriptions of RE principles applied to ATM services of everyday work, as done currently and as envisioned after introduction of a new technology or way of working. The guidance material has been proposed as part of the safety assessment methodology of SESAR (Single European Sky ATM Research), and as stand-alone guidance for ATM design processes. The methodology was validated via a test case on the i4D/CTA (Controlled Time of Arrival) concept. Operational examples from the application of the developed guidance to the i4D/CTA concept are provided. Initial evaluation of the guidance suggests that the methodology (1) provides a narrative, vocabulary and documentation means of project discussions on resilience; (2) brings the discussions of safety and resilience closer to operational practice; (3) facilitates a broader systemic and integrative perspective on operational, management, business, safety, environmental, and human performance aspects; and (4) can extend the vocabulary of safety assessment to include the description of emergent properties, to better support functional changes in ATM. - Highlights: • Guidance material for safety assessment based on systemic thinking is proposed. • It operationalizes Resilience Engineering principles in Air Traffic Management, including a case study. • It enables description of expected changes in work-as-done when introducing a new

A new method for the assessment of patient safety competencies during a medical school clerkship using an objective structured clinical examination

Directory of Open Access Journals (Sweden)

Renata Mahfuz Daud-Gallotti

2011-01-01

Full Text Available INTRODUCTION: Patient safety is seldom assessed using objective evaluations during undergraduate medical education. OBJECTIVE: To evaluate the performance of fifth-year medical students using an objective structured clinical examination focused on patient safety after implementation of an interactive program based on adverse events recognition and disclosure. METHODS: In 2007, a patient safety program was implemented in the internal medicine clerkship of our hospital. The program focused on human error theory, epidemiology of incidents, adverse events, and disclosure. Upon completion of the program, students completed an objective structured clinical examination with five stations and standardized patients. One station focused on patient safety issues, including medical error recognition/disclosure, the patient-physician relationship and humanism issues. A standardized checklist was completed by each standardized patient to assess the performance of each student. The student's global performance at each station and performance in the domains of medical error, the patient-physician relationship and humanism were determined. The correlations between the student performances in these three domains were calculated. RESULTS: A total of 95 students participated in the objective structured clinical examination. The mean global score at the patient safety station was 87.59 ± 1.24 points. Students' performance in the medical error domain was significantly lower than their performance on patient-physician relationship and humanistic issues. Less than 60% of students (n = 54 offered the simulated patient an apology after a medical error occurred. A significant correlation was found between scores obtained in the medical error domains and scores related to both the patient-physician relationship and humanistic domains. CONCLUSIONS: An objective structured clinical examination is a useful tool to evaluate patient safety competencies during the medical

Risk Assessment During the Final Phase of an Uncontrolled Re-Entry

Science.gov (United States)

Gaudel, A.; Hourtolle, C.; Goester, J. F.; Fuentes, N.

2013-09-01

As French National Space Agency, CNES is empowered to monitor compliance with technical regulations of the French Space Operation Act, FSOA, and to take all necessary measures to ensure the safety of people, property, public health and environment for all space operations involving French responsibility at international level.Therefore, CNES developed ELECTRA that calculates the risk for ground population involved in three types of events: rocket launching, controlled re-entry and uncontrolled re-entry. For the first two cases, ELECTRA takes into account degraded cases due to a premature stop of propulsion.Major evolutions were implemented recently on ELECTRA to meet new users' requirements, like the risk assessment during the final phase of uncontrolled re-entry, that can be combined with the computed risk for each country involved by impacts.The purpose of this paper is to provide an overview of the ELECTRA method and main functionalities, and then to highlight these recent improvements.

Additional safety assessment of ITER - Addition safety investigation of the INB ITER

International Nuclear Information System (INIS)

2012-01-01

This assessment aims at re-assessing safety margins in the light of events which occurred in Fukushima Daiichi, i.e. extreme natural events challenging the safety of installations. After a presentation of some characteristics of the ITER installation (location, activities, buildings, premise detritiation systems, electric supply, handling means, radioactive materials, chemical products, nuclear risks, specific risks), the report addresses the installation robustness by identifying cliff-edge effect risks which can be related to a loss of confinement of radioactive materials, explosions, a significant increase of exposure level, a possible effect on water sheets, and so on. The next part addresses the various aspects related to a seismic risk: installation sizing (assessment methodology, seismic risk characterization in Cadarache), sizing protection measures, installation compliance, and margin assessment. External flooding is the next addressed risk: installation sizing with respect to this specific risk, protection measures, installation compliance, margin assessment, and studied additional measures. Other extreme natural phenomena are considered (meteorological conditions, earthquake and flood) which may have effects on other installations (dam, canal). Then, the report addresses technical risks like the loss of electric supplies and cooling systems, the way a crisis is managed in terms of technical and human means and organization in different typical accidental cases. Subcontracting practices are also discussed. A synthesis proposes an overview of this additional safety assessment and discusses the impact which could have additional measures which could be implemented

ELECTRA © Launch and Re-Entry Safety Analysis Tool

Science.gov (United States)

Lazare, B.; Arnal, M. H.; Aussilhou, C.; Blazquez, A.; Chemama, F.

2010-09-01

French Space Operation Act gives as prime objective to National Technical Regulations to protect people, properties, public health and environment. In this frame, an independent technical assessment of French space operation is delegated to CNES. To perform this task and also for his owns operations CNES needs efficient state-of-the-art tools for evaluating risks. The development of the ELECTRA© tool, undertaken in 2007, meets the requirement for precise quantification of the risks involved in launching and re-entry of spacecraft. The ELECTRA© project draws on the proven expertise of CNES technical centers in the field of flight analysis and safety, spaceflight dynamics and the design of spacecraft. The ELECTRA© tool was specifically designed to evaluate the risks involved in the re-entry and return to Earth of all or part of a spacecraft. It will also be used for locating and visualizing nominal or accidental re-entry zones while comparing them with suitable geographic data such as population density, urban areas, and shipping lines, among others. The method chosen for ELECTRA© consists of two main steps: calculating the possible reentry trajectories for each fragment after the spacecraft breaks up; calculating the risks while taking into account the energy of the fragments, the population density and protection afforded by buildings. For launch operations and active re-entry, the risk calculation will be weighted by the probability of instantaneous failure of the spacecraft and integrated for the whole trajectory. ELECTRA©’s development is today at the end of the validation phase, last step before delivery to users. Validation process has been performed in different ways: numerical application way for the risk formulation; benchmarking process for casualty area, level of energy of the fragments entries and level of protection housing module; best practices in space transportation industries concerning dependability evaluation; benchmarking process for

Examination of rotator cuff re-tear

International Nuclear Information System (INIS)

Kitahara, Hiroyuki; Yabe, Yoshihiro; Norimatsu, Takahiro; Adachi, Shinji; Sera, Keisuke

2010-01-01

The six-month post-operative re-tear rate in 72 arthroscopic rotator cuff repair cases was 16.3% by MRI. The re-tear rate of massive tears was 50%. We investigated the details of the re-tears by MRI and arthroscopic findings. High re-tear rates were connected with cuff tear size and fatty degeneration of muscle belly. Cases with poor cuff quality in arthroscopically showed high re-tear rate. These results suggest that surgery operation should be performed as soon as possible after diagnosis of cuff tear to obtain good results. Cases with damage of long head of the biceps (LHB) are likely to develop impingement causes of re-tears. Some type of rehabilitation is required to avoid impingement in such cases. (author)

Safety assessment of primary system components at the USNRC

Energy Technology Data Exchange (ETDEWEB)

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

1988-12-31

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

Selected component failure rate values from fusion safety assessment tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, L.C.

1998-09-01

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

Selected Component Failure Rate Values from Fusion Safety Assessment Tasks

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee Charles

1998-09-01

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

Selected component failure rate values from fusion safety assessment tasks

International Nuclear Information System (INIS)

Cadwallader, L.C.

1998-01-01

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

Safety/security interface assessments at commercial nuclear power plants

International Nuclear Information System (INIS)

Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

1985-01-01

The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur

Safety/security interface assessments at commercial nuclear power plants

International Nuclear Information System (INIS)

Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

1985-07-01

The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur. 2 refs

Effect of the interval of training course on understanding of radiation safety and an improvement of re-training course

International Nuclear Information System (INIS)

Miyoshi, Hirokazu; Yamamoto, Yasuyo; Adachi, Akio

2005-01-01

Radiation safety training courses are indispensable educational programs for radiation workers. We have two kinds of courses, which are held before use of radioisotope (beginner's training course) and held annually (re-training course). The interval between two courses was found to give some effects for radiation worker's recognition and knowledge on radiation safety through the result of examination and questionnaire on the radiation safety after training. The average scores of participants indicated that the short interval (3 months) was better than the long interval (almost one year). Furthermore, the average scores of participants in the 2003 training course were higher than those in the 2002 and 2001 training courses. Several participants were found to lack in the basic radiation safety attitude and knowledge. In order to improve these results, the practical training should be given additionally for workers, who lacked in understanding. (author)

Additional safety assessments. Report by the Nuclear Safety Authority - December 2011

International Nuclear Information System (INIS)

2011-12-01

The first part of this voluminous report proposes an assessment of targeted audits performed in French nuclear installations (water pressurized reactors on the one hand, laboratories, factories and waste and dismantling installations on the other hand) on issues related to the Fukushima accident. The examined issues were the protection against flooding and against earthquake, and the loss of electricity supplies and of cooling sources. The second part addresses the additional safety assessments of the reactors and the European resistance tests: presentation of the French electronuclear stock, earthquake, flooding and natural hazards (installation sizing, safety margin assessment), loss of electricity supplies and cooling systems, management of severe accidents, subcontracting conditions. The third part addresses the same issues for nuclear installations other than nuclear power reactors

A patient safety objective structured clinical examination.

Science.gov (United States)

Singh, Ranjit; Singh, Ashok; Fish, Reva; McLean, Don; Anderson, Diana R; Singh, Gurdev

2009-06-01

There are international calls for improving education for health care workers around certain core competencies, of which patient safety and quality are integral and transcendent parts. Although relevant teaching programs have been developed, little is known about how best to assess their effectiveness. The objective of this work was to develop and implement an objective structured clinical examination (OSCE) to evaluate the impact of a patient safety curriculum. The curriculum was implemented in a family medicine residency program with 47 trainees. Two years after commencing the curriculum, a patient safety OSCE was developed and administered at this program and, for comparison purposes, to incoming residents at the same program and to residents at a neighboring residency program. All 47 residents exposed to the training, all 16 incoming residents, and 10 of 12 residents at the neighboring program participated in the OSCE. In a standardized patient case, error detection and error disclosure skills were better among trained residents. In a chart-based case, trained residents showed better performance in identifying deficiencies in care and described more appropriate means of addressing them. Third year residents exposed to a "Systems Approach" course performed better at system analysis and identifying system-based solutions after the course than before. Results suggest increased systems thinking and inculcation of a culture of safety among residents exposed to a patient safety curriculum. The main weaknesses of the study are its small size and suboptimal design. Much further investigation is needed into the effectiveness of patient safety curricula.

Safety Assessment for Decommissioning

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-06-15

In the past few decades, international guidance has been developed on methods for assessing the safety of predisposal and disposal facilities for radioactive waste. More recently, it has been recognized that there is also a need for specific guidance on safety assessment in the context of decommissioning nuclear facilities. The importance of safety during decommissioning was highlighted at the International Conference on Safe Decommissioning for Nuclear Activities held in Berlin in 2002 and at the First Review Meeting of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management in 2003. At its June 2004 meeting, the Board of Governors of the IAEA approved the International Action Plan on Decommissioning of Nuclear Facilities (GOV/2004/40), which called on the IAEA to: ''establish a forum for the sharing and exchange of national information and experience on the application of safety assessment in the context of decommissioning and provide a means to convey this information to other interested parties, also drawing on the work of other international organizations in this area''. In response, in November 2004, the IAEA launched the international project Evaluation and Demonstration of Safety for Decommissioning of Facilities Using Radioactive Material (DeSa) with the following objectives: -To develop a harmonized approach to safety assessment and to define the elements of safety assessment for decommissioning, including the application of a graded approach; -To investigate the practical applicability of the methodology and performance of safety assessments for the decommissioning of various types of facility through a selected number of test cases; -To investigate approaches for the review of safety assessments for decommissioning activities and the development of a regulatory approach for reviewing safety assessments for decommissioning activities and as a basis for regulatory decision making; -To provide a forum

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

Science.gov (United States)

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

2010-07-01

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

Re-Assessment of Concrete Bridges

DEFF Research Database (Denmark)

Thoft-Christensen, Palle

In this paper two aspects of re-assessment of the reliability of concrete bridges are discussed namely modelling of the corrosion of reinforcement and updating of uncertain variables. The main reason for deterioration of concrete bridges is corrosion of the reinforcement. Therefore, modelling...

A RE-ASSESSMENT OF OLDER DRIVERS AS A ROAD SAFETY RISK

Directory of Open Access Journals (Sweden)

Jim LANGFORD

2006-01-01

Full Text Available Older drivers are frequently viewed as overly represented in crashes, particularly when crash involvement per distance travelled is considered. This perception has led to a call for tighter licensing conditions for older drivers, a policy which inevitably results in mobility restrictions for at least some drivers. However there is a growing body of research evidence which shows that as a group, older drivers represent no greater road risk than drivers from other age groups once different levels of driving activity are taken into account. This paper has examined aspects of older drivers' fitness to drive based on survey data and off-road and on-road driving performance from a sample of 905 New Zealand older drivers. The results show that policies which target all older drivers and lead to licensing and mobility restrictions cannot be justified from a safety basis.

Effects of Re-Using a Conceptual Examination Question in Physics

Science.gov (United States)

Sharma, Manjula D.; Sefton, Ian M.; Cole, Martyn; Whymark, Aaron; Millar, Rosemary M.; Smith, Andrew

2005-12-01

We report on a study of what happened when we recycled a conceptual examination question in a first-year university physics course. The question, which was used for three consecutive years, asked about an astronaut's experience of weighing in an orbiting space-craft. Our original intention was to use a phenomenographic approach to look for differences in students' descriptive answers. Having done that, we decided to add a study of the marks that were awarded to those answers. The first time that the question was re-used, the distribution of answers amongst our phenomenographic categories showed a decrease in the common conception that gravity is zero in the satellite and an increase in explanations in terms of free fall. When the question was re-used a second time, that difference was maintained but it was not significantly increased. The distribution of marks for the question was different over the three years in a way that appears to be unrelated to differences in students' conceptual understandings. Differences in the distribution of marks are more likely to be related to differences in marking procedures. We conclude that studies like this one have the potential to contribute to improvements in university assessment procedures. In particular we propose that phenomenographic analysis could be used in the design of marking schemes.

Study of the cost-benefit analysis method for safety. Meeting of the Permanent Group in charge of nuclear reactors on the 5 July 2007

International Nuclear Information System (INIS)

2007-07-01

After a recall of the history of the issue of third decennial visit of the 900 MW reactors, of the IRSN preliminary analysis, of elements given to the Permanent Group, of requests made by the ASN, and a presentation of the analysis performed by the IRSN, this large report presents the cost-benefit analysis method and its potential applications (principle, cost assessment, safety assessment, examples) and reports international experience gained in this area: the risk-informed approach (within the IAEA, in the USA, France and other European countries, the specific cost-benefit approach), existing cost-benefit type methods (comparison between methods used in the USA, in France and in Canada), and monetary assessment of accidents. It reports the application of the cost-benefit method for safety and its limitations, and then its application to modifications which have been implemented after safety re-examinations. It discusses the use of level 1 and 2 safety probabilistic studies, and reports the use of a cost-benefit method for safety within the frame of safety re-examinations

MAPLE-X10 reactor safety assessment

International Nuclear Information System (INIS)

Cotnam, K.D.; Lounsbury, R.I.; Gillespie, G.E.

1990-01-01

This paper reports on the safety assessment of the 10 MW MAPLE-X10 reactor which has involved a substantial component of PSA analysis to supplement deterministic analysis. Initiating events are identified through the use of a master logic diagram. The events are then examined through event sequence diagrams, at the concept design stage, followed by a set of reliability analyses that are coordinated with the event sequence diagrams. Improvements identified through the reliability analyses are incorporated into the design to ensure that safety objectives are attained

HSE's safety assessment principles for criticality safety

International Nuclear Information System (INIS)

Simister, D N; Finnerty, M D; Warburton, S J; Thomas, E A; Macphail, M R

2008-01-01

The Health and Safety Executive (HSE) published its revised Safety Assessment Principles for Nuclear Facilities (SAPs) in December 2006. The SAPs are primarily intended for use by HSE's inspectors when judging the adequacy of safety cases for nuclear facilities. The revised SAPs relate to all aspects of safety in nuclear facilities including the technical discipline of criticality safety. The purpose of this paper is to set out for the benefit of a wider audience some of the thinking behind the final published words and to provide an insight into the development of UK regulatory guidance. The paper notes that it is HSE's intention that the Safety Assessment Principles should be viewed as a reflection of good practice in the context of interpreting primary legislation such as the requirements under site licence conditions for arrangements for producing an adequate safety case and for producing a suitable and sufficient risk assessment under the Ionising Radiations Regulations 1999 (SI1999/3232 www.opsi.gov.uk/si/si1999/uksi_19993232_en.pdf). (memorandum)

A re-examination of throats

KAUST Repository

Kim, J.-W.; Kim, D.; Lindquist, W. B.

2013-01-01

We critically re-examine the concept of a throat in a porous medium as a geometric quantity defined independently of an entry meniscus in a drainage process. To maintain the standard notion of a throat as a locally minimum-area cross section in the pore network, we demonstrate with examples that throats must intersect each other. Using flow simulation, we show that these intersecting throats correspond to capillary pressure controlled entry points during drainage. We have designed a throat-finding algorithm that explicitly locates intersecting throats, using a planar approximation for robustness and speed. The capability of the new algorithm was compared against an existing algorithm in the construction of pore-throat networks from X-ray computed tomography images of consolidated sandstones (7.5–22% porosity) and of an unconsolidated sand pack (32.5% porosity). We show that the probability of throat intersection increases significantly with porosity above 20%; in the sand pack, over 1/4 of all throats intersect with another.

A re-examination of throats

KAUST Repository

Kim, J.-W.

2013-11-01

We critically re-examine the concept of a throat in a porous medium as a geometric quantity defined independently of an entry meniscus in a drainage process. To maintain the standard notion of a throat as a locally minimum-area cross section in the pore network, we demonstrate with examples that throats must intersect each other. Using flow simulation, we show that these intersecting throats correspond to capillary pressure controlled entry points during drainage. We have designed a throat-finding algorithm that explicitly locates intersecting throats, using a planar approximation for robustness and speed. The capability of the new algorithm was compared against an existing algorithm in the construction of pore-throat networks from X-ray computed tomography images of consolidated sandstones (7.5–22% porosity) and of an unconsolidated sand pack (32.5% porosity). We show that the probability of throat intersection increases significantly with porosity above 20%; in the sand pack, over 1/4 of all throats intersect with another.

Safety assessment and verification for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2001-01-01

This publication supports the Safety Requirements on the Safety of Nuclear Power Plants: Design. This Safety Guide was prepared on the basis of a systematic review of all the relevant publications including the Safety Fundamentals, Safety of Nuclear Power Plants: Design, current and ongoing revisions of other Safety Guides, INSAG reports and other publications that have addressed the safety of nuclear power plants. This Safety Guide also provides guidance for Contracting Parties to the Convention on Nuclear Safety in meeting their obligations under Article 14 on Assessment and Verification of Safety. The Safety Requirements publication entitled Safety of Nuclear Power Plants: Design states that a comprehensive safety assessment and an independent verification of the safety assessment shall be carried out before the design is submitted to the regulatory body. This publication provides guidance on how this requirement should be met. This Safety Guide provides recommendations to designers for carrying out a safety assessment during the initial design process and design modifications, as well as to the operating organization in carrying out independent verification of the safety assessment of new nuclear power plants with a new or already existing design. The recommendations for performing a safety assessment are suitable also as guidance for the safety review of an existing plant. The objective of reviewing existing plants against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The methods and the recommendations of this Safety Guide can also be used by regulatory bodies for the conduct of the regulatory review and assessment. Although most recommendations of this Safety Guide are general and applicable to all types of nuclear reactors, some specific recommendations and examples apply mostly to water cooled reactors. Terms such as 'safety assessment', 'safety analysis' and 'independent

Managing acute pain in patients who report lactose intolerance: the safety of an old excipient re-examined.

Science.gov (United States)

Mill, Deanna; Dawson, Jessica; Johnson, Jacinta Lee

2018-05-01

Lactose intolerance is exceedingly common, reportedly affecting up to 70% of the world's population, leading to both abdominal and systemic symptoms. Current treatment focuses predominantly on restricting dietary consumption of lactose. Given lactose is one of the most commonly used excipients in the pharmaceutical industry, consideration must be given to the lactose content and therefore safety of pharmaceutical preparations prescribed for patients with lactose intolerance. This article summarizes the current literature examining the likelihood of inducing adverse effects through the administration of lactose-containing pharmaceutical preparations in patients reporting lactose intolerance, describes how to assess this risk on an individual patient basis and reviews suitable analgesic options for this population. A case study is presented detailing a patient reporting lactose intolerance who insists on treatment with the lactose-free product codeine/ibuprofen (Nurofen Plus) rather than other codeine-free analgesics. It is important to assess the likelihood of lactose as an excipient inducing symptoms in this scenario, as reluctance to cease codeine could suggest codeine dependence, an issue that is becoming increasingly common in countries such as Australia and Canada. Given codeine dependence is associated with serious sequelae including hospitalization and death, the patient must either be reassured the lactose component in their prescribed analgesics will not induce symptoms or an alternative treatment strategy must be confirmed. General recommendations applying theory from the literature to the management of acute pain in lactose-intolerant patients are discussed and specific treatment options are outlined. Although large inter-individual variability is reported, most lactose-intolerant patients can tolerate the small quantities of lactose found in pharmaceutical preparations. Cumulative lactose exposure can be assessed in patients taking multiple medications

Long-term maintenance of reducing conditions in a spent nuclear fuel repository. A re-examination of critical factors

Energy Technology Data Exchange (ETDEWEB)

Gascoyne, M. [Gascoyne GeoProjects Inc, Pinawa, MB (Canada)

1999-04-01

Penetration of oxidising groundwaters to depths of 500 m in a permeable bedrock, over a glacial/interglacial cycle, may compromise the stability of a spent nuclear fuel repository and cause the release and migration towards the surface of actinides and associated fission products. This report examines the potential for the penetration of oxygen (O{sub 2}) to depths of 500 m in a fractured crystalline rock environment, typical of the Fennoscandian Shield. Previous studies performed for the Swedish program of nuclear waste disposal (principally the SITE-94 safety assessment) have indicated that O{sub 2} might reach repository depths during a deglaciation when melt-water from the base of an ice sheet could enter the bedrock, driven by strong hydraulic gradients. This report re-examines aspects of this scenario and finds that: 1. The capacity of flow-path minerals to scavenge O{sub 2} from recharging groundwater may be lower than expected due to a previously unrecognised depletion of Fe(II)-bearing minerals in the active flow-paths in a fractured crystalline rock. 2. Assumptions in the SITE-94 assessment, such as the use of a continental-scale flow model, the lack of structural controls on groundwater flow, a preferred horizontal permeability, and the use of permeabilities to depths of 10 km that are up to two orders of magnitude greater than comparable environments, are disproportionately simplistic and represent an extremely conservative case. 3. Assumptions of a thin, discontinuous permafrost, a warm-based ice sheet, and high-O{sub 2} content melt-water at the repository site are unrealistic and overly conservative. A more realistic scenario, which includes a greater influence of permafrost, a cold based ice sheet, lower bedrock permeabilities and a more-limited, regional-scale flow path, is recommended as being more appropriate for use in the safety assessment. Under this revised scenario, it is believed that O{sub 2} will not penetrate to repository depths over

Long-term maintenance of reducing conditions in a spent nuclear fuel repository. A re-examination of critical factors

International Nuclear Information System (INIS)

Gascoyne, M.

1999-04-01

Penetration of oxidising groundwaters to depths of 500 m in a permeable bedrock, over a glacial/interglacial cycle, may compromise the stability of a spent nuclear fuel repository and cause the release and migration towards the surface of actinides and associated fission products. This report examines the potential for the penetration of oxygen (O 2 ) to depths of 500 m in a fractured crystalline rock environment, typical of the Fennoscandian Shield. Previous studies performed for the Swedish program of nuclear waste disposal (principally the SITE-94 safety assessment) have indicated that O 2 might reach repository depths during a deglaciation when melt-water from the base of an ice sheet could enter the bedrock, driven by strong hydraulic gradients. This report re-examines aspects of this scenario and finds that: 1. The capacity of flow-path minerals to scavenge O 2 from recharging groundwater may be lower than expected due to a previously unrecognised depletion of Fe(II)-bearing minerals in the active flow-paths in a fractured crystalline rock. 2. Assumptions in the SITE-94 assessment, such as the use of a continental-scale flow model, the lack of structural controls on groundwater flow, a preferred horizontal permeability, and the use of permeabilities to depths of 10 km that are up to two orders of magnitude greater than comparable environments, are disproportionately simplistic and represent an extremely conservative case. 3. Assumptions of a thin, discontinuous permafrost, a warm-based ice sheet, and high-O 2 content melt-water at the repository site are unrealistic and overly conservative. A more realistic scenario, which includes a greater influence of permafrost, a cold based ice sheet, lower bedrock permeabilities and a more-limited, regional-scale flow path, is recommended as being more appropriate for use in the safety assessment. Under this revised scenario, it is believed that O 2 will not penetrate to repository depths over the next 100,000 years

The Safety Assessment Framework Tool (SAFRAN) - Description, Overview and Applicability

International Nuclear Information System (INIS)

Alujevic, Luka

2014-01-01

The SAFRAN tool (Safety Assessment Framework) is a user-friendly software application that incorporates the methodologies developed in the SADRWMS (Safety Assessment Driven Radioactive Waste Management Solutions) project. The International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of all types of radioactive waste, including disused sources, small volumes, legacy and decommissioning waste, operational waste, and large volume naturally occurring radioactive material residues. SAFRAN provides aid in: Describing the predisposal RW management activities in a systematic way, Conducting the SA (safety assessment) with clear documentation of the methodology, assumptions, input data and models, Establishing a traceable and transparent record of the safety basis for decisions on the proposed RW management solutions, Demonstrating clear consideration of and compliance with national and international safety standards and recommendations. The SAFRAN tool allows the user to visibly, systematically and logically address predisposal radioactive waste management and decommissioning challenges in a structured way. It also records the decisions taken in such a way that it constitutes a justifiable safety assessment of the proposed management solutions. The objective of this paper is to describe the SAFRAN architecture and features, properly define the terms safety case and safety assessment, and to predict the future development of the SAFRAN tool and assess its applicability to the construction of a future LILW (Low and Intermediate Level Waste) storage facility and repository in Croatia, taking into account all the capabilities and modelling features of the SAFRAN tool. (author)

Application of fuzzy set theory for safety culture and safety management assessment of Kartini research reactor

International Nuclear Information System (INIS)

Syarip; Hauptmanns, U.

2000-01-01

The safety culture status of nuclear power plant is usually assessed through interview and/or discussions with personnel and management in plant, and an assessment of the pertinent documentation. The approach for safety culture assessment described in IAEA Safety Series, make uses of a questionnaire composed of questions which require 'Yes' or 'No' as an answer. Hence, it is basically a check-list approach which is quite common for safety assessments in industry. Such a procedure ignores the fact that the expert answering the question usually has knowledge which goes far beyond a mere binary answer. Additionally, many situations cannot readily be described in such restricted terms. Therefore, it was developed a checklist consisting of questions which are formulated such that they require more than a simple 'yes' or 'no' as an answer. This allows one to exploit the expert knowledge of the analyst appropriately by asking him to qualify the degree of compliance of each of the topics examined. The method presented has proved useful in assessing the safety culture and quality of safety management of the research reactor. The safety culture status and the quality of safety management of Kartini research reactor is rated as 'average'. The method is also flexible and allows one to add questions to existing areas or to introduce new areas covering related topics

Safety assessment methodologies and their application in development of near surface waste disposal facilities - the ASAM project

International Nuclear Information System (INIS)

Metcalf, P.

2003-01-01

The scope of ASAM project covers near surface disposal facilities for all types of low and intermediate level wastes with emphasis of the post-closure safety assessment.The objectives are to explore practical application to a range of disposal facilities for a number of purposes e.g. development of design concepts, safety re-assessment, upgrading safety and to develop practical approaches to assist regulators, operators and other experts in review of safety assessment. The task of the Co-ordination Group are: reassessment of existing facilities - use of safety assessment in decision making on selection of options (volunteer site Hungary); disused sealed sources - evaluation of disposability of disused sealed sources in near surface facilities (volunteer site Saratov, Russia); mining and minerals processing waste - evaluation of long-term safety (volunteer site pmc S. Africa). An agreement on the scope and objectives of the project are reached and the further consideration, such as human intrusion/institutional control/security; waste from oil/gas industry; very low level waste; categorization of sealed sources coordinated with other IAEA activities are outlined

Use of the Home Safety Self-Assessment Tool (HSSAT) within Community Health Education to Improve Home Safety.

Science.gov (United States)

Horowitz, Beverly P; Almonte, Tiffany; Vasil, Andrea

2016-10-01

This exploratory research examined the benefits of a health education program utilizing the Home Safety Self-Assessment Tool (HSSAT) to increase perceived knowledge of home safety, recognition of unsafe activities, ability to safely perform activities, and develop home safety plans of 47 older adults. Focus groups in two senior centers explored social workers' perspectives on use of the HSSAT in community practice. Results for the health education program found significant differences between reported knowledge of home safety (p = .02), ability to recognize unsafe activities (p = .01), safely perform activities (p = .04), and develop a safety plan (p = .002). Social workers identified home safety as a major concern and the HSSAT a promising assessment tool. Research has implications for reducing environmental fall risks.

Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

International Nuclear Information System (INIS)

2002-01-01

A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

Re-assessment of woodfuel supply and demand relationships in Kampong Thom Province, Cambodia

International Nuclear Information System (INIS)

Top, Neth; Ty, Sokhun; Mizoue, Nobuya; Ito, Satoshi; Kai, Shigetaka; Nakao, Toshio

2006-01-01

This study re-examined a previously published relationship between supply and demand for woodfuel at different spatial scales within Kampong Thom Province, Cambodia. We considered three different scales: The first was the whole area of the province. The second scale calculated village-scale data in zones of 1, 3, and 5km from each village and then aggregated for all villages of the province. The third scale also calculated data for the three zones at the village-scale, but the data were then aggregated according to three regional groupings based on population density and forest availability. When evaluating woodfuel supply, we excluded biomass increment of trees with diameter larger than 30cm because it was found that local people rarely collect trees larger than 30cm for fuel. On the demand side, dead wood, which was included in the previous assessment, and woodfuel obtained from non-forest sources were excluded to enable comparison of supply and demand that related only to living trees originating from forests. The re-assessment revealed large decreases in both supply and demand at each scale as compared with our previous assessment; supply reduced by 46% and demand by 36% at the whole province scale. However, the ratios of supply to demand at each scale examined were very similar for both assessments. This re-assessment therefore supports our previous findings: there is a deficiency in woodfuel resources in areas of high population density along the main road due to high woodfuel demand and a predominance of agricultural land and regrowth forest in close proximity to villages. This study underscores the usefulness of taking detailed woodfuel consumption patterns into account when assessing the impact of woodfuel demand on forests. For a more accurate assessment on the sustainability of woodfuel resources and utilization, further research is needed to project future woodfuel demand, not only for green wood from forested sources, but also for dead wood and

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

International Nuclear Information System (INIS)

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

1990-01-01

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

COHeRE: Cross-Ontology Hierarchical Relation Examination for Ontology Quality Assurance.

Science.gov (United States)

Cui, Licong

Biomedical ontologies play a vital role in healthcare information management, data integration, and decision support. Ontology quality assurance (OQA) is an indispensable part of the ontology engineering cycle. Most existing OQA methods are based on the knowledge provided within the targeted ontology. This paper proposes a novel cross-ontology analysis method, Cross-Ontology Hierarchical Relation Examination (COHeRE), to detect inconsistencies and possible errors in hierarchical relations across multiple ontologies. COHeRE leverages the Unified Medical Language System (UMLS) knowledge source and the MapReduce cloud computing technique for systematic, large-scale ontology quality assurance work. COHeRE consists of three main steps with the UMLS concepts and relations as the input. First, the relations claimed in source vocabularies are filtered and aggregated for each pair of concepts. Second, inconsistent relations are detected if a concept pair is related by different types of relations in different source vocabularies. Finally, the uncovered inconsistent relations are voted according to their number of occurrences across different source vocabularies. The voting result together with the inconsistent relations serve as the output of COHeRE for possible ontological change. The highest votes provide initial suggestion on how such inconsistencies might be fixed. In UMLS, 138,987 concept pairs were found to have inconsistent relationships across multiple source vocabularies. 40 inconsistent concept pairs involving hierarchical relationships were randomly selected and manually reviewed by a human expert. 95.8% of the inconsistent relations involved in these concept pairs indeed exist in their source vocabularies rather than being introduced by mistake in the UMLS integration process. 73.7% of the concept pairs with suggested relationship were agreed by the human expert. The effectiveness of COHeRE indicates that UMLS provides a promising environment to enhance

Risk Assessment of Girls : Are There Any Sex Differences in Risk Factors for Re-offending and in Risk Profiles?

NARCIS (Netherlands)

van der Put, Claudia E.; Dekovic, Maja; Hoeve, Machteld; Stams, Geert Jan J. M.; van der Laan, Peter H.; Langewouters, Femke E. M.

2014-01-01

The aims of this study were (a) to investigate sex differences in risk factors for re-offending and (b) to provide a risk assessment model for girls. The data of 1,396 adolescents who committed a criminal offense were examined. Both generic and sex-specific risk factors for re-offending were found.

Assessment of Safety Culture

International Nuclear Information System (INIS)

Bilic Zabric, T.; Kavsek, D.

2006-01-01

A strong safety culture leads to more effective conduct of work and a sense of accountability among managers and employees, who should be given the opportunity to expand skills by training. The resources expended would thus result in tangible improvements in working practices and skills, which encourage further improvement of safety culture. In promoting an improved safety culture, NEK has emphasized both national and organizational culture with an appropriate balance of behavioural sciences and quality management systems approaches. In recent years there has been particular emphasis put on an increasing awareness of the contribution that human behavioural sciences can make to develop good safety practices. The purpose of an assessment of safety culture is to increase the awareness of the present culture, to serve as a basis for improvement and to keep track of the effects of change or improvement over a longer period of time. There is, however, no single approach that is suitable for all purposes and which can measure, simultaneously, all the intangible aspects of safety culture, i.e. the norms, values, beliefs, attitudes or the behaviours reflecting the culture. Various methods have their strengths and weaknesses. To prevent significant performance problems, self-assessment is used. Self-assessment is the process of identifying opportunities for improvement actively or, in some cases, weaknesses that could cause more serious errors or events. Self-assessments are an important input to the corrective action programme. NEK has developed questionnaires for safety culture self-assessment to obtain information that is representative of the whole organization. Questionnaires ensure a greater degree of anonymity, and create a less stressful situation for the respondent. Answers to questions represent the more apparent and conscious values and attitudes of the respondent. NEK proactively co-operates with WANO, INPO, IAEA in the areas of Safety Culture and Human

A study on optimization of the nuclear safety system

International Nuclear Information System (INIS)

Lee, Sang Hoon; Koh, Byung Joon; Kim, Jin Soo; Kim, Byoung Do; Cho, Seong Won; Kwon, Seog Kwon; Choi, Kwang Sik

1986-12-01

The number of nuclear facilities (nuclear power plants, research reactors, nuclear fuel facilities) under construction or in operation in Korea continues to increase and this has brought about increased importance and concerns toward nuclear safety in Korea. Also, domestic nuclear related organizations are increasingly carrying out the design/construction of nuclear power plants and the development /supply of nuclear fuels. In order to flexibly respond to these changes and to suggest direction to take, it is necessary to re-examine the current nuclear safety regulation system. This study is carried out in two stages and this report describes the results of the analysis and the assessment of the nuclear licencing system of such foreign countries as sweden and German, as the first of the two. In this regard, this study includes the analysis on the backgrounds on the choice of nuclear licensing system, the analysis on the licensing procedures, the analysis on the safety inspection system and the enforcement laws, the analysis on the structure and function of the regulatory, business and research organizations as well as the analysis on the relationship between the safety research and the regulatory duties. In this study, the German safety inspection system and the enforcement procedures and the Swedish nuclear licensing system are analyzed in detail. By comparing and assessing the finding with the current Korea Nuclear Licensing System, this study points out some reform measures of the Korean system that needs to improved. With the changing situations in mind, this study aims to develop the nuclear safety regulation system optimized for Korean situation by re-examining the current regulation system. (Author)

German - Ukrainian collaboration in the assessment of digital I and C systems for safety applications in NPPs

International Nuclear Information System (INIS)

Yastrebenetsky, M.; Vinogradskaia, S.; Wach, D.; Mulka, B.

2001-01-01

German - Ukrainian collaboration in safety assessment of digital Instrumentation and Control (IC) systems began to be in progress since 1995 as part of the established collaboration in the field of Ukrainian NPP safety declared by the German Ministry BMU and Ukrainian Ministry of Environmental Protection and Nuclear Safety and aimed at the support of the Ukrainian Regulatory Body in supervision and licensing of NPPs. The collaboration in IC was triggered by the contract between Rovno NPP (Ukraine) and Siemens (Germany) on procurement of digital emergency protection system for Unit 4. The collaboration has been realized between regulatory authorities and supporting organizations of both countries: GRS/ISTec - Germany and Nuclear Regulatory Authority and State Scientific Technical Center of Nuclear and Radiation Safety (SSTC NRS) - Ukraine. From the beginning the collaboration was intended to cover not only the single specific system, but also a great number of tasks concerned with safety assessment of digital IC systems. As a result the existing Ukrainian standards on IC assessment have been re-evaluated and supplemented by requirements concerning software-based digital IC safety systems. (authors)

German - Ukrainian collaboration in the assessment of digital I and C systems for safety applications in NPPs

Energy Technology Data Exchange (ETDEWEB)

Yastrebenetsky, M.; Vinogradskaia, S. [State Scientific Technical Center of Nuclear and Radiation Safety, Kharkov (Ukraine); Wach, D.; Mulka, B. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

2001-07-01

German - Ukrainian collaboration in safety assessment of digital Instrumentation and Control (IC) systems began to be in progress since 1995 as part of the established collaboration in the field of Ukrainian NPP safety declared by the German Ministry BMU and Ukrainian Ministry of Environmental Protection and Nuclear Safety and aimed at the support of the Ukrainian Regulatory Body in supervision and licensing of NPPs. The collaboration in IC was triggered by the contract between Rovno NPP (Ukraine) and Siemens (Germany) on procurement of digital emergency protection system for Unit 4. The collaboration has been realized between regulatory authorities and supporting organizations of both countries: GRS/ISTec - Germany and Nuclear Regulatory Authority and State Scientific Technical Center of Nuclear and Radiation Safety (SSTC NRS) - Ukraine. From the beginning the collaboration was intended to cover not only the single specific system, but also a great number of tasks concerned with safety assessment of digital IC systems. As a result the existing Ukrainian standards on IC assessment have been re-evaluated and supplemented by requirements concerning software-based digital IC safety systems. (authors)

Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Baji, C.; Vamos, G.; Toth, J.

2001-01-01

The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

IAEA safety requirements for safety assessment of fuel cycle facilities and activities

International Nuclear Information System (INIS)

Jones, G.

2013-01-01

The IAEA's Statute authorises the Agency to establish standards of safety for protection of health and minimisation of danger to life and property. In that respect, the IAEA has established a Safety Fundamentals publication which contains ten safety principles for ensuring the protection of workers, the public and the environment from the harmful effects of ionising radiation. A number of these principles require safety assessments to be carried out as a means of evaluating compliance with safety requirements for all nuclear facilities and activities and to determine the measures that need to be taken to ensure safety. The safety assessments are required to be carried out and documented by the organisation responsible for operating the facility or conducting the activity, are to be independently verified and are to be submitted to the regulatory body as part of the licensing or authorisation process. In addition to the principles of the Safety Fundamentals, the IAEA establishes requirements that must be met to ensure the protection of people and the environment and which are governed by the principles in the Safety Fundamentals. The IAEA's Safety Requirements publication 'Safety Assessment for Facilities and Activities', establishes the safety requirements that need to be fulfilled in conducting and maintaining safety assessments for the lifetime of facilities and activities, with specific attention to defence in depth and the requirement for a graded approach to the application of these safety requirements across the wide range of fuel cycle facilities and activities. Requirements for independent verification of the safety assessment that needs to be carried out by the operating organisation, including the requirement for the safety assessment to be periodically reviewed and updated are also covered. For many fuel cycle facilities and activities, environmental impact assessments and non-radiological risk assessments will be required. The

Risk monitor - a tool for operational safety assessment risk monitor - user's manual

International Nuclear Information System (INIS)

Hari Prasad, M.; Vinod, Gopika; Saraf, R.K.; Ghosh, A.K.

2006-06-01

Probabilistic Safety Assessment has become a key tool as on today to identify and understand Nuclear Power Plant vulnerabilities. As a result of the availability of these PSA studies, there is a desire to use them to enhance plant safety and to operate the nuclear stations in the most efficient manner. Risk Monitor is a PC based tool, which computes the real time safety level and assists plant personnel to manage day-to-day activities. Risk Monitor is a PC based user friendly software tool used for modification and re-analysis of a nuclear Power plant. Operation of Risk Monitor is based on PSA methods for assisting in day to day applications. Risk Monitoring programs can assess the risk profile and are used to optimize the operation of Nuclear Power Plants with respect to a minimum risk level over the operating time. This report presents the background activities of Risk Monitor, its application areas and the step by step procedure for the user.to interact with the software. This software can be used with the PSA model of any Nuclear Power Plant. (author)

Experiences in assessing safety culture

International Nuclear Information System (INIS)

Spitalnik, J.

2002-01-01

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

Safety assessment of discharge chute isolation barrier preparation and installation

International Nuclear Information System (INIS)

Meichle, R.H.

1994-01-01

This analysis examines activities associated with the installation of isolation barriers in the K Basins at the Hanford Reservation. This revision adds evaluation of barrier drops on stored fuel and basin floor, identifies fuel which will be moved and addresses criticality issues with sludge. The safety assessment is made for the activities for the preparation and installation of the discharge chute isolation barriers. The safety assessment includes a hazard assessment and comparisons of potential accidents/events to those addressed by the current safety basis documentation. No significant hazards were identified. An evaluation against the USQ evaluation questions was made and the determination made that the activities do not represent a USQ. Hazard categorization techniques were used to provide a basis for readiness review classifications

OSART Independent Safety Culture Assessment (ISCA) Guidelines

International Nuclear Information System (INIS)

2016-01-01

Safety culture is understood as an important part of nuclear safety performance. This has been demonstrated by the analysis of significant events such as Chernobyl, Davis Besse, Vandellos II, Asco, Paks, Mihamma and Forsmark, among others. In order to enhance safety culture, one essential activity is to perform assessments. IAEA Safety Standard Series No. GS-R-3, The Management System for Facilitites and Activities, states requirements for continuous improvement of safety culture, of which self, peer and independent safety culture assessments constitute an essential part. In line with this requirement, the Independent Safety Culture Assessment (ISCA) module is offered as an add-on module to the IAEA Operational Safety Review Team (OSART) programme. The OSART programme provides advice and assistance to Member States to enhance the safety of nuclear power plants during commissioning and operation. By including the ISCA module in an OSART mission, the receiving organization benefits from the synergy between the technical and the safety culture aspects of the safety review. The joint operational safety and safety culture assessment provides the organization with the opportunity to better understand the interactions between technical, human, organizational and cultural aspects, helping the organization to take a systemic approach to safety through identifying actions that fully address the root causes of any identified issue. Safety culture assessments provide insight into the fundamental drivers that shape organizational patterns of behaviour, safety consciousness and safety performance. The complex nature of safety culture means that the analysis of the results of such assessments is not as straightforward as for other types of assessment. The benefits of the results of nuclear safety culture assessments are maximized only if appropriate tools and guidance for these assessments is used; hence, this comprehensive guideline has been developed. The methodology explained

Metrics design for safety assessment

NARCIS (Netherlands)

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

2016-01-01

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

Safety assessment of cast steel valve housing using NDE- and FM-methods

Energy Technology Data Exchange (ETDEWEB)

Blauel, J G; Hodulak, L [Fraunhofer-Institut fuer Werkstoffmechanik, Freiburg im Breisgau (Germany)

1988-12-31

This document presents some results concerning the safety assessment of cast steel valve housing. This risk assessment is performed through the Non Destructive Examination and Fracture Mechanics methods. (TEC).

Correlation between safety climate and contractor safety assessment programs in construction.

Science.gov (United States)

Sparer, Emily H; Murphy, Lauren A; Taylor, Kathryn M; Dennerlein, Jack T

2013-12-01

Contractor safety assessment programs (CSAPs) measure safety performance by integrating multiple data sources together; however, the relationship between these measures of safety performance and safety climate within the construction industry is unknown. Four hundred and one construction workers employed by 68 companies on 26 sites and 11 safety managers employed by 11 companies completed brief surveys containing a nine-item safety climate scale developed for the construction industry. CSAP scores from ConstructSecure, Inc., an online CSAP database, classified these 68 companies as high or low scorers, with the median score of the sample population as the threshold. Spearman rank correlations evaluated the association between the CSAP score and the safety climate score at the individual level, as well as with various grouping methodologies. In addition, Spearman correlations evaluated the comparison between manager-assessed safety climate and worker-assessed safety climate. There were no statistically significant differences between safety climate scores reported by workers in the high and low CSAP groups. There were, at best, weak correlations between workers' safety climate scores and the company CSAP scores, with marginal statistical significance with two groupings of the data. There were also no significant differences between the manager-assessed safety climate and the worker-assessed safety climate scores. A CSAP safety performance score does not appear to capture safety climate, as measured in this study. The nature of safety climate in construction is complex, which may be reflective of the challenges in measuring safety climate within this industry. Am. J. Ind. Med. 56:1463-1472, 2013. © 2013 Wiley Periodicals, Inc. © 2013 Wiley Periodicals, Inc.

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

Energy Technology Data Exchange (ETDEWEB)

Ghofrani, Javad

2016-05-26

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

International Nuclear Information System (INIS)

Ghofrani, Javad

2016-01-01

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

Safety assessment principles for nuclear plants

International Nuclear Information System (INIS)

1992-01-01

The present Safety Assessment Principles result from the revision of those which were drawn up following a recommendation arising from the Sizewell-B enquiry. The principles presented here relate only to nuclear safety; there is a section on risks from normal operation and accident conditions and the standards against which those risks are assessed. A major part of the document deals with the principles that cover the design of nuclear plants. The revised Safety assessment principles are aimed primarily at the safety assessment of new nuclear plants but they will also be used in assessing existing plants. (UK)

Development of safety related technology and infrastructure for safety assessment

International Nuclear Information System (INIS)

Venkat Raj, V.

1997-01-01

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

Thinking of the safety assessment of HLW disposal

International Nuclear Information System (INIS)

Li Honghui; Zhao Shuaiwei; Liu Jianqin; Liu Wei; Wan Lei; Yang Zhongtian; An Hongxiang; Sun Qinghong

2014-01-01

The function and the research methods of safety assessment are discussed. Two methods about safety assessment and the requirement of safety assessment are introduced. The key parameters and influence factors in nuclide transport of safety assessment are specialized. The works will be done on safety assessment is discussed which will give some suggests for the development of safety assessment. (authors)

ALARP considerations in criticality safety assessments

International Nuclear Information System (INIS)

Bowden, Russell L.; Barnes, Andrew; Thorne, Peter R.; Venner, Jack

2003-01-01

Demonstrating that the risk to the public and workers is As Low As Reasonably Practicable (ALARP) is a fundamental requirement of safety cases for nuclear facilities in the United Kingdom. This is embodied in the Safety Assessment Principles (SAPs) published by the Regulator, the essence of which is incorporated within the safety assessment processes of the various nuclear site licensees. The concept of ALARP within criticality safety assessments has taken some time to establish in the United Kingdom. In principle, the licensee is obliged to search for a deterministic criticality safety solution, such as safe geometry vessels and passive control features, rather than placing reliance on active measurement devices and plant administrative controls. This paper presents a consideration of some ALARP issues in relation to the development of criticality safety cases. The paper utilises some idealised examples covering a range of issues facing the criticality safety assessor, including new plant design, operational plant and decommissioning activities. These examples are used to outline the elements of the criticality safety cases and present a discussion of ALARP in the context of criticality safety assessments. (author)

Safety culture assessment developed by JANTI

International Nuclear Information System (INIS)

Hamada, Jun

2009-01-01

Japan's JCO accident in September 1999 provided a real-life example of what can happen when insufficient attention is paid to safety culture. This accident brought to light the importance of safety culture and reinforced the movement to foster a safety culture. Despite this, accidents and inappropriate conduct have continued to occur. Therefore, there is a strong demand to instill a safety culture throughout the nuclear power industry. In this context, Japan's nuclear power regulator, the Nuclear and Industrial Safety Agency (NISA), decided to include in its safety inspections assessments of the safety culture found in power utilities' routine safety operations to get signs of deterioration in the organizational climate. In 2007, NISA constructed guidelines for their inspectors to carry out these assessments. At the same time, utilities have embarked on their own independent safety culture initiatives, such as revising their technical specifications and building effective PDCA cycle to promote safety culture. In concert with these developments, JANTI has also instituted safety culture assessments. (author)

H12: Examination of safety assessment aims, procedures and results from a wider perspective

International Nuclear Information System (INIS)

Neall, F.B; Smith, P.A.

2004-04-01

Safety assessment (SA) are a familiar tool for the evaluation of disposal concepts for radioactive waste. There is, however, often confusion in the wider community about the aims, methods and results used in SA. This report aims to present the H12 SA in a way that makes the assessment process clearer and the implications of the results more meaningful both to workers within the SA field and to a wider technical audience. The reasonableness of the assessment results, the quality of the models and databases and redundancy within the natural and engineered barrier system have been considered. A number of recent and somewhat older SAs that address a range of different waste types, host rocks and disposal concepts have been considered, and comparisons made to H12. A further aim is to put both doses and timescales in a more meaningful context. It has been necessary to: consider ways of demonstrating the meaningfulness of calculations that give results for many thousands of years in the future; provide a framework timescale as a context for SA results over long times; demonstrate the smallness of the risk associated with the doses by comparison with other radiological and non-radiological risks. The perception of risk, which is a critical issue for public acceptance of radioactive waste disposal and must be considered when seeking to present safety assessment results 'in perspective' to a wider audience, is also discussed. It is concluded that H12 is comparable in many ways to assessments carried out internationally. Some assumptions are somewhat arbitrary reflecting the generic stage of the Japanese programme, and are likely to become better founded in future exercises. Nevertheless, H12 provides a clear and well-founded message that it is feasible to site and construct a safe repository from HLW in Japan. (author)

Specifications of the International Atomic Energy Agency's international project on safety assessment driven radioactive waste management solutions

International Nuclear Information System (INIS)

Ghannadi, M.; Asgharizadeh, F.; Assadi, M. R.

2008-01-01

Radioactive waste is produced in the generation of nuclear power and the production and use of radioactive materials in the industry, research, and medicine. The nuclear waste management facilities need to perform a safety assessment in order to ensure the safety of a facility. Nuclear safety assessment is a structured and systematic way of examining a proposed facility, process, operation and activity. In nuclear waste management point of view, safety assessment is a process which is used to evaluate the safety of radioactive waste management and disposal facilities. In this regard the International Atomic Energy Agency is planed to implement an international project with cooperation of some member states. The Safety Assessment Driving Radioactive Waste Management Solutions Project is an international programme of work to examine international approaches to safety assessment in aspects of p redisposal r adioactive waste management, including waste conditioning and storage. This study is described the rationale, common aspects, scope, objectives, work plan and anticipated outcomes of the project with refer to International Atomic Energy Agency's documents, such as International Atomic Energy Agency's Safety Standards, as well as the Safety Assessment Driving Radioactive Waste Management Solutions project reports

Rad waste disposal safety analysis / Integrated safety assessment of a waste repository

International Nuclear Information System (INIS)

Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

2012-04-01

We developed CYPRUS+and adopted PID and RES method for the development of scenario. Safety performance assessment program was developed using GoldSim for the safety assessment of disposal system for the disposal of spnet fuels and wastes resulting from the pyrpoprocessing. Biosphere model was developed and verified in cooperation with JAEA. The capability to evaluate post-closure performance and safety was added to the previously developed program. And, nuclide migration and release to the biosphere considering site characteristics was evaluated by using deterministic and probabilistic approach. Operational safety assessment for drop, fire, and earthquake was also statistically evaluated considering well-established input parameter distribution. Conservative assessment showed that dose rate is below the limit value of low- and intermediate-level repository. Gas generation mechanism within engineered barrier was defined and its influence on safety was evaluated. We made probabilistic safety assessment by obtaining the probability distribution functions of important input variables and also made a sensitivity analysis. The maximum annual dose rate was shown to be below the safety limit value of 10 mSv/yr. The structure and element of safety case was developed to increase reliability of safety assessment methodology for a deep geological repository. Finally, milestone for safety case development and implementation strategy for each safety case element was also proposed

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

International Nuclear Information System (INIS)

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

2002-03-01

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

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

Questionnaire responses concerning safety issues in MR examination

International Nuclear Information System (INIS)

Yamaguchi-Sekino, Sachiko; Nakai, Toshiharu; Muranaka, Hiroyuki

2011-01-01

Recently, the rising numbers of medical implants and scanners with higher static magnetic field have increased safety concerns for magnetic resonance (MR) examination. To determine future safety focus, we distributed anonymous questionnaires to 3250 members of the Japanese Society for Magnetic Resonance in Medicine (JSMRM) and received 978 responses. Safety issues on the questionnaire concentrated on the handling of patients with implants (Q7-18, appendix), acoustic trauma due to scanning (Q19-21, appendix), and MR compatibility within the scanner room (Q22-25, appendix). Ninety-three percent of respondents indicated they had encountered cases with implants or medical materials of unknown MR compatibility; 21.7% reported heating problems and 15.0%, nerve stimulation problems, in patients with implants during MR examination. Although 88.7% of respondents recognized the term ''MR compatibility'', 68.2% indicated limited detailed understanding of the term. Eleven percent had had cases with suspected acoustic injury from MR scanner noise. Scanner noise levels were not clarified in any way in 37.4% cases, but 69.5% applied ear protection to patients. Labeling of ''MR compatibility'' of equipment brought into the MR scanner room was reported by 71.9%. More than 50% experienced MR compatibility issues related to equipment brought into the MR scanner room. With regard to safety issues on metallic objects which are implanted in MR workers, 88.1% indicated they would continue current operations even the implant is inside the body. Respondents identified lectures and seminars by professional societies, safety training by manufacturers, and information from the Internet and literature as the 3 main sources for up-dating safety information for MR examination. (author)

Marked point process framework for living probabilistic safety assessment and risk follow-up

International Nuclear Information System (INIS)

Arjas, Elja; Holmberg, Jan

1995-01-01

We construct a model for living probabilistic safety assessment (PSA) by applying the general framework of marked point processes. The framework provides a theoretically rigorous approach for considering risk follow-up of posterior hazards. In risk follow-up, the hazard of core damage is evaluated synthetically at time points in the past, by using some observed events as logged history and combining it with re-evaluated potential hazards. There are several alternatives for doing this, of which we consider three here, calling them initiating event approach, hazard rate approach, and safety system approach. In addition, for a comparison, we consider a core damage hazard arising in risk monitoring. Each of these four definitions draws attention to a particular aspect in risk assessment, and this is reflected in the behaviour of the consequent risk importance measures. Several alternative measures are again considered. The concepts and definitions are illustrated by a numerical example

Examination of the Safety of Pediatric Vaccine Schedules in a Non-Human Primate Model: Assessments of Neurodevelopment, Learning, and Social Behavior

Science.gov (United States)

Curtis, Britni; Liberato, Noelle; Rulien, Megan; Morrisroe, Kelly; Kenney, Caroline; Yutuc, Vernon; Ferrier, Clayton; Marti, C. Nathan; Mandell, Dorothy; Burbacher, Thomas M.; Sackett, Gene P.

2015-01-01

Background In the 1990s, the mercury-based preservative thimerosal was used in most pediatric vaccines. Although there are currently only two thimerosal-containing vaccines (TCVs) recommended for pediatric use, parental perceptions that vaccines pose safety concerns are affecting vaccination rates, particularly in light of the much expanded and more complex schedule in place today. Objectives The objective of this study was to examine the safety of pediatric vaccine schedules in a non-human primate model. Methods We administered vaccines to six groups of infant male rhesus macaques (n = 12–16/group) using a standardized thimerosal dose where appropriate. Study groups included the recommended 1990s Pediatric vaccine schedule, an accelerated 1990s Primate schedule with or without the measles–mumps–rubella (MMR) vaccine, the MMR vaccine only, and the expanded 2008 schedule. We administered saline injections to age-matched control animals (n = 16). Infant development was assessed from birth to 12 months of age by examining the acquisition of neonatal reflexes, the development of object concept permanence (OCP), computerized tests of discrimination learning, and infant social behavior. Data were analyzed using analysis of variance, multilevel modeling, and survival analyses, where appropriate. Results We observed no group differences in the acquisition of OCP. During discrimination learning, animals receiving TCVs had improved performance on reversal testing, although some of these same animals showed poorer performance in subsequent learning-set testing. Analysis of social and nonsocial behaviors identified few instances of negative behaviors across the entire infancy period. Although some group differences in specific behaviors were reported at 2 months of age, by 12 months all infants, irrespective of vaccination status, had developed the typical repertoire of macaque behaviors. Conclusions This comprehensive 5-year case–control study, which closely examined

The Safety Case and Safety Assessment for the Disposal of Radioactive Waste

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-09-15

This Safety Guide provides guidance and recommendations on meeting the safety requirements in respect of the safety case and supporting safety assessment for the disposal of radioactive waste. The safety case and supporting safety assessment provide the basis for demonstration of safety and for licensing of radioactive waste disposal facilities and assist and guide decisions on siting, design and operations. The safety case is also the main basis on which dialogue with interested parties is conducted and on which confidence in the safety of the disposal facility is developed. This Safety Guide is relevant for operating organizations preparing the safety case as well as for the regulatory body responsible for developing the regulations and regulatory guidance that determine the basis and scope of the safety case. Contents: 1. Introduction; 2. Demonstrating the safety of radioactive waste disposal; 3. Safety principles and safety requirements; 4. The safety case for disposal of radioactive waste; 5. Radiological impact assessment for the period after closure; 6. Specific issues; 7. Documentation and use of the safety case; 8. Regulatory review process.

Re-assessment of safety and ageing; Reevaluations de surete et vieillissement

Energy Technology Data Exchange (ETDEWEB)

Boudot, R.; Dallery, D. [Electricite de France (EDF/DPN), Div. Production Nucleaire 93 - Saint-Denis (France); Delmas, J. [Electricite de France (EDF/DIN), 93 - Saint-Denis (France); Balley, J.; Acalet, R. [Electricite de France (EDF-CIPN), 13 - Marseille (France); Bignan, G.; Mergui, C.; Mazoyer, R.; Maegey, M. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Laigneau, P. [AREVA NP, 92 - Paris La Defense (France)

2007-03-15

The first article presents the research program led conjointly by Cea, EDF and Areva to face ageing of nuclear power plants. 3 axis have been defined: 1) the damaging processes in equipment, 2) the damaging processes in concrete structures, and 3) inspection and monitoring techniques. In the context of the preparation of the third decennial safety review of its 900 MWe PWR units, EDF was asked by the Nuclear Authority (ANS) to write up documentation (DAPE) whose purpose is to show that the aging processes are mastered and adequately handled in the unit concerned and that the carrying on of reactor operation for the next 10 years will not degrade its safety standard. The second article presents the specific purposes of the second (VD2) and third (VD3) decennial safety reviews. The last article deals with the renovation work planned on the Masurca research reactor. Masurca installation is dedicated to the setting of neutron experiments necessary for the validation of neutron calculations in fast neutron spectra configurations. The reactor upgrade will concern: -) the power system, -) the instrumentation and control of the reactor, -) the ventilation of the reactor building, -) the seismic reinforcement of an auxiliary building, -) fire protection, and -) radiation monitoring. (A.C.)

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

International Nuclear Information System (INIS)

2009-01-01

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

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

International Nuclear Information System (INIS)

2009-01-01

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

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

International Nuclear Information System (INIS)

2010-01-01

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

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

International Nuclear Information System (INIS)

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Regulatory review of safety cases and safety assessments - associated challenges

International Nuclear Information System (INIS)

Bennett, D.G.; Ben Belfadhel, M.; Metcalf, P.E.

2006-01-01

Regulatory reviews of safety cases and safety assessments are essential for credible decision making on the licensing or authorization of radioactive waste disposal facilities. Regulatory review also plays an important role in developing the safety case and in establishing stakeholders' confidence in the safety of the facility. Reviews of safety cases for radioactive waste disposal facilities need to be conducted by suitably qualified and experienced staff, following systematic and well planned review processes. Regulatory reviews should be sufficiently comprehensive in their coverage of issues potentially affecting the safety of the disposal system, and should assess the safety case against clearly established criteria. The conclusions drawn from a regulatory review, and the rationale for them should be reproducible and documented in a transparent and traceable way. Many challenges are faced when conducting regulatory reviews of safety cases. Some of these relate to issues of project and programme management, and resources, while others derive from the inherent difficulties of assessing the potential long term future behaviour of engineered and environmental systems. The paper describes approaches to the conduct of regulatory reviews and discusses some of the challenges faced. (author)

Intermediate probabilistic safety assessment approach for safety critical digital systems

International Nuclear Information System (INIS)

Taeyong, Sung; Hyun Gook, Kang

2001-01-01

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

Safety assessments for potential exposures

International Nuclear Information System (INIS)

Dunn, D.I.

2012-04-01

Safety Assessment of potential exposures have been carried out in major practices, namely: industrial radiography, gamma irradiators and electron accelerators used in industry and research, and radiotherapy. This paper focuses on reviewing safety assessment methodologies and using developed software to analyse radiological accidents, also review, and discuss these past accidents.The primary objective of the assessment is to assess the adequacy of planned or existing measures for protection and safety and to identify any additional measures that should be put in place. As such, both routine use of the source and the probability and magnitude of potential exposures arising from accidents or incidents should be considered. Where the assessment indicates that there is a realistic possibility of an accident affecting workers or members of the public or having consequences for the environment, the registrant or licensee should prepare a suitable emergency plan. A safety assessment for normal operation addresses all the conditions under which the radiation source operates as expected, including all phases of the lifetime of the source. Due account needs to be taken of the different factors and conditions that will apply during non-operational phases, such as installation, commissioning and maintenance. (author)

Development of a Safety Assessment Information System for the Management of Periodic Safety Assessment Activities

International Nuclear Information System (INIS)

Song, Tae Young

2007-01-01

At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas

Development of a Safety Assessment Information System for the Management of Periodic Safety Assessment Activities

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Young [Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)

2007-07-01

At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas.

Safety assessment for spent fuel storage facilities

International Nuclear Information System (INIS)

1994-01-01

This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

'Perspectives' re-examined

International Nuclear Information System (INIS)

Daglish, J.

1980-01-01

A one day conference held in September 1980 to discuss radioactive waste management is reviewed. Quotations and summaries are given from a number of the papers presented including those dealing with the safety, costs, treatment and disposal of high-level radioactive wastes, the conversion of these wastes to solid forms (vitrification and other immobilisation techniques), the importance of timing in disposal operations due to the heat ouput of the wastes, container engineering, geological aspects of radioactive waste repositories, and public attitudes. (U.K.)

'Perspectives' re-examined

Energy Technology Data Exchange (ETDEWEB)

Daglish, J

1980-12-01

A one-day conference held in September 1980 to discuss radioactive waste management is reviewed. Quotations and summaries are given from a number of the papers presented including those dealing with the safety, costs, treatment and disposal of high-level radioactive wastes, the conversion of these wastes to solid forms (vitrification and other immobilisation techniques), the importance of timing in disposal operations due to the heat output of the wastes, container engineering, geological aspects of radioactive waste repositories, and public attitudes.

DMPD: Cytokines, PGE2 and endotoxic fever: a re-assessment. [Dynamic Macrophage Pathway CSML Database

Lifescience Database Archive (English)

Full Text Available 15967158 Cytokines, PGE2 and endotoxic fever: a re-assessment. Blatteis CM, Li S, L... (.svg) (.html) (.csml) Show Cytokines, PGE2 and endotoxic fever: a re-assessment. PubmedID 15967158 Title C...ytokines, PGE2 and endotoxic fever: a re-assessment. Authors Blatteis CM, Li S, L

Consideration of aging in probabilistic safety assessment

International Nuclear Information System (INIS)

Titina, B.; Cepin, M.

2007-01-01

Probabilistic safety assessment is a standardised tool for assessment of safety of nuclear power plants. It is a complement to the safety analyses. Standard probabilistic models of safety equipment assume component failure rate as a constant. Ageing of systems, structures and components can theoretically be included in new age-dependent probabilistic safety assessment, which generally causes the failure rate to be a function of age. New age-dependent probabilistic safety assessment models, which offer explicit calculation of the ageing effects, are developed. Several groups of components are considered which require their unique models: e.g. operating components e.g. stand-by components. The developed models on the component level are inserted into the models of the probabilistic safety assessment in order that the ageing effects are evaluated for complete systems. The preliminary results show that the lack of necessary data for consideration of ageing causes highly uncertain models and consequently the results. (author)

Five-Year NRHP Re-Evaluation of Historic Buildings Assessment

Energy Technology Data Exchange (ETDEWEB)

Ullrich, R A; Heidecker, K R

2011-09-12

The Lawrence Livermore National Laboratory (LLNL) 'Draft Programmatic Agreement among the Department of Energy and the California State Historic Preservation Officer Regarding Operation of Lawrence Livermore National Laboratory' requires a review and re-evaluation of the eligibility of laboratory properties for the National Register of Historic Places (NRHP) every five years. The original evaluation was published in 2005; this report serves as the first five-year re-evaluation. This re-evaluation includes consideration of changes within LLNL to management, to mission, and to the built environment. it also determines the status of those buildings, objects, and districts that were recommended as NRHP-eligible in the 2005 report. Buildings that were omitted from the earlier building list, those that have reached 50 years of age since the original assessment, and new buildings are also addressed in the re-evaluation.

A Methodology for Safety Culture Impact Assessment

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

A Methodology for Safety Culture Impact Assessment

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2014-01-01

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

BARC-risk monitor- a tool for operational safety assessment in nuclear power plants

International Nuclear Information System (INIS)

Vinod, Gopika; Saraf, R.K.; Babar, A.K.; Hadap, Nikhil

2000-12-01

Probabilistic safety assessment has become a key tool as on today to identify and understand nuclear power plant vulnerabilities. As a result of the availability of these PSA studies, there is a desire to use them to enhance plant safety and to operate the nuclear stations in the most efficient manner. Risk monitor is a PC based tool, which computes the real time safety level and assists plant personnel to manage day-to-day activities. Risk monitor is a PC based user friendly software tool used for modification and re-analysis of a nuclear power plant. Operation of risk monitor is based on PSA methods for assisting in day to day applications. Risk monitoring programs can assess the risk profile and are used to optimise the operation of nuclear power plants with respect to a minimum risk level over the operating time. This report presents the background activities of risk monitor, its application areas and also gives the status of such tools in international scenarios. The software is based on the PSA model of Kaiga generating station and would be applicable to similar design configuration. (author)

Regulatory review of safety cases and safety assessments for near surface

International Nuclear Information System (INIS)

Nys, V.

2003-01-01

The activities of the ASAM Regulatory Review Working Group are presented. Regulatory review of the safety assessment is made. It includes the regulatory review of post-closure safety assessment; safety case development and confidence building. The ISAM methodology is reviewed and SA system description is presented. Recommendations on the review process management are given

Safety Auditing and Assessments

Science.gov (United States)

Goodin, James Ronald (Ronnie)

2005-01-01

Safety professionals typically do not engage in audits and independent assessments with the vigor as do our quality brethren. Taking advantage of industry and government experience conducting value added Independent Assessments or Audits benefits a safety program. Most other organizations simply call this process "internal audits." Sources of audit training are presented and compared. A relation of logic between audit techniques and mishap investigation is discussed. An example of an audit process is offered. Shortcomings and pitfalls of auditing are covered.

Using driving simulators to assess driving safety.

Science.gov (United States)

Boyle, Linda Ng; Lee, John D

2010-05-01

Changes in drivers, vehicles, and roadways pose substantial challenges to the transportation safety community. Crash records and naturalistic driving data are useful for examining the influence of past or existing technology on drivers, and the associations between risk factors and crashes. However, they are limited because causation cannot be established and technology not yet installed in production vehicles cannot be assessed. Driving simulators have become an increasingly widespread tool to understand evolving and novel technologies. The ability to manipulate independent variables in a randomized, controlled setting also provides the added benefit of identifying causal links. This paper introduces a special issue on simulator-based safety studies. The special issue comprises 25 papers that demonstrate the use of driving simulators to address pressing transportation safety problems and includes topics as diverse as neurological dysfunction, work zone design, and driver distraction. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

Confidence building in safety assessments

International Nuclear Information System (INIS)

Grundfelt, Bertil

1999-01-01

Future generations should be adequately protected from damage caused by the present disposal of radioactive waste. This presentation discusses the core of safety and performance assessment: The demonstration and building of confidence that the disposal system meets the safety requirements stipulated by society. The major difficulty is to deal with risks in the very long time perspective of the thousands of years during which the waste is hazardous. Concern about these problems has stimulated the development of the safety assessment discipline. The presentation concentrates on two of the elements of safety assessment: (1) Uncertainty and sensitivity analysis, and (2) validation and review. Uncertainty is associated both with respect to what is the proper conceptual model and with respect to parameter values for a given model. A special kind of uncertainty derives from the variation of a property in space. Geostatistics is one approach to handling spatial variability. The simplest way of doing a sensitivity analysis is to offset the model parameters one by one and observe how the model output changes. The validity of the models and data used to make predictions is central to the credibility of safety assessments for radioactive waste repositories. There are several definitions of model validation. The presentation discusses it as a process and highlights some aspects of validation methodologies

Examining the Relationship between Safety Management System Implementation and Safety Culture in Collegiate Flight Schools

Science.gov (United States)

Robertson, Mike Fuller

2017-01-01

Safety Management Systems (SMS) are becoming the industry standard for safety management throughout the aviation industry. As the Federal Aviation Administration (FAA) continues to mandate SMS for different segments, the assessment of an organization's safety culture becomes more important. An SMS can facilitate the development of a strong…

Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

Energy Technology Data Exchange (ETDEWEB)

Kinker, M., E-mail: M.Kinker@iaea.org [International Atomic Energy Agency (IAEA), Vienna (Austria); Avila, R.; Hofman, D., E-mail: rodolfo@facilia.se [FACILIA AB, Stockholm (Sweden); Jova Sed, L., E-mail: jovaluis@gmail.com [Centro Nacional de Seguridad Nuclear (CNSN), La Habana (Cuba); Ledroit, F., E-mail: frederic.ledroit@irsn.fr [IRSN PSN-EXP/SSRD/BTE, (France)

2013-07-01

In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

International Nuclear Information System (INIS)

Kinker, M.; Avila, R.; Hofman, D.; Jova Sed, L.; Ledroit, F.

2013-01-01

In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

Safety assessment of the magnetic resonance imaging (MRI) facility at the 37 Military Hospital

International Nuclear Information System (INIS)

Mintah, R.O

2010-01-01

Safety assessment of the operation of the Magnetic Resonance Imaging (MRI) system at 37 Military Hospital was done. Protocols were developed to assess the radiological health and safety impact of some selected MR imaging procedures on patients, staff and the general public. The parameters considered to be assessed were; specific absorption rate (SAR); temperature rise in the body, variation of the magnetic field gradients, RF energy used, sound pressure level and the potential for missile effects. The Smart brain and Routine lumbar Spine examination cards specific to brain and lumbar spine anatomy techniques were used. For brain examinations the T1 W-SE sagittal PH, and the T1 W-SE Tra-PH protocols gave the highest SAR values with a mean value 1.6 W/kg for 3 minutes. For the lumbar spine examinations, the T1 W-TSE axial protocols exposed patients to the highest consistent SAR value of 2.8 W /kg. The T2W-TSE axial protocol gave the highest SAR value of 3.8 W/kg with a mean value of 3.1 W/kg with the highest exposure time of 4 minutes with a mean of 2.5 minutes. These SAR values were within the limits recommended by the United States Food and Drug Administration (USFDA). To optimize patient, staff and public safety, local guidelines for safety assessment were developed which include: in house screening with a metal detector, filling in the screening form and ensuring that safety requirements are met before entering the MR room. The choice of protocols that minimize SAR values and strict compliance to safety protocols developed at the MRI facility therefore should be followed and continuously updated to achieve maximum safety for staff, patient and the general public in and around an MRI facility. (au)

Fatigue in seafarers working in the offshore oil and gas re-supply industry: effects of safety climate, psychosocial work environment and shift arrangement.

Science.gov (United States)

Hystad, Sigurd W; Saus, Evelyn-Rose; Sætrevik, Bjørn; Eid, Jarle

2013-01-01

This study examined the influence of safety climate and psychosocial work environment on the reported fatigue of seafarers working in the offshore oil and gas re-supply industry (n = 402). We found that seafarers who reported high psychological demands and perceived the organisational-level safety climate negatively,reported significantly more mental fatigue, physical fatigue, and lack of energy. In addition, seafarers who reported having high levels of job control reported being significantly less mentally fatigued. We also found some combined effects of safety climate and shift arrangement. Organisational-level safety climate did not influence the levels of physical fatigue in seafarers working on the night shift. On the contrary, seafarers working during the days reported to be more physically fatigued when they perceived the organisational-level climate to be negative compared with the positive. The opposite effect was found for group-level safety climate: seafarers working during the nights reported to be more physically fatigued when they perceived the group-level climate to be negative compared with the positive. The results from this study point to the importance of taking into consideration aspects of the psychosocial work environment and safety climate,and their potential impact on fatigue and safety in the maritime organisations.

Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Syarip, S., E-mail: syarip@batan.go.id [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

2014-10-15

The self-assessment of safety culture and safety management status of Kartini research reactor is a step to foster safety culture and management by identifying good practices and areas for improvement, and also to improve reactor safety in a whole. The method used in this assessment is based on questionnaires provided by the Forum for Nuclear Cooperation in Asia (FNCA), then reviewed by experts. Based on the assessment and evaluation results, it can be concluded that there were several good practices in maintaining the safety status of Kartini reactor such as: reactor operators and radiation protection workers were aware and knowledgeable of the safety standards and policies that apply to their operation, readily accept constructive criticism from their management and from the inspectors of regulatory body that address safety performance. As a proof, for the last four years the number of inspection/audit findings from Regulatory Body (BAPETEN) tended to decrease while the reactor utilization and its operating hour increased. On the other hands there were also some comments and recommendations for improvement of reactor safety culture, such as that there should be more frequent open dialogues between employees and managers, to grow and attain a mutual support to achieve safety goals. (author)

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

International Nuclear Information System (INIS)

Chung, Dae Wook

2011-01-01

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

Safety assessment of indoor live fire range, May 1989

Energy Technology Data Exchange (ETDEWEB)

NONE

1989-05-01

The following Safety Assessment (SA) pertains to the indoor live fire range (LFR) at EG&G Mound Applied Technology plant. The purpose of the indoor LFR is to conduct training with live ammunition for all designated personnel. The SA examines the risks that are attendant to the operation of an indoor LFR for this purpose.

Terrain Safety Assessment in Support of the Mars Science Laboratory Mission

Science.gov (United States)

Kipp, Devin

2012-01-01

In August 2012, the Mars Science Laboratory (MSL) mission will pioneer the next generation of robotic Entry, Descent, and Landing (EDL) systems by delivering the largest and most capable rover to date to the surface of Mars. The process to select the MSL landing site took over five years and began with over 50 initial candidate sites from which four finalist sites were chosen. The four finalist sites were examined in detail to assess overall science merit, EDL safety, and rover traversability on the surface. Ultimately, the engineering assessments demonstrated a high level of safety and robustness at all four finalist sites and differences in the assessment across those sites were small enough that neither EDL safety nor rover traversability considerations could significantly discriminate among the final four sites. Thus the MSL landing site at Gale Crater was selected from among the four finalists primarily on the basis of science considerations.

Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

International Nuclear Information System (INIS)

Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

2008-01-01

Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design

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

International Nuclear Information System (INIS)

1992-01-01

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

Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Independent assessment for new nuclear reactor safety

Directory of Open Access Journals (Sweden)

D'Auria Francesco

2017-01-01

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

Independent assessment for new nuclear reactor safety

International Nuclear Information System (INIS)

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

2017-01-01

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

Safety assessment and quality control of medical x-ray facilities in some hospitals in Ghana

International Nuclear Information System (INIS)

Darko, E.O.; Charles, D.F.

1998-01-01

Safety assessment and quality control measurements of diagnostic x-ray installations were carried out in five hospitals in Ghana. The study was focused on the siting, design and construction of the buildings housing the x-ray units, assessment of safety systems and devices and measurements of the technical performance, and film processing conditions. The location, inadequacies in the design/construction, unavailability of relevant safety systems and devices, violation of basic safety principles and poor performance of some of the x-ray facilities indicate the need to improve quality control programmes, safety culture and enforcement of regulatory standards in diagnostic x-ray examinations in Ghana. (author). 8 refs., 11 tabs., 8 figs

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

Science.gov (United States)

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

2015-04-22

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

Procedures to relate the NII safety assessment principles for nuclear reactors to risk

CERN Document Server

Kelly, G N; Hemming, C R

1985-01-01

Within the framework of the Public Inquiry into the proposed pressurised water reactor (PWR) at Sizewell, estimates were made of the levels of individual and societal risk from a PWR designed in a manner which would conform to the safety assessment principles formulated by the Nuclear Installations Inspectorate (NII). The procedures used to derive these levels of risk are described in this report. The opportunity has also been taken to revise the risk estimates made at the time of the Inquiry by taking account of additional data which were not then available, and to provide further quantification of the likely range of uncertainty in the predictions. This re-analysis has led to small changes in the levels of risk previously evaluated, but these are not sufficient to affect the broad conclusions reached before. For a reactor just conforming to the NII safety assessment principles a maximum individual risk of fatal cancer of about 10 sup - sup 6 per year of reactor operation has been estimated; the societal ris...

AGR core safety assessment methodologies

International Nuclear Information System (INIS)

McLachlan, N.; Reed, J.; Metcalfe, M.P.

1996-01-01

To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

Integral Safety Assessment of Underground Storage of CO2 in Barendrecht, the Netherlands

International Nuclear Information System (INIS)

Vijgen, L.; Nitert, M.; Buijtendijk, B.; Van Dalen, A.

2009-10-01

The DCMR Environmental Protection Agency Rijnmond in the Netherlands conducted an Integral Safety Assessment of Underground Storage of CO2 in Barendrecht, the Netherlands, in cooperation with the involved safety and supervision authorities. The following aspects of the entire storage project and its safety issues have been examined: the compressor station in Pernis; the underground pipes between the compressor station and the injection locations; and the injection locations Barendrecht-Ziedewij and Barendrecht. [nl

Uncorrected and Corrected Distance Visual Acuity, Predictability, Efficacy, and Safety after Femtosecond Laser in Situ Keratomileusis (FS-LASIK) and Refractive Lenticule extraction (ReLEx) for Moderate and High Myopia

DEFF Research Database (Denmark)

Vestergaard, Anders; Justesen, Birgitte Larsen; Melsen, Charlotte

Title: Uncorrected and Corrected Distance Visual Acuity, Predictability, Efficacy, and Safety after Femtosecond Laser in Situ Keratomileusis (FS-LASIK) and Refractive Lenticule extraction (ReLEx) for Moderate and High Myopia. Vestergaard A., Justesen B., Melsen C., Lyhne N., Department of Ophthal......Title: Uncorrected and Corrected Distance Visual Acuity, Predictability, Efficacy, and Safety after Femtosecond Laser in Situ Keratomileusis (FS-LASIK) and Refractive Lenticule extraction (ReLEx) for Moderate and High Myopia. Vestergaard A., Justesen B., Melsen C., Lyhne N., Department...... predictability, efficacy and safety after femtosecond LASIK (FS-LASIK) with ReLEx. Setting: Department of Ophthalmology, Odense University Hospital, Denmark. Methods: Retrospective study of results after FS-LASIK and ReLEx (including ReLEx flex, ReLEx pseudo-smile, and ReLEx smile). In total, 228 eyes were...... treated with FS-LASIK and 83 eyes with ReLEx, at the Department of Ophthalmology, Odense University Hospital in the period of April to November 2011. Only otherwise healthy myopic eyes with up to 3.00 D of astigmatism and with CDVA ≤ 0.30 (logMAR) before surgery were included in this study. FS-LASIK flaps...

The Impact of Family Engagement and Child Welfare Services on Maltreatment Re-reports and Substantiated Re-reports.

Science.gov (United States)

Fuller, Tamara; Zhang, Saijun

2017-08-01

Despite decades of debate about the most effective ways to intervene with families reported to child protective services (CPS), little evidence exists regarding the types of services or approach that reduce children's risk of additional maltreatment. The current study used data collected during a statewide experimental evaluation of CPS to examine the impact of numerous service variables, family engagement, and family characteristics on the risk of maltreatment re-reports and substantiated re-reports among families initially reported for neglect and risk of harm. The sample included 4,868 families with screened-in reports that were randomly assigned to receive either an investigation or an assessment. The results of the Cox regression analyses found that service duration, intensity, and breadth were unrelated to maltreatment re-report or substantiated re-reports, but caseworker ratings of the service-need match were associated with both. The provision of domestic violence services was related to decreased risk of maltreatment re-reports. Increased levels of family engagement were associated with lowered risk of both maltreatment re-reports and substantiated re-reports. Once the effects of services, engagement, and family characteristics were taken into account, CPS response pathway (investigation or assessment) had no relationship to maltreatment re-reports or substantiated re-reports.

Audit of data and code use in the SR-Can safety assessment

Energy Technology Data Exchange (ETDEWEB)

Hicks, T.W.; Baldwin, T.D. [Galson Sciences Ltd, 5 Grosvenor House, Melton R oad, Oakham, Rutland LE15 6AX (United Kingdom)

2008-03-15

Building on the findings of previous studies on data and code quality assurance (QA) in safety assessments, this report provides a review of data and code QA in the SR-Can safety assessment. The data quality audit aimed to check that the selection and use of data in the SR-Can safety assessment was appropriate, focusing on the data that underpin representations of and assumptions about canister, insert, buffer, and backfill behaviour. The SR-Can Data Report provided the initial focus for examining the traceability and reliability of data used in the safety assessment; the Data Report is one of the series of SR-Can safety assessment reports and, in this review, it was anticipated that it would provide the primary source of data on the canister, insert, buffer, and backfill. However, other safety assessment reports (the SR-Can Main Report, the Initial State Report, the Fuel and Canister Process Report, and the Buffer and Backfill Process Report) were found to provide key information on data used in the safety assessment. The quality audit of codes aimed to check that code use in the SR-Can safety assessment has been justified through a transparent and traceable process of code development and selection. The Model Summary Report provided the focus for reviewing the QA status of the codes used in the safety assessment. As well as highlighting a number of concerns regarding QA aspects of specific data sets, parameter values, and codes used in the SR-Can safety assessment (which are presented in the report), the review has led to several general observations on data and code QA that should be considered by SKB in the development and implementation of a QA system for the SR-Site safety assessment: - The SR-Site safety assessment and associated QA records should include information that demonstrates that a full QA system has been implemented in order to build confidence in the validity of the assessment. - The data and parameter values used directly in the safety

A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

International Nuclear Information System (INIS)

Ohi, Takao; Kawasaki, Daisuke; Chiba, Tamotsu; Takase, Toshio; Hane, Koji

2013-01-01

A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner. (author)

Procedures for self-assessment of operational safety

International Nuclear Information System (INIS)

1997-08-01

Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and and the safety culture as a whole. The concepts developed in this report present the basic approach to self-assessment taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject

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

Science.gov (United States)

Guo, Jingjing

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

Safety assessment of outdoor live fire range

Energy Technology Data Exchange (ETDEWEB)

NONE

1989-05-01

The following Safety Assessment (SA) pertains to the outdoor live fire range facility (LFR). The purpose of this facility is to supplement the indoor LFR. In particular it provides capacity for exercises that would be inappropriate on the indoor range. This SA examines the risks that are attendant to the training on the outdoor LFR. The outdoor LFR used by EG&G Mound is privately owned. It is identified as the Miami Valley Shooting Grounds. Mondays are leased for the exclusive use of EG&G Mound.

Calcium lignosulphonate: re-evaluation of relevant endpoints to re-confirm validity and NOAEL of a 90-day feeding study in rats.

Science.gov (United States)

Thiel, Anette; Braun, William; Cary, Maurice G; Engelhardt, Jeffery A; Goodman, Dawn G; Hall, William C; Romeike, Annette; Ward, Jerrold M

2013-08-01

A 90-day feeding study in Han/Wistar rats with calcium lignosulphonate was evaluated by the EFSA. The study was considered to be inadequate due to potentially impaired health status of the animals based upon a high incidence of minimal lymphoid hyperplasia in mesenteric/mandibular lymph nodes and Peyer's patches, and minimal lymphoid cell infiltration in the liver in all animals. The EFSA Panel further disagreed with the conclusion that the treatment-related observation of foamy histiocytosis in mesenteric lymph nodes was non-adverse and asked whether this observation would progress to something more adverse over time. A PWG was convened to assess the sections of lymph nodes, Peyer's patches and liver. In addition, all lymphoid tissues were re-examined. The clinical pathology and animal colony health screening data were re-evaluated. The question whether the foamy histiocytosis could progress to an adverse finding with increasing exposure duration was addressed by read-across. In conclusion, the animals on the 90-day feeding study were in good health, the study was adequate for safety evaluation, and the foamy histiocytes in the mesenteric lymph nodes were not considered adverse, but rather an adaptive response that was considered unlikely to progress to an adverse condition with time. The NOAEL was re-affirmed to be 2000 mg/kgbw/d. Copyright © 2013 Elsevier Inc. All rights reserved.

Safety assessment of allylalkoxybenzene derivatives used as flavouring substances - methyl eugenol and estragole

NARCIS (Netherlands)

Smith, R.L.; Adams, T.B.; Doull, J.; Feron, V.J.; Goodman, J.I.; Marnett, L.J.; Portoghese, P.S.; Waddell, W.J.; Wagner, B.M.; Rogers, A.E.; Caldwell, J.; Sipes, I.G.

2002-01-01

This publication is the seventh in a series of safety evaluations performed by the Expert Panel of the Flavor and Extract Manufacturers' Association (FEMA). In 1993, the Panel initiated a comprehensive program to re-evaluate the safety of more than 1700 GRAS flavouring substances under conditions of

On the re-characterisation of defects in structural integrity assessments due to snap-through

International Nuclear Information System (INIS)

Burdekin, F.M.; Sherry, A.H.

2006-01-01

TAGSI was asked by HSE/NII to comment on the effectiveness of the R6 procedure which permits a surface defect failing an initial assessment to be re-assessed as a through thickness defect of increased length. If the defect passes the re-assessment it may then be considered acceptable. In TAGSI's view re-characterisation needs to be considered separately for contained plastic collapse and ductile or brittle fracture of the remaining ligament. For plastic collapse and ductile tearing of the remaining ligament the R6 rules for re-characterisation appear generally satisfactory. For brittle failure of the ligament in many cases the R6 recommendations will be found satisfactory but for cases of large difference between static and dynamic toughness it is recommended that scoping calculations should be carried out for the defect geometry, loading and material properties of concern

Light water reactor sequence timing: its significance to probabilistic safety assessment modeling

International Nuclear Information System (INIS)

Bley, D.C.; Buttemer, D.R.; Stetkar, J.W.

1988-01-01

This paper examines event sequence timing in light water reactor plants from the viewpoint of probabilistic safety assessment (PSA). The analytical basis for the ideas presented here come primarily from the authors' work in support of more than 20 PSA studies over the past several years. Timing effects are important for establishing success criteria for support and safety system response and for identifying the time available for operator recovery actions. The principal results of this paper are as follows: 1. Analysis of event sequence timing is necessary for meaningful probabilistic safety assessment - both the success criteria for systems performance and the probability of recovery are tightly linked to sequence timing. 2. Simple engineering analyses based on first principles are often sufficient to provide adequate resolution of the time available for recovery of PSA scenarios. Only those parameters that influence sequence timing and its variability and uncertainty need be examined. 3. Time available for recovery is the basic criterion for evaluation of human performance, whether time is an explicit parameter of the operator actions analysis or not. (author)

Uncertainty analysis in safety assessment

International Nuclear Information System (INIS)

Lemos, Francisco Luiz de; Sullivan, Terry

1997-01-01

Nuclear waste disposal is a very complex subject which requires the study of many different fields of science, like hydro geology, meteorology, geochemistry, etc. In addition, the waste disposal facilities are designed to last for a very long period of time. Both of these conditions make safety assessment projections filled with uncertainty. This paper addresses approaches for treatment of uncertainties in the safety assessment modeling due to the variability of data and some current approaches used to deal with this problem. (author)

NUMO's approach for long-term safety assessment - 59404

International Nuclear Information System (INIS)

Ebashi, Takeshi; Kaku, Kenichi; Ishiguro, Katsuhiko

2012-01-01

One of NUMO's policies for ensuring safety is staged and flexible project implementation and decision-making based on iterative confirmation of safety. The safety assessment takes the central role in multiple lines of reasoning and argumentation by providing a quantitative evaluation of long-term safety; a key aspect is uncertainty management. This paper presents NUMO's basic strategies for long-term safety assessment based on the above policy. NUMO's approach considering Japanese boundary conditions is demonstrated as a starting-point for evaluating the long-term safety of an actual site. In Japan, the Act on Final Disposal of Specified Radioactive Waste states that the siting process shall consist of three stages. The Nuclear Waste Management Organization of Japan (NUMO) is responsible for geological disposal of vitrified high-level waste and some types of TRU waste. NUMO has chosen to implement a volunteer approach to siting. NUMO decided to prepare the so-called 2010 technical report, which sets out three safety policies, one of which is staged project implementation and decision-making based on iterative confirmation of safety. Based on this policy, NUMO will gradually integrate relevant interdisciplinary knowledge to build a safety case when a formal volunteer application is received that would allow site investigations to be initiated. The safety assessment takes the central role in multiple lines of reasoning and argumentation by providing a quantitative evaluation of long-term safety; one of a key aspect is uncertainty management. This paper presents the basic strategies for NUMO's long-term safety assessment based on the above policy. In concrete terms, the common procedures involved in safety assessment are applied in a stepwise manner, based on integration of knowledge obtained from site investigations/evaluations and engineered measures. The results of the safety assessment are then reflected in the planning of site investigations and engineered

Physical design correlates of efficiency and safety in emergency departments: a qualitative examination.

Science.gov (United States)

Pati, Debajyoti; Harvey, Thomas E; Pati, Sipra

2014-01-01

The objective of this study was to explore and identify physical design correlates of safety and efficiency in emergency department (ED) operations. This study adopted an exploratory, multimeasure approach to (1) examine the interactions between ED operations and physical design at 4 sites and (2) identify domains of physical design decision-making that potentially influence efficiency and safety. Multidisciplinary gaming and semistructured interviews were conducted with stakeholders at each site. Study data suggest that 16 domains of physical design decisions influence safety, efficiency, or both. These include (1) entrance and patient waiting, (2) traffic management, (3) subwaiting or internal waiting areas, (4) triage, (5) examination/treatment area configuration, (6) examination/treatment area centralization versus decentralization, (7) examination/treatment room standardization, (8) adequate space, (9) nurse work space, (10) physician work space, (11) adjacencies and access, (12) equipment room, (13) psych room, (14) staff de-stressing room, (15) hallway width, and (16) results waiting area. Safety and efficiency from a physical environment perspective in ED design are mutually reinforcing concepts--enhancing efficiency bears positive implications for safety. Furthermore, safety and security emerged as correlated concepts, with security issues bearing implications for safety, thereby suggesting important associations between safety, security, and efficiency.

Probabilistic safety assessment

International Nuclear Information System (INIS)

Hoertner, H.; Schuetz, B.

1982-09-01

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

IAEA Issues Report on Mission to Review Japan's Nuclear Power Plant Safety Assessment Process

International Nuclear Information System (INIS)

2012-01-01

Full text: A team of international nuclear safety experts has delivered its report on a mission it conducted from 21-31 January 2012 to review Japan's process for assessing nuclear safety at the nation's nuclear power plants. International Atomic Energy Agency (IAEA) officials delivered the IAEA Mission Report to Japanese officials yesterday and made it publicly available today. Following the 11 March 2011 accident at TEPCO's Fukushima Daiichi Nuclear Power Station, Japan's Nuclear and Industrial Safety Agency (NISA) announced the development of a revised safety assessment process for the nation's nuclear power reactors. At the request of the Government of Japan, the IAEA organized a team of five IAEA and three international nuclear safety experts and visited Japan to review NISA's approach to the Comprehensive Assessments for the Safety of Existing Power Reactor Facilities and how NISA examines the results submitted by nuclear operators. A Preliminary Summary Report was issued on 31 January. 'The mission report provides additional information regarding the team's recommendations and overall finding that NISA's instructions to power plants and its review process for the Comprehensive Safety Assessments are generally consistent with IAEA Safety Standards', said team leader James Lyons, Director of the IAEA's Nuclear Installation Safety Division. National safety assessments and their peer review by the IAEA are a key component of the IAEA Action Plan on Nuclear Safety, which was approved by the Agency's Member States following last year's nuclear accident at Fukushima Daiichi Nuclear Power Station. The IAEA safety review mission held meetings in Tokyo with officials from NISA, the Japanese Nuclear Energy Safety Organization (JNES), and the Kansai Electric Power Company (KEPCO), and the team visited the Ohi Nuclear Power Station to see an example of how Japan's Comprehensive Safety Assessment is being implemented by nuclear operators. In its report delivered today

Safety investigation of 'Mutsu', the first nuclear ship in Japan (the correspondence to the guideline of safety design examination, etc.)

International Nuclear Information System (INIS)

1981-01-01

Japan Nuclear Ship Development Agency had made previously application for the permission of the alteration of the reactor installation in the nuclear ship Mutsu (the first of this kind in Japan), based on the overall safety investigation of the ship made by JNDA. Taking the opportunity of the governmental safety examination concerning the permission, the correspondence of the safety aspects of the n.s. Mutsu to the existing guidelines for the safety of nuclear reactor facilities was examined. These results to further enhance the safety of the n.s. Mutsu are described concerning the following matters: the safety design examination guideline for power-generating LWR facilities (58 items); the safety evaluation guideline for power-generating LWR facilities (the analysis of abnormal transients during operation and accidents); the safety countermeasures to be adopted in the reactor plant of the n.s. Mutsu from the situation of the TMI nuclear accident in U.S. (7 in design and 10 in operation management); the analysis simulating the TMI accident. (J.P.N.)

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

Application of petrographic examination techniques to the assessment of fire-damaged concrete and masonry structures

International Nuclear Information System (INIS)

Ingham, Jeremy P.

2009-01-01

The number of building fires has doubled over the last 50 years. There has never been a greater need for structures to be assessed for fire damage to ensure safety and enable appropriate repairs to be planned. Fortunately, even after a severe fire, concrete and masonry structures are generally capable of being repaired rather than demolished. By allowing direct examination of microcracking and mineralogical changes, petrographic examination has become widely used to determine the depth of fire damage for reinforced concrete elements. Petrographic examination can also be applied to fire-damaged masonry structures built of materials such as stone, brick and mortar. Petrography can ensure accurate detection of damaged geomaterials, which provides cost savings during building repair and increased safety reassurance. This paper comprises a review of the role of petrography in fire damage assessments, drawing on a range of actual fire damage investigations.

Complementary safety assessments - Report by the French Nuclear Safety Authority

International Nuclear Information System (INIS)

2011-12-01

As an immediate consequence of the Fukushima accident, the French Authority of Nuclear Safety (ASN) launched a campaign of on-site inspections and asked operators (mainly EDF, AREVA and CEA) to make complementary assessments of the safety of the nuclear facilities they manage. The approach defined by ASN for the complementary safety assessments (CSA) is to study the behaviour of nuclear facilities in severe accidents situations caused by an off-site natural hazard according to accident scenarios exceeding the current baseline safety requirements. This approach can be broken into 2 phases: first conformity to current design and secondly an approach to the beyond design-basis scenarios built around the principle of defence in depth. 38 inspections were performed on issues linked to the causes of the Fukushima crisis. It appears that some sites have to reinforce the robustness of the heat sink. The CSA confirmed that the processes put into place at EDF to detect non-conformities were satisfactory. The complementary safety assessments demonstrated that the current seismic margins on the EDF nuclear reactors are satisfactory. With regard to flooding, the complementary safety assessments show that the complete reassessment carried out following the flooding of the Le Blayais nuclear power plant in 1999 offers the installations a high level of protection against the risk of flooding. Concerning the loss of electrical power supplies and the loss of cooling systems, the analysis of EDF's CSA reports showed that certain heat sink and electrical power supply loss scenarios can, if nothing is done, lead to core melt in just a few hours in the most unfavourable circumstances. As for nuclear facilities that are not power or experimental reactors, some difficulties have appeared to implement the CSA approach that was initially devised for reactors. Generally speaking, ASN considers that the safety of nuclear facilities must be made more robust to improbable risks which are not

Assessing the Content of YouTube Videos in Educating Patients Regarding Common Imaging Examinations.

Science.gov (United States)

Rosenkrantz, Andrew B; Won, Eugene; Doshi, Ankur M

2016-12-01

To assess the content of currently available YouTube videos seeking to educate patients regarding commonly performed imaging examinations. After initial testing of possible search terms, the first two pages of YouTube search results for "CT scan," "MRI," "ultrasound patient," "PET scan," and "mammogram" were reviewed to identify educational patient videos created by health organizations. Sixty-three included videos were viewed and assessed for a range of features. Average views per video were highest for MRI (293,362) and mammography (151,664). Twenty-seven percent of videos used a nontraditional format (eg, animation, song, humor). All videos (100.0%) depicted a patient undergoing the examination, 84.1% a technologist, and 20.6% a radiologist; 69.8% mentioned examination lengths, 65.1% potential pain/discomfort, 41.3% potential radiation, 36.5% a radiology report/results, 27.0% the radiologist's role in interpretation, and 13.3% laboratory work. For CT, 68.8% mentioned intravenous contrast and 37.5% mentioned contrast safety. For MRI, 93.8% mentioned claustrophobia, 87.5% noise, 75.0% need to sit still, 68.8% metal safety, 50.0% intravenous contrast, and 0.0% contrast safety. For ultrasound, 85.7% mentioned use of gel. For PET, 92.3% mentioned radiotracer injection, 61.5% fasting, and 46.2% diabetic precautions. For mammography, unrobing, avoiding deodorant, and possible additional images were all mentioned by 63.6%; dense breasts were mentioned by 0.0%. Educational patient videos on YouTube regarding common imaging examinations received high public interest and may provide a valuable patient resource. Videos most consistently provided information detailing the examination experience and less consistently provided safety information or described the presence and role of the radiologist. Copyright © 2016 American College of Radiology. Published by Elsevier Inc. All rights reserved.

Development and applications of a safety assessment system for promoting safety culture in nuclear power plants

International Nuclear Information System (INIS)

Takano, Ken-ichi; Hasegawa, Naoko; Hirose, Ayako; Hayase, Ken-ichi

2004-01-01

For past five years, CRIEPI has been continuing efforts to develop and make applications of a 'safety assessment system' which enable to measure the safety level of organization. This report describe about frame of the system, assessment results and its reliability, and relation between labor accident rate in the site and total safety index (TSI), which can be obtained by the principal factors analysis. The safety assessment in this report is based on questionnaire survey of employee. The format and concrete questionnaires were developed using existing literatures including organizational assessment tools. The tailored questionnaire format involved 124 questionnaire items. The assessment results could be considered as a well indicator of the safety level of organization, safety management, and safety awareness of employee. (author)

Safety culture' is integrating 'human' into risk assessment

International Nuclear Information System (INIS)

Sugimoto, Taiji

2014-01-01

Significance of Fukushima nuclear power accident requested reconsideration of safety standards, of which we had usually no doubt. Risk assessment standard (JIS B 9702), Which was used for repetition of database preparation and cumulative assessment, defined allowable risk and residual risk. However, work site and immediate assessment was indispensable beside such assessment so as to ensure safety. Risk of casualties was absolutely not acceptable in principle and judgments to approve allowable risk needed accountability, which was reminded by safety culture proposed by IAEA and also identified by investigation of organizational cause of Columbia accident. Actor of safety culture would be organization and individual, and mainly individual. Realization of safety culture was conducted by personnel having moral consciousness and firm sense of mission in the course of jobs and working daily with sweat pouring. Safety engineering/technology should have framework integrating human as such totality. (T. Tanaka)

Re criticality assessment following reactor core damage in Fukushima unit 2

International Nuclear Information System (INIS)

Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

2012-01-01

Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

Assessment of molecular events during in vitro re-epithelialization under honey-alginate matrix ambience

Energy Technology Data Exchange (ETDEWEB)

Barui, Ananya [Centre for Healthcare Science and Technology, BESU, Shibpur, Howrah 711103, West Bengal (India); Mandal, Naresh [Dept. of Electronics and Telecommunication Engg., BESU, Shibpur, Howrah 711103, West Bengal (India); Majumder, Subhadipa [Department of Biochemistry, University of Calcutta Ballygunge, Circular Road, Kolkata 700 019, West Bengal (India); Das, Raunak Kumar [School of Medical Science and Technology, IIT, Kharagpur 721 302, West Bengal (India); Sengupta, Sanghamitra [Department of Biochemistry, University of Calcutta Ballygunge, Circular Road, Kolkata 700 019, West Bengal (India); Banerjee, Provas [School of Medical Science and Technology, IIT, Kharagpur 721 302, West Bengal (India); Ray, Ajoy Kumar; RoyChaudhuri, Chirosree [Dept. of Electronics and Telecommunication Engg., BESU, Shibpur, Howrah 711103, West Bengal (India); Chatterjee, Jyotirmoy, E-mail: jchatterjee@smst.iitkgp.ernet.in [School of Medical Science and Technology, IIT, Kharagpur 721 302, West Bengal (India)

2013-08-01

Re-epithelialization is one of the most important stages of cutaneous regeneration and its success requires supportive micro-ambience which may be provided with suitable bio-matrix. Biocompatibility and efficacy of such bio-matrix in re-epithelialization could be explored by multimodal analysis of structural and functional attributes of in vitro wound healing model including evaluation of prime molecular expressions of the epithelial cells during repair. Present study examines the influence of honey-alginate and alginate matrices on re-epithelialization in keratinocyte (HaCaT) population in a 2-D wound model. Cellular viability, proliferation and cell–cell adhesion status were assessed during wound closure using live/dead cell assay and by evaluating expressions of Ki67, p63 and E-cadherin along-with % change in cellular electrical impedance. Efficacy of honey-alginate matrix in comparison to only alginate one was demonstrated by a quicker reduction in wound gap, improved cellular viability, enhanced expressions of Ki67, p63 and its isoforms (TAp63, ΔNp63) as well as E-cadherin. Faster restoration of electrical attribute (% of impedance change) after wounding also indicated better impact of honey-alginate matrix in re-epithelialization. - Highlights: • Role of honey based matrix is evaluated in wound re-epithelialization. • Healing impact of matrix studied in 2D in vitro keratinocyte (HaCaT) wound model. • Faster impedance restoration indicated rapid healing rate of HaCaT under honey. • PCR observations showed faster initiation of cell proliferation under honey. • ICC study indicated better up-regulation of healing markers under honey matrix.

Assessment of molecular events during in vitro re-epithelialization under honey-alginate matrix ambience

International Nuclear Information System (INIS)

Barui, Ananya; Mandal, Naresh; Majumder, Subhadipa; Das, Raunak Kumar; Sengupta, Sanghamitra; Banerjee, Provas; Ray, Ajoy Kumar; RoyChaudhuri, Chirosree; Chatterjee, Jyotirmoy

2013-01-01

Re-epithelialization is one of the most important stages of cutaneous regeneration and its success requires supportive micro-ambience which may be provided with suitable bio-matrix. Biocompatibility and efficacy of such bio-matrix in re-epithelialization could be explored by multimodal analysis of structural and functional attributes of in vitro wound healing model including evaluation of prime molecular expressions of the epithelial cells during repair. Present study examines the influence of honey-alginate and alginate matrices on re-epithelialization in keratinocyte (HaCaT) population in a 2-D wound model. Cellular viability, proliferation and cell–cell adhesion status were assessed during wound closure using live/dead cell assay and by evaluating expressions of Ki67, p63 and E-cadherin along-with % change in cellular electrical impedance. Efficacy of honey-alginate matrix in comparison to only alginate one was demonstrated by a quicker reduction in wound gap, improved cellular viability, enhanced expressions of Ki67, p63 and its isoforms (TAp63, ΔNp63) as well as E-cadherin. Faster restoration of electrical attribute (% of impedance change) after wounding also indicated better impact of honey-alginate matrix in re-epithelialization. - Highlights: • Role of honey based matrix is evaluated in wound re-epithelialization. • Healing impact of matrix studied in 2D in vitro keratinocyte (HaCaT) wound model. • Faster impedance restoration indicated rapid healing rate of HaCaT under honey. • PCR observations showed faster initiation of cell proliferation under honey. • ICC study indicated better up-regulation of healing markers under honey matrix

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

Science.gov (United States)

Ancel, Ersin; Shih, Ann T.

2014-01-01

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

Uncertainty analysis in safety assessment

Energy Technology Data Exchange (ETDEWEB)

Lemos, Francisco Luiz de [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Sullivan, Terry [Brookhaven National Lab., Upton, NY (United States)

1997-12-31

Nuclear waste disposal is a very complex subject which requires the study of many different fields of science, like hydro geology, meteorology, geochemistry, etc. In addition, the waste disposal facilities are designed to last for a very long period of time. Both of these conditions make safety assessment projections filled with uncertainty. This paper addresses approaches for treatment of uncertainties in the safety assessment modeling due to the variability of data and some current approaches used to deal with this problem. (author) 13 refs.; e-mail: lemos at bnl.gov; sulliva1 at bnl.gov

A re-determination and re-assessment of the thermodynamics of sublimation of uranium dioxide

International Nuclear Information System (INIS)

Ackermann, R.J.; Rauh, E.G.; Rand, M.H.

1980-01-01

New mass-spectrometric measurements on the ion-intensity of UO 2 + over urania from 1813 to 2463 K are reported. Although the mean value for the enthalpy of sublimation calculated from these measurements is close to previous values, a detailed examination of the results indicates that there is an appreciable curvature in the log p versus reciprocal-temperature curve for the process: UO 2 (s)→UO 2 (g). This is attributed to a large negative ΔCsub(p) for the sublimation reaction, arising from the sharp increase in Csub(p) (UO 2 (s)) above approximately 1750 K. A thorough re-assessment of the previous studies on the sublimation of urania suggests an 'international' average value of psub(UO 2 )=(1.3+-0.1)x10 -6 atm at 2150 K; Knudsen effusion measurements above 2450 K (p>1x10 -4 atm) are thought to be in error due to departures from molecular flow. Thermal functions for UO 2 (g) have been calculated, assuming a linear molecule and electronic contributions to the partition function based on those of ThO(g). Anharmonicity corrections have been included. When these functions are combined with the thermal functions for UO 2 (s), recently assessed, the third law heat of sublimation at 298.15 K becomes 147.8 kcal.mol -1 with a trend of only 0.2 kcal.mol -1 across the temperature range 1800 to 2400 K. (author)

Safety assessment of foods derived from genetically modified crops

NARCIS (Netherlands)

Kleter, G.A.; Kuiper, H.A.

2003-01-01

The pre-market safety assessment of foods derived from genetically modified crops is carried out according to the consensus approach of "substantial equivalence", in other words: the comparative safety assessment. Currently, the safety assessment of genetically modified foods is harmonized at the

National Waste Repository Novi Han operational safety analysis report. Safety assessment methodology

International Nuclear Information System (INIS)

2003-01-01

The scope of the safety assessment (SA), presented includes: waste management functions (acceptance, conditioning, storage, disposal), inventory (current and expected in the future), hazards (radiological and non-radiological) and normal and accidental modes. The stages in the development of the SA are: criteria selection, information collection, safety analysis and safety assessment documentation. After the review the facilities functions and the national and international requirements, the criteria for safety level assessment are set. As a result from the 2nd stage actual parameters of the facility, necessary for safety analysis are obtained.The methodology is selected on the base of the comparability of the results with the results of previous safety assessments and existing standards and requirements. The procedure and requirements for scenarios selection are described. A radiological hazard categorisation of the facilities is presented. Qualitative hazards and operability analysis is applied. The resulting list of events are subjected to procedure for prioritization by method of 'criticality analysis', so the estimation of the risk is given for each event. The events that fall into category of risk on the boundary of acceptability or are unacceptable are subjected to the next steps of the analysis. As a result the lists with scenarios for PSA and possible design scenarios are established. PSA logical modeling and quantitative calculations of accident sequences are presented

Supplier-induced demand: re-examining identification and misspecification in cross-sectional analysis.

Science.gov (United States)

Peacock, Stuart J; Richardson, Jeffrey R J

2007-09-01

This paper re-examines criticisms of cross-sectional methods used to test for supplier-induced demand (SID) and re-evaluates the empirical evidence using data from Australian medical services. Cross-sectional studies of SID have been criticised on two grounds. First, and most important, the inclusion of the doctor supply in the demand equation leads to an identification problem. This criticism is shown to be invalid, as the doctor supply variable is stochastic and depends upon a variety of other variables including the desirability of the location. Second, cross-sectional studies of SID fail diagnostic tests and produce artefactual findings due to model misspecification. Contrary to this, the re-evaluation of cross-sectional Australian data indicate that demand equations that do not include the doctor supply are misspecified. Empirical evidence from the re-evaluation of Australian medical services data supports the notion of SID. Demand and supply equations are well specified and have very good explanatory power. The demand equation is identified and the desirability of a location is an important predictor of the doctor supply. Results show an average price elasticity of demand of 0.22 and an average elasticity of demand with respect to the doctor supply of 0.46, with the impact of SID becoming stronger as the doctor supply rises. The conclusion we draw from this paper is that two of the main criticisms of the empirical evidence supporting the SID hypothesis have been inappropriately levelled at the methods used. More importantly, SID provides a satisfactory, and robust, explanation of the empirical data on the demand for medical services in Australia.

Assessment of safety culture: Changing regulatory approach in Hungary

International Nuclear Information System (INIS)

Ronaky, Jozsef; Toth, Andras

2002-01-01

Hungarian Atomic Energy Authority (HAEA) is changing its inspection practice and assessment methods of safety performance and safety culture in operating nuclear facilities. The new approach emphasises integrated team inspection of safety cornerstones and systematic assessment of safety performance of operators. (author)

Tolerability of risk, safety assessment principles and their implications for probabilistic safety analysis

International Nuclear Information System (INIS)

Ewing, D.J.F.; Campbell, J.F.

1994-01-01

This paper gives a regulatory view of probabilistic safety assessment as seen by the Nuclear Installations Inspectorate (NII) and in the light of the general regulatory risk aims set out in the Health and Safety Executive's (HSE) The tolerability of risk from nuclear power stations (TOR) and in Safety assessment principles for nuclear plants (SAPs), prepared by NII on behalf of the HSE. Both of these publications were revised and republished in 1992. This paper describes the SAPs, together with the historical background, the motivation for review, the effects of the Sizewell and Hinkley Point C public inquiries, changes since the original versions, comparison with international standards and use in assessment. For new plant, probabilistic safety analysis (PSA) is seen as an essential tool in balancing the safety of the design and in demonstrating compliance with TOR and the SAPs. (Author)

The radiation safety self-assessment program of Ontario Hydro

International Nuclear Information System (INIS)

Armitage, G.; Chase, W.J.

1987-01-01

Ontario Hydro has developed a self-assessment program to ensure that high quality in its radiation safety program is maintained. The self-assessment program has three major components: routine ongoing assessment, accident/incident investigation, and detailed assessments of particular radiation safety subsystems or of the total radiation safety program. The operation of each of these components is described

Re-evaluation of azo dyes as food additives

DEFF Research Database (Denmark)

Pratt, Iona; Larsen, John Christian; Mortensen, Alicja

2013-01-01

additives to be assessed by the Scientific Committee on Food, many years ago, (ii) because of concern regarding possible health effects of artificial colours arising since the original evaluations.Concerns includedbehavioural effects in children, allergic reactions, genotoxicity and possible carcinogenicity......Aryl azo compounds are widely used as colorants (azo dyes) in a wide range of products including textiles, leather, paper, cosmetics, pharmaceuticals and food.As part of its systematic re-evaluation of food additives, the European Food Safety Authority (EFSA) has carried out new risk assessments...

Probabilistic safety assessment for seismic events

International Nuclear Information System (INIS)

1993-10-01

This Technical Document on Probabilistic Safety Assessment for Seismic Events is mainly associated with the Safety Practice on Treatment of External Hazards in PSA and discusses in detail one specific external hazard, i.e. earthquakes

Safety analysis and risk assessment handbook

International Nuclear Information System (INIS)

Peterson, V.L.; Colwell, R.G.; Dickey, R.L.

1997-01-01

This Safety Analysis and Risk Assessment Handbook (SARAH) provides guidance to the safety analyst at the Rocky Flats Environmental Technology Site (RFETS) in the preparation of safety analyses and risk assessments. Although the older guidance (the Rocky Flats Risk Assessment Guide) continues to be used for updating the Final Safety Analysis Reports developed in the mid-1980s, this new guidance is used with all new authorization basis documents. With the mission change at RFETS came the need to establish new authorization basis documents for its facilities, whose functions had changed. The methodology and databases for performing the evaluations that support the new authorization basis documents had to be standardized, to avoid the use of different approaches and/or databases for similar accidents in different facilities. This handbook presents this new standardized approach. The handbook begins with a discussion of the requirements of the different types of authorization basis documents and how to choose the one appropriate for the facility to be evaluated. It then walks the analyst through the process of identifying all the potential hazards in the facility, classifying them, and choosing the ones that need to be analyzed further. It then discusses the methods for evaluating accident initiation and progression and covers the basic steps in a safety analysis, including consequence and frequency binning and risk ranking. The handbook lays out standardized approaches for determining the source terms of the various accidents (including airborne release fractions, leakpath factors, etc.), the atmospheric dispersion factors appropriate for Rocky Flats, and the methods for radiological and chemical consequence assessments. The radiological assessments use a radiological open-quotes templateclose quotes, a spreadsheet that incorporates the standard values of parameters, whereas the chemical assessments use the standard codes ARCHIE and ALOHA

Living probabilistic safety assessment (LPSA)

International Nuclear Information System (INIS)

1999-08-01

Over the past few years many nuclear power plant organizations have performed probabilistic safety assessments (PSAs) to identify and understand key plant vulnerabilities. As a result of the availability of these PSA studies, there is a desire to use them to enhance plant safety and to operate the nuclear stations in the most efficient manner. PSA is an effective tool for this purpose as it assists plant management to target resources where the largest benefit to plant safety can be obtained. However, any PSA which is to be used in this way must have a credible and defensible basis. Thus, it is very important to have a high quality 'living PSA' accepted by the plant and the regulator. With this background in mind, the IAEA has prepared this report on Living Probabilistic Safety Assessment (LPSA) which addresses the updating, documentation, quality assurance, and management and organizational requirements for LPSA. Deficiencies in the areas addressed in this report would seriously reduce the adequacy of the LPSA as a tool to support decision making at NPPs. This report was reviewed by a working group during a Technical Committee Meeting on PSA Applications to Improve NPP Safety held in Madrid, Spain, from 23 to 27 February 1998

Rights, responsibilities and (re)presentation: Using drawings to convey health and safety messages among immigrant workers.

Science.gov (United States)

McKillop, Chris; Parsons, Janet A; Brown, Janet; Scott, Susan; Holness, D Linn

2016-09-27

Immigrant workers who are new to Canada are considered a vulnerable population under the Ontario Ministry of Labour Prevention Strategy for workplace safety. Posters outlining workplace safety rights and responsibilities may not be understandable to new immigrants. To explore visual approaches to making health and safety messages more understandable to new immigrants. This pilot study used arts-based qualitative research methods. Key messages from the Ministry of Labour Health & Safety at Work poster were (re)represented as images by an artist. Recent immigrants engaged in individual interviews and then took part in a focus group, in order to elicit their experiences of health and safety practices, their understanding and feedback concerning the Ministry poster, and the images created. An image-rich version of the poster was developed. The combination of drawings and minimal text was preferred and considered helpful by participants. Barriers to health and safety and work challenges for new immigrants were highlighted. Visual analysis yielded new versions of the poster, as well as a pictorial representation of the research process and study findings. The study demonstrates the value of using image-rich posters with immigrant workers, and the effectiveness of using arts-based methods within the research process.

Safety re-evaluation of the HOR reactor

International Nuclear Information System (INIS)

Verkooijen, A.H.M.; Vries, J.W. de

2001-01-01

State. Requirement C16 in the new licence asks for a periodical integral safety re-evaluation of the HOR reactor every 10 years and starting after 2 years

Probabilistic safety assessment in nuclear power plant management

International Nuclear Information System (INIS)

Holloway, N.J.

1989-06-01

Probabilistic Safety Assessment (PSA) techniques have been widely used over the past few years to assist in understanding how engineered systems respond to abnormal conditions, particularly during a severe accident. The use of PSAs in the design and operation of such systems thus contributes to the safety of nuclear power plants. Probabilistic safety assessments can be maintained to provide a continuous up-to-date assessment (Living PSA), supporting the management of plant operations and modifications

Assessing medical students' perceptions of patient safety: the medical student safety attitudes and professionalism survey.

Science.gov (United States)

Liao, Joshua M; Etchegaray, Jason M; Williams, S Tyler; Berger, David H; Bell, Sigall K; Thomas, Eric J

2014-02-01

To develop and test the psychometric properties of a survey to measure students' perceptions about patient safety as observed on clinical rotations. In 2012, the authors surveyed 367 graduating fourth-year medical students at three U.S. MD-granting medical schools. They assessed the survey's reliability and construct and concurrent validity. They examined correlations between students' perceptions of organizational cultural factors, organizational patient safety measures, and students' intended safety behaviors. They also calculated percent positive scores for cultural factors. Two hundred twenty-eight students (62%) responded. Analyses identified five cultural factors (teamwork culture, safety culture, error disclosure culture, experiences with professionalism, and comfort expressing professional concerns) that had construct validity, concurrent validity, and good reliability (Cronbach alphas > 0.70). Across schools, percent positive scores for safety culture ranged from 28% (95% confidence interval [CI], 13%-43%) to 64% (30%-98%), while those for teamwork culture ranged from 47% (32%-62%) to 74% (66%-81%). They were low for error disclosure culture (range: 10% [0%-20%] to 27% [20%-35%]), experiences with professionalism (range: 7% [0%-15%] to 23% [16%-30%]), and comfort expressing professional concerns (range: 17% [5%-29%] to 38% [8%-69%]). Each cultural factor correlated positively with perceptions of overall patient safety as observed in clinical rotations (r = 0.37-0.69, P safety behavioral intent item. This study provided initial evidence for the survey's reliability and validity and illustrated its applicability for determining whether students' clinical experiences exemplify positive patient safety environments.

The Fukushima Daiichi Accident. Technical Volume 2/5. Safety Assessment

International Nuclear Information System (INIS)

2015-08-01

safety systems. Section 2.4 describes the accident management provisions and their implementation. All components of accident management are discussed, both preventive (before core melt) and mitigative (after core melt or severe accident). The section covers hardware provisions, emergency operating procedures, severe accident operating procedures, human resources and organizational arrangements, including training and drills. Interface with the off-site emergency arrangements is also discussed. Section 2.5 deals with the governmental, legal and regulatory framework for nuclear safety in Japan up to the time of the Fukushima Daiichi accident. It evaluates this framework and its contribution to the accident, and identifies lessons learned. Section 2.6 analyses the human and organizational aspects of the accident. It examines the main stakeholders of nuclear safety in Japan and shows how their actions were interrelated and interconnected, thereby reinforcing basic assumptions about nuclear safety that prevented them from adequately preparing for such an accident. The section analyses why the accident happened despite advancements in nuclear safety in areas such as solid design, peer reviews, regulatory frameworks, safety assessment methodologies, years of successful operating experience, defence in depth, emergency preparedness, severe accident management guidelines (SAMGs) and a strong international commitment to nuclear safety. Finally, Section 2.7 addresses the role of operating experience in improving plant design and operation in order to continuously improve nuclear safety and support defence in depth. The section assesses the TEPCO operating experience programme and the extent to which lessons were learned from events both in Japan and internationally, and the design changes made

Types of safety assessments of near surface repository for radioactive waste

International Nuclear Information System (INIS)

Mateeva, M.

2004-01-01

The purpose of this article is to presents the classification of different types safety assessments of near surface repository for low and intermediate level radioactive waste substantiated with results of safety assessments generated in Bulgaria. The different approach of safety assessments applied for old existing repository as well as for site selection for construction new repository is outlined. The regulatory requirements in Bulgaria define three main types of assessments: Safety assessment; Technical substation of repository safety; Assessment of repository influence on environment that is in form of report prepared from the Ministry of environment and waters on the base of results obtained in two first types of assessments. Additionally first type is subdivided in three categories - preliminary safety assessment, safety assessment and post closure safety assessment, which are generated using deterministic approach. The technical substation of repository safety is generated using probabilistic approach. Safety assessment results that are presented here are based on evaluation of existing old repository type 'Radon' in Novi Han and real site selection procedure for new near surface repository for low and intermediate level radioactive waste from nuclear power station in Kozloduy. The important role of safety assessment for improvement the repository safety as well as for repository licensing, correct site selection and right choice of engineer barriers and repository design is discussed using generated results. (author)

Examination on the safety of handling the fuel elements in the nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

1977-01-01

This is the report of the Examination Committee on Total Inspection and Repair Technologies for Mutsu to the Director of Science and Technology Agency and the Minister of Transport dated July 29, 1977. The committee concluded before that the total inspection on safety and the repair of shielding can be carried out as the fuel elements are loaded, and the safety can be secured sufficiently. It was decided at the meeting of ministers concerned with Mutsu on May 17 that the safety concerning handling the fuel elements of Mutsu should be examined by the committee. Under the premise that the fuel elements are loaded again and used after the total inspection on safety and the repair of shielding, the committee examined the methods and the basic concept of safety about the taking-out, transport and preservation of the fuel elements, and the conclusions obtained are reported. The contents of the examination are the outline of the fuel elements, the present condition of the fuel elements, the safety concerning taking-out, transport and preservation of the fuel elements, and the other measures required for securing safety. The committee thinks that the safety can be secured sufficiently if the works are carried out carefully. (Kako, I.)

Data used for safety assessment of reprocessing facilities

International Nuclear Information System (INIS)

Nomura, Yasushi; Suzuki, Atsuyuki; Kanagawa, Akira

1990-08-01

For safety assessment of a reprocessing facility, it is important to know performance of radioactive materials in their accidental release and transfer. Accordingly, it is necessary to collect and prepare data for use in analyses for their performance. In JAERI, experiments such as for data acquisition, for source-term evaluation and for radioactive material transfer, are now planned to be performed. Prior to these experiments, it is decided to investigate data in use for accidental safety assessment of reprocessing plants and their based experimental data, thus to make it possible to recommend reasonable values for safety analysis parameters by evaluating the investigated results, to select the experimental items, to edit a safety assessment handbook and so on. In this line of objectives, JAERI rewarded a two-year contract of investigation to Nuclear Safety Research Association, to make a working group under a special committee on data investigation for reprocessing facility safety assessment. This report is a collection of results reviewed and checked by the working group. The contents consist of two parts, one for investigation and review of data used for safety assessment of domestic or oversea reprocessing facilities, and the other for investigation, review and evaluation of ANSI recommended American standard data reported by E. Walker together with their based experimental data resorting to the original referred reports. (author)

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

International Nuclear Information System (INIS)

Hedin, A.

2008-01-01

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

The role of IAEA in the seismic assessment and upgrading of existing NPPs. Seismic safety of nuclear power plants in Eastern Europe

Energy Technology Data Exchange (ETDEWEB)

Guerpinar, A; Godoy, A [International Atomic Energy Agency, Vienna (IAEA). Div. of Nuclear Installation Safety

1997-03-01

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on `Benchmark study for the seismic analysis and testing of WWER type nuclear power plants`. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

The role of IAEA in the seismic assessment and upgrading of existing NPPs. Seismic safety of nuclear power plants in Eastern Europe

International Nuclear Information System (INIS)

Guerpinar, A.; Godoy, A.; . Div. of Nuclear Installation Safety)

1997-01-01

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

Safety and reliability assessment

International Nuclear Information System (INIS)

1979-01-01

This report contains the papers delivered at the course on safety and reliability assessment held at the CSIR Conference Centre, Scientia, Pretoria. The following topics were discussed: safety standards; licensing; biological effects of radiation; what is a PWR; safety principles in the design of a nuclear reactor; radio-release analysis; quality assurance; the staffing, organisation and training for a nuclear power plant project; event trees, fault trees and probability; Automatic Protective Systems; sources of failure-rate data; interpretation of failure data; synthesis and reliability; quantification of human error in man-machine systems; dispersion of noxious substances through the atmosphere; criticality aspects of enrichment and recovery plants; and risk and hazard analysis. Extensive examples are given as well as case studies

Safety assessment as basis for the decision making process

International Nuclear Information System (INIS)

Ilie, P.; Didita, L.; Danchiv, A.

2005-01-01

This paper deals with the safety assessment for a new near surface repository, particularly for the early stage of repository development using ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) safety assessment methodology. In this stage of the repository life cycle the main purpose of the safety assessment is to demonstrate that the plant is capable to be constructed and operated safely. The paper is based on development of the ASAM (Application of the Safety Assessment Methodologies for Near-Surface Disposal Facilities) Decision Support Subgroup of the Common Aspects Working Group. The implications of decision making for the application of the ISAM methodology on post-closure safety assessment are analysed. Some important elements of the decision-making process with impact on key components of the ISAM process are described. Following the development of Decision Support Subgroup of the ASAM Common Aspects Working Group the proposed change of ISAM methodology is analysed. This approach puts all activities in a decision context where the first iteration of the safety assessment is based on the existing state of knowledge and the initial engineering design. Confidence in the process is accomplished through the direct inclusion of all decision makers and stakeholders in the formulation of decisions, the definition of the state of knowledge, and decision making activities. The decision process is developed in context of undertaking assessments with little site-specific information, this situation is specifically for new planned repository. Limited site-specific information can result in a high degree of uncertainty, therefore it is important first of all to identify the sources of uncertainty arising from the limited nature of the site-specific information and then to apply appropriate approaches to manage the uncertainties and to determine whether the uncertainties are important to the overall safety of the disposal facility

Assessment of safety culture at INPP

International Nuclear Information System (INIS)

Lesin, S.

2002-01-01

Safety Culture covers all main directions of plant activities and the plant departments involved through integration into the INPP Quality Assurance System. Safety Culture is represented by three components. The first is the clear INPP Safety and Quality Assurance Policy. Based on the Policy INPP is safely operated and managers' actions firstly aim at safety assurance. The second component is based on personal responsibility for safety and attitude of each employee of the plant. The third component is based on commitment to safety and competence of managers and employees of the plant. This component links the first two to ensure efficient management of safety at the plant. The above mentioned components including the elements which may significantly affect Safety Culture are also presented in the attachment. The concept of such model implies understanding of effect of different factors on the level of Safety Culture in the organization. In order to continuously correct safety problems, self-assessment of the Safety Culture level is performed at regular intervals. (author)

Safety assessment for radioactive waste disposal facility

International Nuclear Information System (INIS)

Thanaletchumy Karuppiah; Mohd Abdul Wahab Yusof; Nik Marzuki Nik Ibrahim; Nurul Wahida Ahmad Khairuddin

2008-08-01

Safety assessments are used to evaluate the performance of a radioactive waste disposal facility and its impact on human health and the environment. This paper presents the overall information and methodology to carry out the safety assessment for a long term performance of a disposal system. A case study was also conducted to gain hands-on experience in the development and justification of scenarios, the formulation and implementation of models and the analysis of results. AMBER code using compartmental modeling approach was used to represent the migration and fate of contaminants in this training. This safety assessment is purely illustrative and it serves as a starting point for each development stage of a disposal facility. This assessment ultimately becomes more detail and specific as the facility evolves. (Author)

Examination of the relationship between management and clinician perception of patient safety climate and patient satisfaction.

Science.gov (United States)

Mazurenko, Olena; Richter, Jason; Kazley, Abby Swanson; Ford, Eric

2017-04-25

The aim of this study was to explore the relationship between managers and clinicians' agreement on deeming the patient safety climate as high or low and the patients' satisfaction with those organizations. We used two secondary data sets: the Hospital Survey on Patient Safety Culture (2012) and the Hospital Consumer Assessment of Healthcare Providers and Systems (2012). We used ordinary least squares regressions to analyze the relationship between the extent of agreement between managers and clinicians' perceptions of safety climate in relationship to patient satisfaction. The dependent variables were four Hospital Consumer Assessment of Healthcare Providers and Systems patient satisfaction scores: communication with nurses, communication with doctors, communication about medicines, and discharge information. The main independent variables were four groups that were formed based on the extent of managers and clinicians' agreement on four patient safety climate domains: communication openness, feedback and communication about errors, teamwork within units, and teamwork across units. After controlling for hospital and market-level characteristics, we found that patient satisfaction was significantly higher if managers and clinicians reported that patient safety climate is high or if only clinicians perceived the climate as high. Specifically, manager and clinician agreement on high levels of communication openness (β = 2.25, p = .01; β = 2.46, p = .05), feedback and communication about errors (β = 3.0, p = .001; β = 2.89, p = .01), and teamwork across units (β = 2.91, p = .001; β = 3.34, p = .01) was positively and significantly associated with patient satisfaction with discharge information and communication about medication. In addition, more favorable perceptions about patient safety climate by clinicians only yielded similar findings. Organizations should measure and examine patient safety climate from multiple perspectives and be aware that individuals

Preliminary safety assessment of the WIPP facility

International Nuclear Information System (INIS)

Balestri, R.J.; Torres, B.W.; Pahwa, S.B.; Brannen, J.P.

1979-01-01

This paper summarizes the efforts to perform a safety assessment of the Waste Isolation Pilot Plant (WIPP) facility being proposed for southeastern New Mexico. This preliminary safety assessment is limited to a consequence assessment in terms of the dose to a maximally exposed individual as a result of introducing the radionuclides into the biosphere. The extremely low doses to the organs as a result of the liquid breach scenarios are contrasted with the background radiation

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.

Promoting and assessment of safety culture within regulatory body

International Nuclear Information System (INIS)

Awasthi, Sumit; Bhattacharya, D.; Koley, J.; Krishnamurthy, P.R.

2015-01-01

Regulators have an important role to play in assisting organizations under their jurisdiction to develop positive safety cultures. It is therefore essential for the regulator to have a robust safety culture as an inherent strategy and communication of this strategy to the organizations it supervises. Atomic Energy Regulatory Board (AERB) emphasizes every utility to institute a good safety culture during various stages of a NPP. The regulatory requirement for establishing organisational safety culture within utility at different stages are delineated in the various AERB safety codes which are presented in the paper. Although the review and assessment of the safety culture is a part of AERB’s continual safety supervision through existing review mechanism, AERB do not use any specific indicators for safety culture assessment. However, establishing and nurturing a good safety culture within AERB helps in encouraging the utility to institute the same. At the induction level AERB provides training to its staffs for regulatory orientation which include a specific course on safety culture. Subsequently, the junior staffs are mentored by seniors while involving them in various regulatory processes and putting them as observers during regulatory decision making process. Further, AERB established a formal procedure for assessing and improving safety culture within its staff as a management system process. The paper describes as a case study the above safety culture assessment process established within AERB

Probabilistic safety assessment for research reactors

International Nuclear Information System (INIS)

1986-12-01

Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

An examination of safety reports involving electronic flight bags and portable electronic devices

Science.gov (United States)

2014-06-01

The purpose of this research was to develop a better understanding of safety considerations with the use of Electronic Flight Bags (EFBs) and Portable Electronic Devices (PEDs) by examining safety reports from Aviation Safety Reporting System (ASRS),...

Safety assessment of radioactive wastes storage 'Mironova Gora'

International Nuclear Information System (INIS)

Serbryakov, B.; Karamushka, V.; Ostroborodov, V.

2000-01-01

A project of transforming the radioactive wastes storage 'Mironova Gora' is under development. A safety assessment of this storage facility was performed to gain assurance on the design decision. The assessment, which was based on the safety assessment methods developed for radioactive wastes repositories, is presented in this paper. (author)

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

International Nuclear Information System (INIS)

Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

2014-01-01

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

Energy Technology Data Exchange (ETDEWEB)

Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

2014-10-15

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

Control and metallurgical examination on safety injection piping

International Nuclear Information System (INIS)

Thebault, Y.; Grandjean, Y.; Gauthier, V.; Lambert, B.; Debustcher, B.

1998-01-01

From 1992 until 1997, cracking phenomena by thermal fatigue regarding safety injection piping were evidenced on several PWR 900 MW reactors. These events led EDF to the implementation of a first maintenance programme. In December 1996, a new leak occurred on an EDF 900 MW PWR in operation and was located on a safety injection pipe. In site inspections and metallurgical examinations carried out in the EDF hot Laboratory evidenced defects inside the pipe, out of the welding areas. These degradations are the consequence of a fatigue cracking phenomenon with thermal cycling linked to permanent tensile stresses. Following this incident, a programme of non destructive testing was implemented on all the EDF 900 MW plants. These inspections exhibited the same defects on other PWR 900 MW units. The results of the metallurgical examinations and also in site inspection results allowed EDF to understand the phenomenon and to validate an inspection programme on the one hand and a modification of the design of the circuits on the other hand. (authors)

Self-assessment of operational safety for nuclear power plants

International Nuclear Information System (INIS)

1999-12-01

Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and the safety culture as a whole. Although the primary beneficiaries of the self-assessment process are the plant and operating organization, the results of the self-assessments are also used, for example, to increase the confidence of the regulator in the safe operation of an installation, and could be used to assist in meeting obligations under the Convention on Nuclear Safety. Such considerations influence the form of assessment, as well as the type and detail of the results. The concepts developed in this report present the basic approach to self-assessment, taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject. This report will be used in IAEA sponsored workshops and seminars on operational safety that include the topic of self-assessment

A probabilistic safety assessment PEER review: Case study on the use of probabilistic safety assessment for safety decisions

International Nuclear Information System (INIS)

1989-10-01

The purpose of this case study is to illustrate, using an actual example, the organizing and carrying out of an independent peer review of a draft full-scope (level 3) probabilistic safety assessment. The specific findings of the peer review are of less importance than the approach taken, the interaction between sponsor and study team, and the technical and administrative issues that can arise during a peer review. This case study will examine the following issues: how the scope of the peer review was established, based on how it was to be used by the review sponsoring body; how the level of effort was determined, and what this determination meant for the technical quality of the review; how the team of peer reviewers was selected; how the review itself was carried out; what findings were made; what was done with these findings by both the review sponsoring body and the PSA analysis team. 9 refs, 2 figs, 1 tab

Analysis of truncation limit in probabilistic safety assessment

International Nuclear Information System (INIS)

Cepin, Marko

2005-01-01

A truncation limit defines the boundaries of what is considered in the probabilistic safety assessment and what is neglected. The truncation limit that is the focus here is the truncation limit on the size of the minimal cut set contribution at which to cut off. A new method was developed, which defines truncation limit in probabilistic safety assessment. The method specifies truncation limits with more stringency than presenting existing documents dealing with truncation criteria in probabilistic safety assessment do. The results of this paper indicate that the truncation limits for more complex probabilistic safety assessments, which consist of larger number of basic events, should be more severe than presently recommended in existing documents if more accuracy is desired. The truncation limits defined by the new method reduce the relative errors of importance measures and produce more accurate results for probabilistic safety assessment applications. The reduced relative errors of importance measures can prevent situations, where the acceptability of change of equipment under investigation according to RG 1.174 would be shifted from region, where changes can be accepted, to region, where changes cannot be accepted, if the results would be calculated with smaller truncation limit

Risk assessment of safety violations for coal mines

Energy Technology Data Exchange (ETDEWEB)

Megan Orsulaka; Vladislav Kecojevicb; Larry Graysona; Antonio Nietoa [Pennsylvania State University, University Park, PA (United States). Dept of Energy and Mineral Engineering

2010-09-15

This article presents an application of a risk assessment approach in characterising the risks associated with safety violations in underground bituminous mines in Pennsylvania using the Mine Safety and Health Administration (MSHA) citation database. The MSHA database on citations provides an opportunity to assess risks in mines through scrutiny of violations of mandatory safety standards. In this study, quantitative risk assessment is performed, which allows determination of the frequency of occurrence of safety violations (through associated citations) as well as the consequences of them in terms of penalty assessments. Focus is on establishing risk matrices on citation experiences of mines, which can give early indication of emerging potentially serious problems. The resulting frequency, consequence and risk rankings present valuable tools for prioritising resource allocations, determining control strategies, and could potentially contribute to more proactive prevention of incidents and injuries.

Healthcare professionals’ views of feedback on patient safety culture assessment.

OpenAIRE

Zwijnenberg, N.C.; Hendriks, M.; Hoogervorst-Schilp, J.; Wagner, C.

2016-01-01

Background: By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals’ views on the feedback of a patient safety culture assessment. Methods: Twenty four hospitals participated in a patient safety culture assessment in 2012. Hospital departments received feedback in a report and on a web...

Safety assessment of research reactors and preparation of the safety analysis report

International Nuclear Information System (INIS)

1994-01-01

This Safety Guide presents guidelines, approved by international consensus, for the preparation, review and assessment of safety documentation for research reactors such as the Safety Analysis Report. While the Guide is most applicable to research reactors in the design and construction stage, it is also recommended for use during relicensing or reassessment of existing reactors

Using Multilevel Analysis to Examine the Relationship between Upper Secondary Students Internet Safety Awareness, Social Background and Academic Aspirations

Directory of Open Access Journals (Sweden)

Ove Edvard Hatlevik

2014-11-01

Full Text Available Since 2009, most Norwegian students in upper secondary have had access to their own personal computer at school. Hence, with the increased access to technology, the importance of online connectedness has increased for adolescents’ social interaction and communication. It is, therefore, important to identify and understand the concept of Internet safety among upper secondary school students. A total of 4216 students from 238 classrooms in 23 upper secondary schools completed an Internet safety assessment. The aim of the study was to operationalize and measure Internet safety in a school context, and to further examine the factors predicting students’ Internet safety awareness and responsibility. Our analysis revealed substantial variation in Internet safety awareness between schools, classrooms and students. Overall, the findings indicate that students’ social backgrounds are determining for their development and understanding of Internet safety awareness.

Safety Assessment of Polyether Lanolins as Used in Cosmetics.

Science.gov (United States)

Becker, Lillian C; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan; Heldreth, Bart

The Cosmetic Ingredient Review (CIR) Expert Panel (Panel) assessed the safety of 39 polyether lanolin ingredients as used in cosmetics. These ingredients function mostly as hair conditioning agents, skin conditioning agent-emollients, and surfactant-emulsifying agents. The Panel reviewed available animal and clinical data, from previous CIR safety assessments of related ingredients and components. The similar structure, properties, functions, and uses of these ingredients enabled grouping them and using the available toxicological data to assess the safety of the entire group. The Panel concluded that these polyether lanolin ingredients are safe in the practices of use and concentration as given in this safety assessment.

Upgrading the safety assessment of exported nuclear power plants

International Nuclear Information System (INIS)

Rosen, M.

1978-01-01

An examination of the safety aspects of exported nuclear power plants demonstrates that additional and somewhat special considerations exist for these plants, and thus that some new approaches may be required to insure their safety. In view of the generally small regulatory staffs of importing countries, suggestions are given for measures which should be taken by the various organizations involved in the export and import of nuclear power facilities to raise the level of the very essential safety assessment. These include the upgrading of the 'export edition' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which highlights the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical Operating Regulations (Technical Specifications for Operation). Consideration is also given to upgrading the regulatory effort and to the obligations of principal organizations involved with exported nuclear plants, including national and international, for insuring the importing countries' technical readiness and the adequacy of the regulatory effort. Special attention is directed towards the project contract as a means of implementing programmes to achieve these goals. (author)

Probabilistic safety assessment as a standpoint for decision making

International Nuclear Information System (INIS)

Cepin, M.

2001-01-01

This paper focuses on the role of probabilistic safety assessment in decision-making. The prerequisites for use of the results of probabilistic safety assessment and the criteria for the decision-making based on probabilistic safety assessment are discussed. The decision-making process is described. It provides a risk evaluation of impact of the issue under investigation. Selected examples are discussed, which highlight the described process. (authors)

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

Directory of Open Access Journals (Sweden)

Järvis Marina

2016-10-01

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

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

Computer-simulated safety test of a fink-type roof truss

International Nuclear Information System (INIS)

Afolaxan, J.O.

2003-01-01

The safety of a Fink-type truss system following the BS5950/sup 1/ design requirements is examined. The design point approach of reliability analysis is developed to determine the safety indices for the individual members and the joints which are assumed filet-welded. Consequently, for assumed loading options, the entire system safety indices are reported indicating a need for the re-appraisal of the adopted criteria. (author)

Novi Han Radioactive Waste Repository post-closure safety assessment, ver.2

International Nuclear Information System (INIS)

Mateeva, M.

2003-01-01

The methodology for the post-closure safety assessment is presented. The assessment context includes regulatory framework (protection principles); scope and time frame; radiological and technical requirements; modeling etc. The description of the Novi Han disposal system contains site location. meteorological, hydrological and seismological characteristics; waste and repository description and human activities characteristics. The next step in the methodology is scenario development and justification. The systematic generation os exposure scenarios is considered as central to the post-closure safety assessment. The most important requirements for the systematic scenario generation approach are: transparency, comprehensiveness (all possible FEPs influencing the the disposal system and the radionuclide release should be considered); relevant future evolutions; identification of critical issues and investigation of the robustness of the system. For the source-pathway-receptor analysis the Process System is divided into near-field, geosphere/atmosphere and biosphere, describing the key facets controlling the potential radionuclide migration to the environment. The schematic division of the Novi Han near-field Process System into lower-level conceptual features is presented and discussed. As a result of the examinations of the FEPs three classes of scenarios are identified for the Novi Han post-closure safety assessment: Environmental evolution scenarios (geological change and climate change); future human action scenarios (human intrusion and archaeological action); Scenarios with very low probability (terrorism, crashes, explosions). The safety assessment iteration leads to identification of a modern scenario generation approach, assessment of key radionuclide releases, geological and hydrological evaluation, identification of the key parameters from sensitivity analysis etc. Examples of conceptual models are given. For the mathematical modeling the AMBER code is used

Need for an "integrated safety assessment" of GMOs, linking food safety and environmental considerations.

Science.gov (United States)

Haslberger, Alexander G

2006-05-03

Evidence for substantial environmental influences on health and food safety comes from work with environmental health indicators which show that agroenvironmental practices have direct and indirect effects on human health, concluding that "the quality of the environment influences the quality and safety of foods" [Fennema, O. Environ. Health Perspect. 1990, 86, 229-232). In the field of genetically modified organisms (GMOs), Codex principles have been established for the assessment of GM food safety and the Cartagena Protocol on Biosafety outlines international principles for an environmental assessment of living modified organisms. Both concepts also contain starting points for an assessment of health/food safety effects of GMOs in cases when the environment is involved in the chain of events that could lead to hazards. The environment can act as a route of unintentional entry of GMOs into the food supply, such as in the case of gene flow via pollen or seeds from GM crops, but the environment can also be involved in changes of GMO-induced agricultural practices with relevance for health/food safety. Examples for this include potential regional changes of pesticide uses and reduction in pesticide poisonings resulting from the use of Bt crops or influences on immune responses via cross-reactivity. Clearly, modern methods of biotechnology in breeding are involved in the reasons behind the rapid reduction of local varieties in agrodiversity, which constitute an identified hazard for food safety and food security. The health/food safety assessment of GM foods in cases when the environment is involved needs to be informed by data from environmental assessment. Such data might be especially important for hazard identification and exposure assessment. International organizations working in these areas will very likely be needed to initiate and enable cooperation between those institutions responsible for the different assessments, as well as for exchange and analysis of

Ending on a positive: Examining the role of safety leadership decisions, behaviours and actions in a safety critical situation.

Science.gov (United States)

Donovan, Sarah-Louise; Salmon, Paul M; Horberry, Timothy; Lenné, Michael G

2018-01-01

Safety leadership is an important factor in supporting safe performance in the workplace. The present case study examined the role of safety leadership during the Bingham Canyon Mine high-wall failure, a significant mining incident in which no fatalities or injuries were incurred. The Critical Decision Method (CDM) was used in conjunction with a self-reporting approach to examine safety leadership in terms of decisions, behaviours and actions that contributed to the incidents' safe outcome. Mapping the analysis onto Rasmussen's Risk Management Framework (Rasmussen, 1997), the findings demonstrate clear links between safety leadership decisions, and emergent behaviours and actions across the work system. Communication and engagement based decisions featured most prominently, and were linked to different leadership practices across the work system. Further, a core sub-set of CDM decision elements were linked to the open flow and exchange of information across the work system, which was critical to supporting the safe outcome. The findings provide practical implications for the development of safety leadership capability to support safety within the mining industry. Copyright © 2017 Elsevier Ltd. All rights reserved.

Ensuring the quality of occupational safety risk assessment.

Science.gov (United States)

Pinto, Abel; Ribeiro, Rita A; Nunes, Isabel L

2013-03-01

In work environments, the main aim of occupational safety risk assessment (OSRA) is to improve the safety level of an installation or site by either preventing accidents and injuries or minimizing their consequences. To this end, it is of paramount importance to identify all sources of hazards and assess their potential to cause problems in the respective context. If the OSRA process is inadequate and/or not applied effectively, it results in an ineffective safety prevention program and inefficient use of resources. An appropriate OSRA is an essential component of the occupational safety risk management process in industries. In this article, we performed a survey to elicit the relative importance for identified OSRA tasks to enable an in-depth evaluation of the quality of risk assessments related to occupational safety aspects on industrial sites. The survey involved defining a questionnaire with the most important elements (tasks) for OSRA quality assessment, which was then presented to safety experts in the mining, electrical power production, transportation, and petrochemical industries. With this work, we expect to contribute to the main question of OSRA in industries: "What constitutes a good occupational safety risk assessment?" The results obtained from the questionnaire showed that experts agree with the proposed OSRA process decomposition in steps and tasks (taxonomy) and also with the importance of assigning weights to obtain knowledge about OSRA task relevance. The knowledge gained will enable us, in the near future, to build a framework to evaluate OSRA quality for industrial sites. © 2012 Society for Risk Analysis.

Evaluation method of iodine re-evolution from an in-containment water pool after a loss of coolant accident, Part II: Evaluation of pH and iodine re-evolution

International Nuclear Information System (INIS)

Kim, Tae Hyeon; Jeong, Ji Hwan

2016-01-01

Highlights: • It is required to evaluate re-evolved iodine from sump water after LOCA. • Transport of iodine and chemicals influencing pH were analyzed using CFD. • Chemical conditions of the iodine-rich region suppress iodine re-evolution. • The current evaluation method for I 2 re-evolution is excessively conservative. - Abstract: Radioactive iodine that is released during a postulated loss of coolant accident is dissolved into the containment spray water and transported into the in-containment refueling water storage tank (IRWST). The re-evolution of iodine from the water is a safety concern. In this study, three-dimensional computational fluid dynamics (CFD) analyses are conducted in order to analyze the transport of chemical species including iodine in the IRWST and to calculate the amount of iodine that re-evolves from the IRWST water. The CFD analyses demonstrate that the pH of water is high where the iodine concentration is high. Considering that the creation rate of molecular iodine declines as the pH increases, it can be understood that the iodine re-evolution is not so strong in practical situations because the chemical conditions of the iodine-rich region suppress the re-evolution of the iodine. In addition, four different methods for evaluating the amount of re-evolved iodine are examined. The amount of re-evolved iodine calculated using the total-volume-average values, which are currently used for safety analyses, appear to be significantly higher than those determined using other methods. The amount of re-evolved iodine estimated using a realistic method with a conservative assumption of volatilization appears to be approximately one thousandth of that evaluated using the current method. This implies that the current method is very conservative.

Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

International Nuclear Information System (INIS)

Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

2008-01-01

The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design

Exploiting data from safety investigations and processes to assess performance of safety management aspects

NARCIS (Netherlands)

Karanikas, Nektarios

2016-01-01

This paper presents an alternative way to use records from safety investigations as a means to support the evaluation of safety management (SM) aspects. Datasets from safety investigation reports and progress records of an aviation organization were analyzed with the scope of assessing safety

Twenty-third water reactor safety information meeting: Volume 2, Human factors research; Advanced I and C hardware and software; Severe accident research; Probabilistic risk assessment topics; Individual plant examination: Proceedings

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1996-03-01

This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 2, present topics in human factors research, advanced instrumentation and control hardware and software, severe accident research, probabilistic risk assessment, and individual plant examination. Individual papers have been cataloged separately.

Assessment of Contributions to Patient Safety Knowledge by the Agency for Healthcare Research and Quality-Funded Patient Safety Projects

Science.gov (United States)

Sorbero, Melony E S; Ricci, Karen A; Lovejoy, Susan; Haviland, Amelia M; Smith, Linda; Bradley, Lily A; Hiatt, Liisa; Farley, Donna O

2009-01-01

Objective To characterize the activities of projects funded in Agency for Healthcare Research and Quality (AHRQ)' patient safety portfolio and assess their aggregate potential to contribute to knowledge development. Data Sources Information abstracted from proposals for projects funded in AHRQ' patient safety portfolio, information on safety practices from the AHRQ Evidence Report on Patient Safety Practices, and products produced by the projects. Study Design This represented one part of the process evaluation conducted as part of a longitudinal evaluation based on the Context–Input–Process–Product model. Principal Findings The 234 projects funded through AHRQ' patient safety portfolio examined a wide variety of patient safety issues and extended their work beyond the hospital setting to less studied parts of the health care system. Many of the projects implemented and tested practices for which the patient safety evidence report identified a need for additional evidence. The funded projects also generated a substantial body of new patient safety knowledge through a growing number of journal articles and other products. Conclusions The projects funded in AHRQ' patient safety portfolio have the potential to make substantial contributions to the knowledge base on patient safety. The full value of this new knowledge remains to be confirmed through the synthesis of results. PMID:21456108

NPP Krsko periodic safety review. Safety assessment and analyses

International Nuclear Information System (INIS)

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

2002-01-01

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

Radioactive waste management: the contribution of expert assessments to the implementation of safe management channels

International Nuclear Information System (INIS)

Besnus, F.; Jouve, A.C.

2011-01-01

The national Radioactive Materials and Waste Management (PNGMDR) sets objectives and defines waste management channels for all radioactive wastes produced in France. Within this framework, IRSN (Institute for Radioprotection and Nuclear Safety) expertise aims at assessing the consistency and robustness of the technical solutions set in place by the plan. As a result of this assessment, the main safety issues and priorities for upgrading the safety of the various facilities that will receive and treat waste are identified on the one hand, while possible foreseen weaknesses in terms of storage or treatment capacities are put into light on the other hand. To carry out such assessment, IRSN backs on its 'in depth' knowledge of facilities, acquired through the examination of each major step of waste management facility life (creation, commissioning, re-examination of safety...). This knowledge feeds in turn the examination of the waste management strategies implemented by operators. In addition, special attention is given to the achievement of waste packages of favourable properties as well as to the conditions for their safe disposal, since these two aspects are most often key factors for optimizing the safety of the whole management channel. By its capacity to overlook all steps of waste management channels, from production to final disposal, IRSN intends to contribute to the objective of enhancing the global safety of the management of radioactive waste. (authors)

Safety assessment of a lithium target

International Nuclear Information System (INIS)

Burgazzi, Luciano; Roberta, Ferri; Barbara, Giannone

2006-01-01

This paper addresses the safety assessment of the lithium target of the International Fusion Materials Irradiation Facility (IFMIF) through evaluating the most important risk factors related to system operation and verifying the fulfillment of the safety criteria. The hazard assessment is based on using a well-structured Failure Mode and Effect Analysis (FMEA) procedure by detailing on a component-by-component basis all the possible failure modes and identifying their effects on the plant. Additionally, a systems analysis, applying the fault tree technique, is performed in order to evaluate, from a probabilistic standpoint, all the relevant and possible failures of each component required for safe system operation and assessing the unavailability of the lithium target system. The last task includes the thermal-hydraulic transient analysis of the target lithium loop, including operational and accident transients. A lithium target loop model is developed, using the RELAP5/Mod3.2 thermal-hydraulic code, which has been modified to include specific features of IFMIF itself. The main conclusions are that target safety is fulfilled, the hazards associated with lithium operation are confined within the IFMIF security boundaries, the environmental impact is negligible, and the plant responds to the simulated transients by being able to reach steady conditions in a safety situation

Environment, safety and health progress assessment manual

International Nuclear Information System (INIS)

1992-12-01

On June 27, 1989, the Secretary of Energy announced a 1O-Point Initiative to strengthen environment,safety, and health (ES ampersand H) programs, and waste management activities at involved conducting DOE production, research, and testing facilities. One of the points independent Tiger Team Assessments of DOE operating facilities. The Office of Special Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This volume contains appendices to the Environment, Safety and Health Progress Assessment Manual

Schools and Disasters: Safety and Mental Health Assessment and Interventions for Children.

Science.gov (United States)

Lai, Betty S; Esnard, Ann-Margaret; Lowe, Sarah R; Peek, Lori

2016-12-01

This article draws on experiences and lessons from global disasters and utilizes the United Nations Comprehensive School Safety Framework to highlight the necessary role of safe schools in protecting children, as well as adult staff, from the immediate threats and long-term implications of disasters. Specifically, we focus on three well-established pillars of school safety: Pillar I: Safe Learning Facilities; Pillar II: Disaster Management; and Pillar III: Risk Reduction and Resilience Education. In addition, we propose a potential fourth pillar, which underscores the function of schools in postdisaster mental health assessment and intervention for children. We argue that schools offer a central location and trusted institutional space for mental health assessment and intervention after disasters. We also examine the important linkages between schools, child mental health, and household and family recovery. We conclude with recommendations for filling gaps in research and practice related to ensuring the safety of schools and the associated health and well-being of children in the face of future disasters.

The role of natural analogues in safety assessment and acceptability

International Nuclear Information System (INIS)

Papp, Toenis

1987-01-01

The safety assessment must evaluate the level of safety for a repository, the confidence that can be placed on the assessment and how well the repository can meet the acceptance criteria of the society. Many of the processes and phenomena that govern the long term performance of a deep geologic repository for radioactive waste also take place in nature. To investigate these natural analogues and try to validate the models on which the safety assessment are based is a main task in the effort to build of confidence in the safety assessments. The assessment of the safety of a repository can, however, not only be based on good models. The possible role of natural analogues or natural evidence in other parts of the safety assessment is discussed. Specially with regard to - the need to demonstrate that all relevant processes have been taken into account, and that the important ones have been validated to an acceptable level for relevant parameters spans, -the definition and analysis of external scenarios for the safety assessment and for the claim that all reasonable scenarios have been addressed, - the public confidence in the long-term relevance of the acceptance criteria. (author)

Safety assessment of HLW geological disposal system

International Nuclear Information System (INIS)

Naito, Morimasa

2006-01-01

In accordance with the Japanese nuclear program, the liquid waste with a high level of radioactivity arising from reprocessing is solidified in a stable glass matrix (vitrification) in stainless steel fabrication containers. The vitrified waste is referred to as high-level radioactive waste (HLW), and is characterized by very high initial radioactivity which, even though it decreases with time, presents a potential long-term risk. It is therefore necessary to thoroughly manage HLW from human and his environment. After vitrification, HLW is stored for a period of 30 to 50 years to allow cooling, and finally disposed of in a stable geological environment at depths greater than 300 m below surface. The deep underground environment, in general, is considered to be stable over geological timescales compared with surface environment. By selecting an appropriate disposal site, therefore, it is considered to be feasible to isolate the waste in the repository from man and his environment until such time as radioactivity levels have decayed to insignificance. The concept of geological disposal in Japan is similar to that in other countries, being based on a multibarrier system which combines the natural geological environment with engineered barriers. It should be noted that geological disposal concept is based on a passive safety system that does not require any institutional control for assuring long term environmental safety. To demonstrate feasibility of safe HLW repository concept in Japan, following technical steps are essential. Selection of a geological environment which is sufficiently stable for disposal (site selection). Design and installation of the engineered barrier system in a stable geological environment (engineering measures). Confirmation of the safety of the constructed geological disposal system (safety assessment). For site selection, particular consideration is given to the long-term stability of the geological environment taking into account the fact

Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

International Nuclear Information System (INIS)

Armstrong, D.L.

1994-08-01

This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment

Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

Energy Technology Data Exchange (ETDEWEB)

Armstrong, D.L.

1994-08-01

This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment.

A min cut-set-wise truncation procedure for importance measures computation in probabilistic safety assessment

Energy Technology Data Exchange (ETDEWEB)

Duflot, Nicolas [Universite de technologie de Troyes, Institut Charles Delaunay/LM2S, FRE CNRS 2848, 12, rue Marie Curie, BP2060, F-10010 Troyes cedex (France)], E-mail: nicolas.duflot@areva.com; Berenguer, Christophe [Universite de technologie de Troyes, Institut Charles Delaunay/LM2S, FRE CNRS 2848, 12, rue Marie Curie, BP2060, F-10010 Troyes cedex (France)], E-mail: christophe.berenguer@utt.fr; Dieulle, Laurence [Universite de technologie de Troyes, Institut Charles Delaunay/LM2S, FRE CNRS 2848, 12, rue Marie Curie, BP2060, F-10010 Troyes cedex (France)], E-mail: laurence.dieulle@utt.fr; Vasseur, Dominique [EPSNA Group (Nuclear PSA and Application), EDF Research and Development, 1, avenue du Gal de Gaulle, 92141 Clamart cedex (France)], E-mail: dominique.vasseur@edf.fr

2009-11-15

A truncation process aims to determine among the set of minimal cut-sets (MCS) produced by a probabilistic safety assessment (PSA) model which of them are significant. Several truncation processes have been proposed for the evaluation of the probability of core damage ensuring a fixed accuracy level. However, the evaluation of new risk indicators as importance measures requires to re-examine the truncation process in order to ensure that the produced estimates will be accurate enough. In this paper a new truncation process is developed permitting to estimate from a single set of MCS the importance measure of any basic event with the desired accuracy level. The main contribution of this new method is to propose an MCS-wise truncation criterion involving two thresholds: an absolute threshold in addition to a new relative threshold concerning the potential probability of the MCS of interest. The method has been tested on a complete level 1 PSA model of a 900 MWe NPP developed by 'Electricite de France' (EDF) and the results presented in this paper indicate that to reach the same accuracy level the proposed method produces a set of MCS whose size is significantly reduced.

A min cut-set-wise truncation procedure for importance measures computation in probabilistic safety assessment

International Nuclear Information System (INIS)

Duflot, Nicolas; Berenguer, Christophe; Dieulle, Laurence; Vasseur, Dominique

2009-01-01

A truncation process aims to determine among the set of minimal cut-sets (MCS) produced by a probabilistic safety assessment (PSA) model which of them are significant. Several truncation processes have been proposed for the evaluation of the probability of core damage ensuring a fixed accuracy level. However, the evaluation of new risk indicators as importance measures requires to re-examine the truncation process in order to ensure that the produced estimates will be accurate enough. In this paper a new truncation process is developed permitting to estimate from a single set of MCS the importance measure of any basic event with the desired accuracy level. The main contribution of this new method is to propose an MCS-wise truncation criterion involving two thresholds: an absolute threshold in addition to a new relative threshold concerning the potential probability of the MCS of interest. The method has been tested on a complete level 1 PSA model of a 900 MWe NPP developed by 'Electricite de France' (EDF) and the results presented in this paper indicate that to reach the same accuracy level the proposed method produces a set of MCS whose size is significantly reduced.

RSAS: a Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Dixon, B.W.; Bray, M.A.

1985-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

Status report on seismic re-evaluation

International Nuclear Information System (INIS)

1998-01-01

In May 1997, a meeting of the PWG 3 Sub Group on the Seismic Behaviour of Structures agreed several priority objectives, of which one was the production of a status report on seismic re-evaluation. Seismic re-evaluation is identified as the process of carrying out a re-assessment of the safety of existing nuclear power plants for a specified seismic hazard. This may be necessary when no seismic hazard was considered in the original design of the plant, the relevant codes and regulations have been revised, the seismic hazard for the site has been re-assessed or there is a need to assess the capacity of the plant for severe accident conditions and behaviour beyond the design basis. Re-evaluation may also be necessary to resolve an issue, or to assess the impact of new findings or knowledge. A questionnaire on the subject was issued to all members of the Seismic Sub Group in the summer of 1997, and responses to the questionnaire had been received from most members by the end of 1997. This report is based on the responses to the questionnaire, together with comment and discussion within the group. The questionnaire covered the following main topics of interest in relation to seismic re-evaluation: General and Legislative Framework, Overall Approach, Input Definition and Analysis Methods, Scope of Plant and Assessment of As-built Situation, Assessment criteria, Outcome of Re-evaluations, Research. The responses to the questionnaire have been collated and reviewed with the objective of comparing current practice in the field of seismic re-evaluation in member countries, and a number of important points have been identified in relation to the position of seismic re-evaluation in the nuclear power industry throughout the world. It is evident that seismic re-evaluation is a relatively mature process that has been developing for some time, with most countries adopting similar practices, often based on principles which have been developed in the US nuclear industry. Seismic

Preliminary safety examination on thorium molten-salt nuclear energy synergetics

International Nuclear Information System (INIS)

Furukawa, Kazuo

1991-01-01

The new global fission industry for the next century should keep a strong public acceptance, which means to ensure an enough rational safety feature not only in the engineering issue but also in all issues of integral fuel-cycle system. In this sense, the safety characteristics of the Thorium Molten-Salt Nuclear Energy Synergetic System (THORIMS-NES) is widely examined relating with the several aspects of environmental (including resources, radio-waste, etc.) social (including anti-nuclear proliferation and terrorism, etc), basic technological, engineering, institutional, and economical aspects. From this examination it seems that this system is verified as one of the most promising measures of North-South problem, greenhouse effect, etc in the world. (author). 11 refs., 3 figs., 5 tabs

Safety assessment and detection methods of genetically modified organisms.

Science.gov (United States)

Xu, Rong; Zheng, Zhe; Jiao, Guanglian

2014-01-01

Genetically modified organisms (GMOs), are gaining importance in agriculture as well as the production of food and feed. Along with the development of GMOs, health and food safety concerns have been raised. These concerns for these new GMOs make it necessary to set up strict system on food safety assessment of GMOs. The food safety assessment of GMOs, current development status of safety and precise transgenic technologies and GMOs detection have been discussed in this review. The recent patents about GMOs and their detection methods are also reviewed. This review can provide elementary introduction on how to assess and detect GMOs.

Nuclear power and probabilistic safety assessment (PSA): past through future applications

Science.gov (United States)

Stamatelatos, M. G.; Moieni, P.; Everline, C. J.

1995-03-01

Nuclear power reactor safety in the United States is about to enter a new era -- an era of risk- based management and risk-based regulation. First, there was the age of `prescribed safety assessment,' during which a series of design-basis accidents in eight categories of severity, or classes, were postulated and analyzed. Toward the end of that era, it was recognized that `Class 9,' or `beyond design basis,' accidents would need special attention because of the potentially severe health and financial consequences of these accidents. The accident at Three Mile Island showed that sequences of low-consequence, high-frequency events and human errors can be much more risk dominant than the Class 9 accidents. A different form of safety assessment, PSA, emerged and began to gain ground against the deterministic safety establishment. Eventually, this led to the current regulatory requirements for individual plant examinations (IPEs). The IPEs can serve as a basis for risk-based regulation and management, a concept that may ultimately transform the U.S. regulatory process from its traditional deterministic foundations to a process predicated upon PSA. Beyond the possibility of a regulatory environment predicated upon PSA lies the possibility of using PSA as the foundation for managing daily nuclear power plant operations.

Guidelines for Self-assessment of Research Reactor Safety

International Nuclear Information System (INIS)

2018-01-01

Self-assessment is an organization’s internal process to review its current status, processes and performance against predefined criteria and thereby to provide key elements for the organization’s continual development and improvement. Self-assessment helps the organization to think through what it is expected to do, how it is performing in relation to these expectations, and what it needs to do to improve performance, fulfil the expectations and achieve better compliance with the predefined criteria. This publication provides guidelines for a research reactor operating organization to perform a self-assessment of the safety management and the safety of the facility and to identify gaps between the current situation and the IAEA safety requirements for research reactors. These guidelines also provide a methodology for Member States, regulatory bodies and operating organizations to perform a self-assessment of their application of the provisions of the Code of Conduct on the Safety of Research Reactors. This publication also addresses planning, implementation and follow-up of actions to enhance safety and strengthen application of the Code. The guidelines are applicable to all types of research reactor and critical and subcritical assemblies, at all stages in their lifetimes, and to States, regulatory bodies and operating organizations throughout all phases of research reactor programmes. Research reactor operating organizations can use these guidelines at any time to support self-assessments conducted in accordance with the organization’s integrated management system. These guidelines also serve as a tool for an organization to prepare to receive an IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission. An important result of this is the opportunity for an operating organization to identify focus areas and make safety improvements in advance of an INSARR mission, thereby increasing the effectiveness of the mission and efficiency of the

Safety Culture Assessment at Regulatory Body - PNRA Experience of Implementing IAEA Methodology for Safety Culture Self Assessment

International Nuclear Information System (INIS)

Bhatti, S.A.N.; Arshad, N.

2016-01-01

The prevalence of a good safety culture is equally important for all kind of organizations involved in nuclear business including operating organizations, designers, regulator, etc., and this should be reflected through all the processes and activities of these organizations. The need for inculcating safety culture into regulatory processes and practices is gradually increasing since the major accident at Fukushima. Accordingly, several international fora in last few years repeatedly highlighted the importance of prevalence of safety culture in regulatory bodies as well. The utilisation of concept of safety culture always remained applicable in regulatory activities of PNRA in the form of core values. After the Fukushima accident, PNRA considered it important to check the extent of utilisation of safety culture concept in organizational activities and decided to conduct its “Safety Culture Self-Assessment (SCSA)” for presenting itself as a role model in-order to endorse the fact that safety culture at regulatory authority plays an important role to influence safety culture at licenced facilities.

ASCOT guidelines revised 1996 edition. Guidelines for organizational self-assessment of safety culture and for reviews by the assessment of safety culture in organizations team

International Nuclear Information System (INIS)

1996-01-01

In order to properly assess safety culture, it is necessary to consider the contribution of all organizations which have an impact on it. Therefore, while assessing the safety culture in an operating organization it is necessary to address at least its interfaces with the local regulatory agency, utility corporate headquarters and supporting organizations. These guidelines are primarily intended for use by any organization wishing to conduct a self-assessment of safety culture. They should also serve as a basis for conducting an international peer review of the organization's self-assessment carried out by an ASCOT (Assessment of Safety Culture in Organizations Team) mission

Problems of rotator cuff re-tear cases. Examination of operative findings

International Nuclear Information System (INIS)

Ishitani, Eiichi; Matsuura, Komei; Shin, Kunichika; Kawamoto, Taisaku; Hosokawa, Akira

2007-01-01

High re-tear rate is reported after rotator cuff repair in large and massive tear cases. Previously, we reported that 15% of patients after rotator cuff repair showed re-tear in MRI findings. In this study, 116 consecutive rotator cuff repaired patients who had been operated arthroscopically since 2003 were evaluated for size, torn site, fixation methods, mobility, and tendon quality. In addition, rehabilitation methods and occupation after revision were also evaluated. There were four patients who underwent re-operation. Rotator cuff repairs by revision surgery failed mechanically due to two reasons: the main factor of failure was suture material breakage in three cases and the second was tendon pulling through sutures in one case. The lack of thread strength was suggested. In addition, it is important to pull cuff stump to greater tubercle without excessive tension. Of re-operated cases, three engaging in manual labor suffered large and massive tear. In two cases, premature return to manual labor suggested cause of re-tear. (author)

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

IRSN-ANCCLI partnership. Work session on Complementary safety assessments - November 2011

International Nuclear Information System (INIS)

Lachaume, Jean-Luc; Lheureux, Yves; Sene, Monique; Sene, Raymond; Jorel, Martial; Lavarenne, Caroline; Rousseau, Jean-Marie; Rebour, Vincent; Baumont, David; Dupuy, Patricia

2011-11-01

After an overview by the ASN of complementary safety assessments and an assessment of 'post-Fukushima' inspections of basic nuclear installations, the contributions (Power Point presentations) of this seminar proposed: the opinion of the Gravelines CLI (local information commission) on the Gravelines complementary safety assessment report, an analysis and discussion by the GSIEN on reports of complementary assessment of safety of nuclear installations with respect to the Fukushima accident, an analysis by the IRSN of complementary safety assessments performed by operators, the IRSN approach to analyze complementary safety assessments, reports on installation conditions, external flooding and seismic hazard, 'meltdown prevention' aspects in the management of accidental situations in EDF reactors

Initial development of a practical safety audit tool to assess fleet safety management practices.

Science.gov (United States)

Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

2012-07-01

Work-related vehicle crashes are a common cause of occupational injury. Yet, there are few studies that investigate management practices used for light vehicle fleets (i.e. vehicles less than 4.5 tonnes). One of the impediments to obtaining and sharing information on effective fleet safety management is the lack of an evidence-based, standardised measurement tool. This article describes the initial development of an audit tool to assess fleet safety management practices in light vehicle fleets. The audit tool was developed by triangulating information from a review of the literature on fleet safety management practices and from semi-structured interviews with 15 fleet managers and 21 fleet drivers. A preliminary useability assessment was conducted with 5 organisations. The audit tool assesses the management of fleet safety against five core categories: (1) management, systems and processes; (2) monitoring and assessment; (3) employee recruitment, training and education; (4) vehicle technology, selection and maintenance; and (5) vehicle journeys. Each of these core categories has between 1 and 3 sub-categories. Organisations are rated at one of 4 levels on each sub-category. The fleet safety management audit tool is designed to identify the extent to which fleet safety is managed in an organisation against best practice. It is intended that the audit tool be used to conduct audits within an organisation to provide an indicator of progress in managing fleet safety and to consistently benchmark performance against other organisations. Application of the tool by fleet safety researchers is now needed to inform its further development and refinement and to permit psychometric evaluation. Copyright © 2012 Elsevier Ltd. All rights reserved.

Animal-Free Chemical Safety Assessment

Directory of Open Access Journals (Sweden)

George D Loizou

2016-07-01

Full Text Available The exponential growth of the Internet of Things and the global popularity and remarkable decline in cost of the mobile phone is driving the digital transformation of medical practice. The rapidly maturing digital, nonmedical world of mobile (wireless devices, cloud computing and social networking is coalescing with the emerging digital medical world of omics data, biosensors and advanced imaging which offers the increasingly realistic prospect of personalized medicine. Described as a potential seismic shift from the current healthcare model to a wellness paradigm that is predictive, preventative, personalized and participatory, this change is based on the development of increasingly sophisticated biosensors which can track and measure key biochemical variables in people. Additional key drivers in this shift are metabolomic and proteomic signatures, which are increasingly being reported as pre-symptomatic, diagnostic and prognostic of toxicity and disease. These advancements also have profound implications for toxicological evaluation and safety assessment of pharmaceuticals and environmental chemicals. An approach based primarily on human in vivo and high-throughput in vitro human cell-line data is a distinct possibility. This would transform current chemical safety assessment practise which operates in a human data poor to a human data rich environment. This could also lead to a seismic shift from the current animal-based to an animal-free chemical safety assessment paradigm.

Effects of outsourcing magnetic resonance examinations from a public university hospital to a private agent.

Science.gov (United States)

Tavakol, Parvin; Labruto, Fausto; Bergstrand, Lott; Blomqvist, Lennart

2011-02-01

Sometimes the measures taken to make a radiology department more effective, such as prioritizing the workload and keeping equipment running for as many hours as staffing permits, are not enough. In such cases, outsourcing radiological examinations is a potential solution for reducing waiting times. To investigate differences in waiting time, quality and costs between magnetic resonance (MR) examinations performed in a university hospital and examinations outsourced to private service. We retrospectively selected a group of consecutive, outsourced MR examinations (n=97) and a control group of in-house MR examinations, matched for type of examination. In each group there were referrals that had a specified preferred timeframe for completion. We measured the percentage of cases in which this timeframe was met and if it was not met, how many days exceeded the preferred time. In referrals without a specified preferred timeframe, we also calculated the waiting time. Quality standards were measured by the percentage of examinations that had to be re-done and re-assessed. Finally, we calculated the cumulative costs, taking into account the costs for re-doing and re-assessing examinations. There was no statistically significant difference between the groups, in either the number of examinations that were not performed within the preferred time or the number of days that exceeded the preferred timeframe. For referrals without a preferred timeframe, the waiting time was shorter for outsourced examinations than those not outsourced. There were no differences in the number of examinations that had to be re-done, but more examinations needed to be re-assessed in the outsourced group than in the in-house group. The calculated costs for outsourced examinations were lower than the costs for internally performed examinations. Outsourcing magnetic resonance examinations may be an effective way of reducing a radiology department's workload. Ways in which to reduce the additional costs

Preliminary investigation on reliability assessment of passive safety system

International Nuclear Information System (INIS)

Huang Changfan; Kuang Bo

2012-01-01

The reliability evaluation of passive safety system plays an important part in probabilistic safety assessment (PSA) of nuclear power plant applying passive safety design, which depends quantitatively on reliabilities of passive safety system. According to the object of reliability assessment of passive safety system, relevant parameters are identified. Then passive system behavior during accident scenarios are studied. A practical example of this method is given for the case of reliability assessment of AP1000 passive heat removal system in loss of normal feedwater accident. Key and design parameters of PRHRS are identified and functional failure criteria are established. Parameter combinations acquired by Latin hyper~ cube sampling (LHS) in possible parametric ranges are input and calculations of uncertainty propagation through RELAP5/MOD3 code are carried out. Based on the calculations, sensitivity assessment on PRHRS functional criteria and reliability evaluation of the system are presented, which might provide further PSA with PRHR system reliability. (authors)

Probabilistic safety assessment of the Fugen NPS

International Nuclear Information System (INIS)

Sotsu, Masutake; Iguchi, Yukihiro; Mizuno, Kouichi; Sato, Shinichirou; Shimizu, Miwako

1999-01-01

We performed a probabilistic safety assessment (PSA) on the Fugen NPS. The main topic of assessment was internal factors. We assessment core damage frequency (level 1 PSA) and containment damage frequency (level 2 PSA) during rated operation, and core damage frequency during shutdown (PSA during shutdowns). Our assessment showed that the core damage frequency of Fugen is well below the IAEA criteria for existing plants, that the conditional containment damage during shutdown is almost the target value of 0.1, and that the core damage frequency during shutdown is almost the same as that assessed during operation. These results confirm that the Fugen plant maintains a sufficient safety margin during shutdowns for regular inspections and for refueling. We developed and verified the effectiveness of an accident management plan incorporating the results of the assessment. (author)

Safety standards for near surface disposal and the safety case and supporting safety assessment for demonstrating compliance with the standards

International Nuclear Information System (INIS)

Metcalf, P.

2003-01-01

The report presents the safety standards for near surface disposal (ICRP guidance and IAEA standards) and the safety case and supporting safety assessment for demonstrating compliance with the standards. Special attention is paid to the recommendations for disposal of long-lived solid radioactive waste. The requirements are based on the principle for the same level of protection of future individuals as for the current generation. Two types of exposure are considered: human intrusion and natural processes and protection measures are discussed. Safety requirements for near surface disposal are discussed including requirements for protection of human health and environment, requirements or safety assessments, waste acceptance and requirements etc

Failure rate data for fusion safety and risk assessment

International Nuclear Information System (INIS)

Cadwallader, L.C.

1993-01-01

The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

Fusion blanket inherent safety assessment

International Nuclear Information System (INIS)

Sze, D.K.; Jung, J.; Cheng, E.T.

1986-01-01

Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

Mathematical Safety Assessment Approaches for Thermal Power Plants

Directory of Open Access Journals (Sweden)

Zong-Xiao Yang

2014-01-01

Full Text Available How to use system analysis methods to identify the hazards in the industrialized process, working environment, and production management for complex industrial processes, such as thermal power plants, is one of the challenges in the systems engineering. A mathematical system safety assessment model is proposed for thermal power plants in this paper by integrating fuzzy analytical hierarchy process, set pair analysis, and system functionality analysis. In the basis of those, the key factors influencing the thermal power plant safety are analyzed. The influence factors are determined based on fuzzy analytical hierarchy process. The connection degree among the factors is obtained by set pair analysis. The system safety preponderant function is constructed through system functionality analysis for inherence properties and nonlinear influence. The decision analysis system is developed by using active server page technology, web resource integration, and cross-platform capabilities for applications to the industrialized process. The availability of proposed safety assessment approach is verified by using an actual thermal power plant, which has improved the enforceability and predictability in enterprise safety assessment.

Buffer and backfill process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrik (comp.)

2006-09-15

This document compiles information on processes in the buffer and deposition tunnel backfill relevant for long-term safety of a KBS-repository. It supports the safety assessment SR-Can, which is a preparatory step for a safety assessment that will support the licence application for a final repository in Sweden. The purpose of the process reports is to document the scientific knowledge of the processes to a level required for an adequate treatment of the processes in the safety assessment. The documentation is not exhaustive from a scientific point of view, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. However, it must be sufficiently detailed to motivate, by arguments founded on scientific understanding, the treatment of each process in the safety assessment. The purpose is further to determine how to handle each process in the safety assessment at an appropriate degree of detail, and to demonstrate how uncertainties are taken care of, given the suggested handling.

Buffer and backfill process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Sellin, Patrik

2006-09-01

This document compiles information on processes in the buffer and deposition tunnel backfill relevant for long-term safety of a KBS-repository. It supports the safety assessment SR-Can, which is a preparatory step for a safety assessment that will support the licence application for a final repository in Sweden. The purpose of the process reports is to document the scientific knowledge of the processes to a level required for an adequate treatment of the processes in the safety assessment. The documentation is not exhaustive from a scientific point of view, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. However, it must be sufficiently detailed to motivate, by arguments founded on scientific understanding, the treatment of each process in the safety assessment. The purpose is further to determine how to handle each process in the safety assessment at an appropriate degree of detail, and to demonstrate how uncertainties are taken care of, given the suggested handling

Exploring relationships between patient safety culture and patients' assessments of hospital care.

Science.gov (United States)

Sorra, Joann; Khanna, Kabir; Dyer, Naomi; Mardon, Russ; Famolaro, Theresa

2014-10-01

The purpose of this study was to examine relationships among 2 Agency for Healthcare Research and Quality measures of hospital patient safety and quality, which reflect different perspectives on hospital performance: the Hospital Survey on Patient Safety Culture (Hospital SOPS)--a hospital employee patient safety culture survey--and the Consumer Assessment of Healthcare Providers and Systems Hospital Survey (CAHPS Hospital Survey)--a survey of the experiences of adult inpatients with hospital care and services. Our hypothesis was that these 2 measures would be positively related. We performed multiple regressions to examine the relationships between the Hospital SOPS measures and CAHPS Hospital Survey measures, controlling for hospital bed size and ownership. Analyses were conducted at the hospital level with each survey's measures using data from 73 hospitals that administered both surveys during similar periods. Higher overall Hospital SOPS composite average scores were associated with higher overall CAHPS Hospital Survey composite average scores (r = 0.41, P G 0.01). Twelve of 15 Hospital SOPS measures were positively related to the CAHPS Hospital Survey composite average score after controlling for bed size and ownership, with significant standardized regression coefficients ranging from 0.25 to 0.38. None of the Hospital SOPS measures were significantly correlated with either of the two single-item CAHPS Hospital Survey measures (hospital rating and willingness to recommend). This study found that hospitals where staff have more positive perceptions of patient safety culture tend to have more positive assessments of care from patients. This finding helps validate both surveys and suggests that improvements in patient safety culture may lead to improved patient experience with care. Further research is needed to determine the generalizability of these results to larger sets of hospitals, to hospital units, and to other settings of care.

A Computer Program for Assessing Nuclear Safety Culture Impact

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

Through several accidents of NPP including the Fukushima Daiichi in 2011 and Chernobyl accidents in 1986, a lack of safety culture was pointed out as one of the root cause of these accidents. Due to its latent influences on safety performance, safety culture has become an important issue in safety researches. Most of the researches describe how to evaluate the state of the safety culture of the organization. However, they did not include a possibility that the accident occurs due to the lack of safety culture. Because of that, a methodology for evaluating the impact of the safety culture on NPP's safety is required. In this study, the methodology for assessing safety culture impact is suggested and a computer program is developed for its application. SCII model which is the new methodology for assessing safety culture impact quantitatively by using PSA model. The computer program is developed for its application. This program visualizes the SCIs and the SCIIs. It might contribute to comparing the level of the safety culture among NPPs as well as improving the management safety of NPP.

Healthcare professionals? views on feedback of a patient safety culture assessment

OpenAIRE

Zwijnenberg, Nicolien C.; Hendriks, Michelle; Hoogervorst-Schilp, Janneke; Wagner, Cordula

2016-01-01

Background By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals? views on the feedback of a patient safety culture assessment. Methods Twenty four hospitals participated in a patient safety culture assessment in 2012. Hospital departments received feedback in a report and on a websi...

Safety assessment of the liquid-fed ceramic melter process

International Nuclear Information System (INIS)

Buelt, J.L.; Partain, W.L.

1980-08-01

As part of its development program for the solidification of high-level nuclear waste, Pacific Northwest Laboratory assessed the safety issues for a complete liquid-fed ceramic melter (LFCM) process. The LFCM process, an adaption of commercial glass-making technology, is being developed to convert high-level liquid waste from the nuclear fuel cycle into glass. This safety assessment uncovered no unresolved or significant safety problems with the LFCM process. Although in this assessment the LFCM process was not directly compared with other solidification processes, the safety hazards of the LFCM process are comparable to those of other processes. The high processing temperatures of the glass in the LFCM pose no additional significant safety concerns, and the dispersible inventory of dried waste (calcine) is small. This safety assessment was based on the nuclear power waste flowsheet, since power waste is more radioactive than defense waste at the time of solidification, and all accident conditions for the power waste would have greater radiological consequences than those for defense waste. An exhaustive list of possible off-standard conditions and equipment failures was compiled. These accidents were then classified according to severity of consequence and type of accident. Radionuclide releases to the stack were calculated for each group of accidents using conservative assumptions regarding the retention and decontamination features of the process and facility. Two recommendations that should be considered by process designers are given in the safety assessment

Safety assessment of novel foods and strategies to determine their safety in use

International Nuclear Information System (INIS)

Edwards, Gareth

2005-01-01

Safety assessment of novel foods requires a different approach to that traditionally used for the assessment of food chemicals. A case-by-case approach is needed which must be adapted to take account of the characteristics of the individual novel food. A thorough appraisal is required of the origin, production, compositional analysis, nutritional characteristics, any previous human exposure and the anticipated use of the food. The information should be compared with a traditional counterpart of the food if this is available. In some cases, a conclusion about the safety of the food may be reached on the basis of this information alone, whereas in other cases, it will help to identify any nutritional or toxicological testing that may be required to further investigate the safety of the food. The importance of nutritional evaluation cannot be over-emphasised. This is essential for the conduct of toxicological studies in order to avoid dietary imbalances, etc., that might lead to interpretation difficulties, but also in the context of its use as food and to assess the potential impact of the novel food on the human diet. The traditional approach used for chemicals, whereby an acceptable daily intake (ADI) is established with a large safety margin relative to the expected exposure, cannot be applied to foods. The assessment of safety in use should be based upon a thorough knowledge of the composition of the food, evidence from nutritional, toxicological and human studies, expected use of the food and its expected consumption. Safety equates to a reasonable certainty that no harm will result from intended uses under the anticipated conditions of consumption

LANL Safety Conscious Work Environment (SCWE) Self-Assessment

Energy Technology Data Exchange (ETDEWEB)

Hargis, Barbara C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2014-01-29

On December 21, 2012 Secretary of Energy Chu transmitted to the Defense Nuclear Facilities Safety Board (DNFSB) revised commitments on the implementation plan for Safety Culture at the Waste Treatment and Immobilization Plant. Action 2-5 was revised to require contractors and federal organizations to complete Safety Conscious Work Environment (SCWE) selfassessments and provide reports to the appropriate U.S. Department of Energy (DOE) - Headquarters Program Office by September 2013. Los Alamos National Laboratory (LANL) planned and conducted a Safety Conscious Work Environment (SCWE) Self-Assessment over the time period July through August, 2013 in accordance with the SCWE Self-Assessment Guidance provided by DOE. Significant field work was conducted over the 2-week period August 5-16, 2013. The purpose of the self-assessment was to evaluate whether programs and processes associated with a SCWE are in place and whether they are effective in supporting and promoting a SCWE.

Safety assessment for the 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generation

Energy Technology Data Exchange (ETDEWEB)

Suk, Ho Chun; Cho, M. S.; Jun, J. S. and others

2001-06-01

This document is a report on the safety assessment for the 24 CANFLEX-NU(CANDU Flexible fuelling - Natural Uranium) fuel bundle demonstration irradiation at Wolsong-1 Generating Station. The CANFLEX fuel bundle as a CANDU advanced fuel has been jointly developed by KAERI/AECL. This document describes the rationale for the demonstration irradiation and comments on the Korean government licensing issues such as the status of the CANFLEX fuel irradiations at NRU research reactor in AECL, status and plan of the CANFLEX fuel irradiations at a CANDU-6 power reactor, status of the water CHF(Critical Heat Flux) test at Stern Laboratories and the CHF correlation. This documents presents an assessment the consequences of postulated accidents with all safety system available during demonstration irradiation of 24 CANFLEX-NU fuel bundles at Wolsong-1 Generating Station. The assessment is made by two kinds of approaches. One approach is based on the document of the safety assessment for the 24 CANFLEX-NU fuel bundle demonstration irradiation at Point Lepreau Generating Station. The other approach is taken from the safety analyses using the analysis methods and assumptions used in the final safety reports on the 600 MWe CANDU-PHWR Wolsung-2, 3, and 4 Nuclear Power Plants for the Korea Electric Power Cooperation. The analyses are not comprehensive reviews of the postulated accidents, but examination of the expected difference in accident consequences because of the presence of 24 CANFLEX fuel bundles in two channels. The approach is to compare the difference to the safety margin for 37-element bundle cases.

Quantitative risk assessment of digitalized safety systems

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Safety assessment of high consequence robotics system

International Nuclear Information System (INIS)

Robinson, D.G.; Atcitty, C.B.

1996-01-01

This paper outlines the use of a failure modes and effects analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, the weigh and leak check system, is to replace a manual process for weight and leakage of nuclear materials at the DOE Pantex facility. Failure modes and effects analyses were completed for the robotics process to ensure that safety goals for the systems have been met. Due to the flexible nature of the robot configuration, traditional failure modes and effects analysis (FMEA) were not applicable. In addition, the primary focus of safety assessments of robotics systems has been the protection of personnel in the immediate area. In this application, the safety analysis must account for the sensitivities of the payload as well as traditional issues. A unique variation on the classical FMEA was developed that permits an organized and quite effective tool to be used to assure that safety was adequately considered during the development of the robotic system. The fundamental aspects of the approach are outlined in the paper

Chewing gum and context-dependent memory effects: a re-examination.

Science.gov (United States)

Miles, Christopher; Johnson, Andrew J

2007-03-01

Two experiments re-examined whether chewing spearmint gum affects initial word learning and/or immediate recall for a word list. Both experiments failed to show effects of chewing gum at learning or recall, nor did they suggest that chewing gum produces a context-dependent memory effect. This was true when extraneous contextual cues at learning and recall were minimised (Experiment 2). Together, the data are inconsistent with [Wilkinson, L., Scholey, A. & Wesnes, K. (2002). Chewing gum selectively improves aspects of memory in healthy volunteers. Appetite, 38, 235-236.] claim that chewing gum aids immediate recall of visually presented words. Our results are consistent with [Baker, J. R., Bezance, J. B., Zellaby, E. & Aggleton, J. P. (2004). Chewing gum can produce context-dependent effects upon memory. Appetite, 43, 207-210.] finding that chewing gum of itself is not a sufficient condition to provoke context-dependent learning with immediate testing.

Research on fuzzy comprehensive assessment method of nuclear power plant safety culture

International Nuclear Information System (INIS)

Xiang Yuanyuan; Chen Xukun; Xu Rongbin

2012-01-01

Considering the traits of safety culture in nuclear plant, 38 safety culture assessment indexes are established from 4 aspects such as safety values, safety institution, safety behavior and safety sub- stances. Based on it, a comprehensive assessment method for nuclear power plant safety culture is constructed by using AHP (Analytic Hierarchy Process) approach and fuzzy mathematics. The comprehensive assessment method has the quality of high precision and high operability, which can support the decision making of safety culture development. (authors)

Safety assessment for Generation IV nuclear systems

International Nuclear Information System (INIS)

Leahy, T.J.

2012-01-01

The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Recent RSWG work has focused on the definition of an integrated safety assessment methodology (ISAM) for evaluating the safety of Generation IV systems. ISAM is an integrated 'tool-kit' consisting of 5 analytical techniques that are available and matched to appropriate stages of Generation IV system concept development: 1) qualitative safety features review - QSR, 2) phenomena identification and ranking table - PIRT, 3) objective provision tree - OPT, 4) deterministic and phenomenological analyses - DPA, and 5) probabilistic safety analysis - PSA. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time

Technical Issues and Proposes on the Legislation of Probabilistic Safety Assessment in Periodic Safety Review

International Nuclear Information System (INIS)

Hwang, Seok-Won; Jeon, Ho-Jun; Na, Jang-Hwan

2015-01-01

Korean Nuclear Power Plants have performed a comprehensive safety assessment reflecting design and procedure changes and using the latest technology every 10 years. In Korea, safety factors of PSR are revised to 14 by revision of IAEA Safety Guidelines in 2003. In the revised safety guidelines, safety analysis field was subdivided into deterministic safety analysis, PSA (Probabilistic safety analysis), and hazard analysis. The purpose to examine PSA as a safety factor on PSR is to make sure that PSA results and assumptions reflect the latest state of NPPs, validate the level of computer codes and analytical models, and evaluate the adequacy of PSA instructions. In addition, its purpose is to derive the plant design change, operating experience of other plants and safety enhancement items as well. In Korea, PSA is introduced as a new factor. Thus, the overall guideline development and long-term implementation strategy are needed. Today in Korea, full-power PSA model revision and low-power and shutdown (LPSD) PSA model development is being performed as a part of the post Fukushima action items for operating plants. The scope of the full-power PSA is internal/external level 1, 2 PSA. But in case of fire PSA, the scope is level 1 PSA using new method, NUREG/CR-6850. In case of LPSD PSA, level 1 PSA for all operating plants, and level 2 PSA for 2 demonstration plants are under development. The result of the LPSD PSA will be used as major input data for plant specific SAMG (Severe Accident Management Guideline). The scope of PSA currently being developed in Korea cannot fulfill 'All Mode, All Scope' requirements recommended in the IAEA Safety Guidelines. Besides the legislation of PSA, step-by-step development strategy for non-performed scopes such as level 3 PSA and new fire PSA is one of the urgent issues in Korea. This paper suggests technical issues and development strategies for each PSA technical elements.

Nirex safety assessment research programme: 1987/88

International Nuclear Information System (INIS)

George, D.; Hodgkinson, D.P.

1987-01-01

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

Safety assessment requirements for onsite transfers of radioactive material

International Nuclear Information System (INIS)

Opperman, E.K.; Jackson, E.J.; Eggers, A.G.

1992-05-01

This document contains the requirements for developing a safety assessment document for an onsite package containing radioactive material. It also provides format and content guidance to establish uniformity in the safety assessment documentation and to ensure completeness of the information provided

Geosphere process report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Skagius, Kristina

2010-11-01

This report documents geosphere processes identified as relevant to the long-term safety of a KBS-3 repository, and forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

Geosphere process report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Skagius, Kristina (ed.) (Kemakta Konsult AB, Stockholm (Sweden))

2010-11-15

This report documents geosphere processes identified as relevant to the long-term safety of a KBS-3 repository, and forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

Research on advanced system safety assessment procedures (4)

International Nuclear Information System (INIS)

Suzuki, Kazuhiko; Shimada, Yukiyasu

2001-03-01

The past research reports in the area of safety engineering proposed the Computer-aided HAZOP system to be applied to Nuclear Reprocessing Facilities. Automated HAZOP system has great advantage compared with human analysts in terms of accuracy of the results, and time required to conduct HAZOP studies. This report surveys the literature on risk assessment and safety design based on the concept of independent protection layers (IPLs). Furthermore, to improve HAZOP System, tool is proposed to construct the basic model and the internal state model. Such HAZOP system is applied to analyze two kinds of processes, where the ability of the proposed system is verified. In addition, risk assessment support system is proposed to integrate safety design environment and assessment result to be used by other plants as well as to enable the underline plant to use other plants' information. This technique can be implemented using web-based safety information systems. (author)

Assessing safety culture using RADAR matrix

International Nuclear Information System (INIS)

Mariscal-Saldana, M. a.; Garcia-Herrero, S.; Toca-Otero, A.

2009-01-01

Santa Maria de Garona nuclear power plant, in collaboration with Burgos University, has proceeded to conduct a pilot project aimed at seeing the possibilities for the RADAR (Results, Approach, Development, Assessment and review) logic of EFQM model, as a tool for self evaluation of Safety Culture in a nuclear power plant. In the work it has sought evidences of Safety culture implanted in the plant, and identify strengths and areas for improvement regarding this Culture. the score obtained by analyzing these strengths and areas for improvements has served to prioritize actions implemented. The nuclear power plant has been submitted voluntarily to the mission SCART (Safety Culture Assessment Review Team), an international review being done for the first time in the world at a plant in operation and the team of experts led by International Agency of Atomic Energy (IAEA) has identified this project as a good practice, an innovative process implemented in the plant, that must be transmitted to other plants. (Author) 10 refs

Developing IAM for Life Cycle Safety Assessment

NARCIS (Netherlands)

Toxopeus, Marten E.; Lutters, Diederick; Nee, Andrew Y.C.; Song, Bin; Ong, Soh-Khim

2013-01-01

This publication discusses aspects of the development of an impact assessment method (IAM) for safety. Compared to the many existing IAM’s for environmentally oriented LCA, this method should translate the impact of a product life cycle on the subject of safety. Moreover, the method should be

Safety assessment of smoke flavouring primary products by the European Food Safety Authority

NARCIS (Netherlands)

Theobald, A.; Arcella, D.; Carere, A.; Croera, C.; Engel, K.H.; Gott, D.; Gurtler, R.; Meier, D.; Pratt, I.; Rietjens, I.M.C.M.; Simon, R.; Walker, R.

2012-01-01

This paper summarises the safety assessments of eleven smoke flavouring primary products evaluated by the European Food Safety Authority (EFSA). Data on chemical composition, content of polyaromatic hydrocarbons and results of genotoxicity tests and subchronic toxicity studies are presented and

Fire safety assessment of tunnel structures

DEFF Research Database (Denmark)

Gkoumas, Konstantinos; Giuliani, Luisa; Petrini, Francesco

2011-01-01

.g. structural and non structural, organizational, human behavior). This is even more truth for the fire safety design of such structures. Fire safety in tunnels is challenging because of the particular environment, bearing in mind also that a fire can occur in different phases of the tunnel’s lifecycle. Plans...... for upgrading fire safety provisions and tunnel management are also important for existing tunnels. In this study, following a brief introduction of issues regarding the above mentioned aspects, the structural performance of a steel rib for a tunnel infrastructure subject to fire is assessed by means...

Safety management system needs assessment.

Science.gov (United States)

2016-04-01

The safety of the traveling public is critical as each year there are approximately 200 highway fatalities in Nebraska and numerous crash injuries. The objective of this research was to conduct a needs assessment to identify the requirements of a sta...

Safety, Pharmacokinetics, Immunogenicity, and Biodistribution of (186)Re-Labeled Humanized Monoclonal Antibody BIWA 4 (Bivatuzumab( in Patients with Early-Stage Breast Cancer.

NARCIS (Netherlands)

Koppe, M.; Schaijk, F. van; Roos, J.C.; Leeuwen, P.; Heider, K.H.; Kuthan, H.; Bleichrodt, R.P.

2004-01-01

The aim of this prospective study was to evaluate the safety, pharmacokinetics, immunogenicity, and biodistribution of (186)Re-labeled humanized anti-CD44v6 monoclonal antibody (MAb( BIWA 4 (Bivatuzumab( in 9 patients with early-stage breast cancer. Radioimmunoscintigraphy (RIS( was performed within

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.

Armenian nuclear power plant: US NRC assistance programme for seismic upgrade and safety analysis

International Nuclear Information System (INIS)

Simos, N.; Perkins, K.; Jo, J.; Carew, J.; Ramsey, J.

2003-01-01

This paper summarizes the U.S. Nuclear Regulatory Commission's (US NRC) technical support program activities associated with the Armenian Nuclear Power Plant (ANPP) safety upgrade. The US NRC program, integrated within the overall IAEA-led initiative for safety re-evaluation of the WWER plants, has as its main thrust the technical support to the Armenian Nuclear Regulatory Authority (ANRA) through close collaboration with the scientific staff at Brookhaven National Laboratory (BNL). Several major technical areas of support to ANRA form the basis of the NRC program. These include the seismic re-evaluation and upgrade of the ANPP, safety evaluation of critical systems, and the generation of the Safety Analysis Report (SAR). Specifically, the seismic re-evaluation of the ANPP is part of a broader activity that involves the re-assessment of the seismic hazard at the site, the identification of the Safe Shutdown Equipment at the plant and the evaluation of their seismic capacity, the detailed modeling and analysis of the critical facilities at ANPP, and the generation of the Floor Response Spectra (FRS). Based on the new spectra that incorporate all new findings (hazard, site soil, structure, etc.), the overall capacity of the main structures and the seismic capacity of the critical systems are being re-evaluated. In addition, analyses of critical safe shutdown systems and safe shutdown processes are being performed to ensure both the capabilities of the operating systems and the enhancement of safety due to system upgrades. At present, one of the principal goals of the US NRC's regulatory assistance activities with ANRA is enhancing ANRA's regulatory oversight of high-priority safety issues (both generic and plant-specific) associated with operation of the ANPP. As such, assisting ANRA in understanding and assessing plant-specific seismic and other safety issues associated with the ANPP is a high priority given the ANPP's being located in a seismically active area

Safety and security risk assessments--now demystified!

Science.gov (United States)

White, Donald E

2011-01-01

Safety/security risk assessments no longer need to spook nor baffle healthcare safety/security managers. This grid template provides at-at-glance quick lookup of the possible threats, the affected people and things, a priority ranking of these risks, and a workable solution for each risk. Using the standard document, spreadsheet, or graphics software already available on your computer, you can easily use a scientific method to produce professional looking risk assessments that get quickly understood by both senior managers and first responders alike!

Visualization of Safety Assessment Result Using GIS in SITES

International Nuclear Information System (INIS)

Yun, Bong-Yo; Park, Joo Wan; Park, Se-Moon; Kim, Chang-Lak

2006-01-01

Site Information and Total Environmental database management System (SITES) is an integrated program for overall data analysis, environmental monitoring, and safety analysis that are produced from the site investigation and environmental assessment of the relevant nuclear facility. SITES is composed of three main modules such as Site Environment Characterization database for Unified and Reliable Evaluation system (SECURE), Safety Assessment INTegration system (SAINT) and Site Useful Data Analysis and ALarm system (SUDAL). The visualization function of safety assessment and environmental monitoring results is designed. This paper is to introduce the visualization design method using Geographic Information System (GIS) for SITES

Performance Assessment and analysis of national building codes with fire safety in all wards of a hospital

Directory of Open Access Journals (Sweden)

M. Mahdinia

2009-04-01

Full Text Available Background and aimsAIDS as a re-emergent disease and Viral hepatitis (B and C as one of thBackground and objective: Fire safety is an important problem in hospitals. Movement less, lack of awareness and special situation of residents are the reasons of this subject. In more countries such as Iran, fire protection regulations have compiled within the framework of national building codes. Current building codes don't create sufficient safety for patient in the hospitals in different situations and more of the advanced countries in the world effort to establish building code, base  on performance. This study to be accomplished with this goal that determination of fire risk level in the wards of a hospital and survey the efficiency of the national building codes. Methodsfire risk assesses is done, using "engineering fire risk assessment method" with the checklist for Data gathering. In this manner, risk calculate in all compartments and in the next  stage for survey the effect of building codes, with this supposition that all compartment is  conforming to building code requirement, risk level calculate in two situation.Resultsthe results of present study reveals that, risk level in all wards is more than one and even though risk less than one is acceptable, consequently minimum of safely situations didn't  produce in most wards. The results show the national building code in the different conditions  don't have appropriate efficient for creation of suitable safety. Conclusionin order to access to a fire safety design with sufficient efficiency, suitable selection is use of risk assessment based on, design methods.

Assessing community child passenger safety efforts in three Northwest Tribes.

Science.gov (United States)

Smith, M L; Berger, L R

2002-12-01

To identify strengths and weaknesses in community based child passenger safety programs by developing a scoring instrument and conducting observations of child restraint use in three Native American communities. The three communities are autonomous Tribal reservations in the Pacific Northwest. Their per capita incomes and rates of unemployment are comparable. In each community, 100 children under 5 years old were observed for car seat use. A six item community assessment tool (100 points maximum) awarded points for such items as the type (primary or secondary) and enforcement of child restraint laws; availability of car seats from distribution programs; extent of educational programs; and access to data on vehicle injuries. For children from birth to 4 years, the car seat use rate ranged from 12%-21%. Rates for infants (71%-80%) far exceeded rates for 1-4 year old children (5%-14%). Community scores ranged from 0 to 31.5 points. There was no correlation between scores and observed car seat use. One reason was the total lack of enforcement of restraint laws. A community assessment tool can highlight weaknesses in child passenger efforts. Linking such a tool with an objective measure of impact can be applied to other injury problems, such as fire safety or domestic violence. The very process of creating and implementing a community assessment can enhance agency collaboration and publicize evidence based "best practices" for injury prevention. Further study is needed to address methodologic issues and to examine crash and medical data in relation to community child passenger safety scores.

Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

International Nuclear Information System (INIS)

Frid, W.; Sandervaag, O.

1997-01-01

The Swedish regulatory approach to safety assessment and severe accident management is briefly described. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis, and analysis of postulated disturbances and accident progression sequences. Management and man-technology-organisation issues, as well as inspections, play a key role in safety assessment. Basis for severe accident management were established by the Government decisions in 1981 and 1986. By the end of 1988, the severe accident mitigation systems and emergency operating procedures were implemented at all Swedish reactors. The severe accident research has continued after 1988 for further verification of the protection provided by the systems and reduction of remaining uncertainties in risk dominant phenomena

Complementary assessment of the safety of French nuclear power plants

International Nuclear Information System (INIS)

Camarcat, N.; Pouget-Abadie, X.

2011-01-01

As an immediate consequence of the Fukushima accident the French nuclear safety Authority (ASN) asked EDF to perform a complementary safety assessment for each nuclear power plant dealing with 3 points: 1) the consequences of exceptional natural disasters, 2) the consequences of total loss of electrical power, and 3) the management of emergency situations. The safety margin has to be assessed considering 3 main points: first a review of the conformity to the initial safety requirements, secondly the resistance to events overdoing what the facility was designed to stand for, and the feasibility of any modification susceptible to improve the safety of the facility. This article details the specifications of such assessment, the methodology followed by EDF, the task organization and the time schedule. (A.C.)

Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

Assessment of safety regulation using an artificial society

International Nuclear Information System (INIS)

Furuta, Kazuo; Nagase, Masaya

2005-01-01

This study proposes using an artificial society to assess impacts of safety regulation on the society. The artificial society used in this study is a multi-agent system, which consists of many agents representing companies. The agents cannot survive unless they get profits by producing some products. Safety regulation functions as the business environment, which the agents will evolve to fit to. We modeled this process of survival and adaptation by the genetic algorithm. Using the proposed model, case simulations were performed to compare various regulation styles, and some interesting insights were obtained how regulation style influences behavior of the agents and then productivity and safety level of the industry. In conclusion, an effective method for assessment of safety regulation has been developed, and then several insights were shown in this study

What's gender got to do with it? Examining masculinities, health and safety and return to work in male dominated skilled trades.

Science.gov (United States)

Stergiou-Kita, Mary; Mansfield, Elizabeth; Colantonio, Angela; Moody, Joel; Mantis, Steve

2016-06-16

Electrical injuries are a common cause of work-related injury in male dominated skilled trades. In this study we explored how issues of gender, masculinities and institutional workplace practices shape expectations of men and their choices when returning to work following a workplace electrical injury. Twelve workers, who suffered an electrical injury, and twelve employer representatives, completed semi-structured interviews. Using thematic analysis we identified key themes related to how masculinities influenced men's health and safety during the return to work process. Strong identification with worker roles can influence injured workers decisions to return to work 'too early'. A desire to be viewed as a strong, responsible, resilient worker may intersect with concerns about job loss, to influence participants' decisions to not report safety issues and workplace accidents, to not disclose post-injury work challenges, and to not request workplace supports. Institutionalized workplace beliefs regarding risk, de-legitimization of the severity of injuries, and the valorization of the "tough" worker can further re-enforce dominant masculine norms and influence return to work processes and health and safety practices. Workplaces are key sites where gender identities are constructed, affirmed and institutionalized. Further research is warranted to examine how established masculine norms and gendered workplace expectations can influence workplace health and safety in male dominated high risk occupations. Future research should also evaluate strategies that encourage men to discuss post-injury work challenges and request supports when work performance or health and safety issues arise during the return to work process.

Integrated Deterministic-Probabilistic Safety Assessment Methodologies

Energy Technology Data Exchange (ETDEWEB)

Kudinov, P.; Vorobyev, Y.; Sanchez-Perea, M.; Queral, C.; Jimenez Varas, G.; Rebollo, M. J.; Mena, L.; Gomez-Magin, J.

2014-02-01

IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) is a family of methods which use tightly coupled probabilistic and deterministic approaches to address respective sources of uncertainties, enabling Risk informed decision making in a consistent manner. The starting point of the IDPSA framework is that safety justification must be based on the coupling of deterministic (consequences) and probabilistic (frequency) considerations to address the mutual interactions between stochastic disturbances (e.g. failures of the equipment, human actions, stochastic physical phenomena) and deterministic response of the plant (i.e. transients). This paper gives a general overview of some IDPSA methods as well as some possible applications to PWR safety analyses. (Author)

Human reliability in probabilistic safety assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

Nowadays a growing interest in medioambiental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processess and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects. (This relevance has been demostrated in the accidents happenned). However in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a guide to carry out a Human Reliability Analysis and c) a selected overwiev of the techniques and methodologies currently applied in this area. (Author)

Probabilistic assessment of NPP safety under aircraft impact

International Nuclear Information System (INIS)

Birbraer, A.N.; Roleder, A.J.; Arhipov, S.B.

1999-01-01

Methodology of probabilistic assessment of NPP safety under aircraft impact is described below. The assessment is made taking into account not only the fact of aircraft fall onto the NPP building, but another casual parameters too, namely an aircraft class, velocity and mass, as well as point and angle of its impact with the building structure. This analysis can permit to justify the decrease of the required structure strength and dynamic loads on the NPP equipment. It can also be especially useful when assessing the safety of existing NPP. (author)

Nirex Safety Assessment Research Programme bibliography, 1990

International Nuclear Information System (INIS)

Cooper, M.J.

1990-10-01

This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research Programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW) and associated radiological assessments. (author)

Understanding and assessing safety culture

International Nuclear Information System (INIS)

Dalling, Ian

1997-01-01

The 'Dalling' integrated model of organisational performance is introduced and described. A principal element of this model is culture, which is dynamically contrasted with the five other interacting critical elements, which comprise: the management system, the knowledge base, corporate leadership, stakeholders and consciousness. All six of these principal driving elements significantly influence health, safety, environmental, security, or any other aspect of organisational performance. It is asserted that the elements of organisational performance must be clearly defined and understood if meaningful measurements are to be carried out and sustained progress made in improving the knowledge of organisational performance. AEA Technology's safety culture research programme is then described together with the application of a safety culture assessment tool to organisations in the nuclear, electricity, transport, and oil and gas industries, both within and outside of the United Kingdom. (author)

Seismic safety of nuclear power plants

International Nuclear Information System (INIS)

Guerpinar, A.; Godoy, A.

2001-01-01

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

An assessment of food hygiene and safety at farmers' markets.

Science.gov (United States)

Worsfold, D; Worsfold, P M; Griffith, C J

2004-04-01

Farmers' markets are becoming a more significant part of the food-retailing sector. A survey of farmers' markets was conducted to assess aspects of food hygiene and safety. The views of the public using the markets were also examined. The range of farm products was wide and the methods utilised varied. The markets were usually temporary outdoor events with few facilities. Traders had received elementary food hygiene training and rated their hygiene standards highly. Less than half had risk management procedures in place, most did not perceive their produce as high-risk. They believed consumers to be mainly interested in food quality and to regard food safety issues highly. Consumers shopped at the markets because of the quality of the products sold. Their overall satisfaction with the markets was high and they raised no concerns about food safety. Given the restricted facilities at farmers' markets and the early phase of implementation of hygiene management systems by market traders, it may be precautionary to restrict the sale of farm products at farmers markets to those that are regarded as low-risk.

Simulation of the ATV Re-Entry Obsrvations

Science.gov (United States)

Bastida Virgili, B.; Krag, H.; Lips, T.; De Pasquale, E.

2010-09-01

The first ATV was launched on 9th March 2008 and, after a successful mission, the last phase was a controlled destructive re-entry on 29th September 2008, shortly after 13:30 UTC, in which the remains of the ATV and its load fell into the South Pacific Ocean. In order to better understand the re-entry processes, an insitu optical observation campaign was launched to record and analyze the ATV controlled re-entry with several instruments on board of two airplanes and also from the ISS. This observation campaign was successful and triggered several different still-ongoing studies on the extraction and analysis of data to draw conclusions on the adequacy of the re-entry break-up and explosion models used for the safety analysis of the ATV re-entry. This paper addresses the validation process for ESA’s model for re-entry survivability and on-ground risk assessment for explosive re-entry events using the observation data. The underlying rationale is to improve the models for the benefit of planning and execution of future controlled re-entries and in risk calculation in case of uncontrolled ones. The re-entry trajectory of the ATV, the explosive event and the trajectories of the fragments are simulated with the existing ESA tools and the EVOLVE explosion model. Additional software has been developed to simulate airborne sensor field of view(FOV) crossings based on the aircraft trajectories, attitude profile, sensor mounts and FOVs. Sensor performance and object radiation are modeled in order to generate synthetic images for the different sensors in the ISS and the two airplanes. These synthetic images and synthetic videos are compared with the available reentry observations of the ATV. This paper will present the software and techniques to generate synthetic imagery. It will give results of the comparison between the simulated and the real trajectories and fragmentation and explain the subsequent validation process of the ESA re-entry tools and the potential

The use of probabilistic safety assessments for improving nuclear safety in Europe

International Nuclear Information System (INIS)

Birkhofer, A.

1992-01-01

The political changes in Europe broadened the scope of international nuclear safety matters considerably. The Western world started to receive reliable and increasingly detailed information on Eastern European nuclear technology and took note of a broad range of technical and administrative problems relevant for nuclear safety in these countries. Reunification made Germany a focus of information exchange on these matters. Here, cooperation with the former German Democratic Republic and with other Eastern European countries as well as safety analyses of Soviet-built nuclear power plants started rather early. Meanwhile, these activities are progressing toward all-European cooperation in the nuclear safety sector. This cooperation includes the use of probabilistic safety assessments (PSAs) addressing applications in both Western and Eastern Europe as well as the further development of this methodology in a converging Europe

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Savannah River Site K-Reactor Probabilistic Safety Assessment

International Nuclear Information System (INIS)

Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O'Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

1992-12-01

This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

A novel safety assessment strategy applied to non-selective extracts.

Science.gov (United States)

Koster, Sander; Leeman, Winfried; Verheij, Elwin; Dutman, Ellen; van Stee, Leo; Nielsen, Lene Munch; Ronsmans, Stefan; Noteborn, Hub; Krul, Lisette

2015-06-01

A main challenge in food safety research is to demonstrate that processing of foodstuffs does not lead to the formation of substances for which the safety upon consumption might be questioned. This is especially so since food is a complex matrix in which the analytical detection of substances, and consequent risk assessment thereof, is difficult to determine. Here, a pragmatic novel safety assessment strategy is applied to the production of non-selective extracts (NSEs), used for different purposes in food such as for colouring purposes, which are complex food mixtures prepared from reference juices. The Complex Mixture Safety Assessment Strategy (CoMSAS) is an exposure driven approach enabling to efficiently assess the safety of the NSE by focussing on newly formed substances or substances that may increase in exposure during the processing of the NSE. CoMSAS enables to distinguish toxicologically relevant from toxicologically less relevant substances, when related to their respective levels of exposure. This will reduce the amount of work needed for identification, characterisation and safety assessment of unknown substances detected at low concentration, without the need for toxicity testing using animal studies. In this paper, the CoMSAS approach has been applied for elderberry and pumpkin NSEs used for food colouring purposes. Copyright © 2015 Elsevier Ltd. All rights reserved.

Japan: The institute for the economy of energy recommends a quick re-start of nuclear reactors

International Nuclear Information System (INIS)

Anon.

2014-01-01

The Japanese Institute for the Economy of the Energy (IEEJ) considers that the sooner the nuclear reactors will re-start, the better the Japanese economy and environment will be. The 48 Japanese reactors were stopped after the Fukushima accident and their restart is linked to the implementation of new measures for reinforcing safety. Until now only 2 reactors Sendai 1 and Sendai 2 have been allowed to re-start. The procedure for the safety assessment of the reactors is slower than expected. A study shows that only 7 reactors may be allowed to re-start before march 2015 and a total of 19 units may be operating in march 2016. In this scenario 2% of the electricity will come from nuclear energy in 2014 and 15% in 2015, natural gas imports will still be necessary for the production of electricity and their global cost is estimated to reach 56 billions euros while Japan's rate of energy independence will drop by 4.6%. (A.C.)

Suggestions on the Development of Safety Culture Assessment Method

International Nuclear Information System (INIS)

Choi, Young Sung; Choi, Kwang Sik; Kim, Woong Sik

2006-01-01

Several efforts have been made to assess safety culture of organization that operates nuclear power plants in Korea. The MOST and KINS played a major role to develop assessment methods and KHNP applied them to its NPPs. This paper explains the two methods developed by KINS briefly and presents the insights obtained from the two different applications. It concludes with some suggestions for safety culture assessment based on the insights

Assessing ultrasonic examination results

International Nuclear Information System (INIS)

Deutsch, V.; Vogt, M.

1977-01-01

Amongst nondestructive examination methods, the ultrasonic examination plays an important role. The reason why its scope of application is so wide is because the sound conducting capacity is the only property the material of a test specimen has to have. As the fields are so manifold, only main aspects can be described briefly. The list of references, however, is very extensive and gives plenty of information of all the problems concerning the assessment of ultrasonic examination results. (orig./RW) [de

Risks and safety aspects related to PET/MR examinations

International Nuclear Information System (INIS)

Brix, Gunnar; Nekolla, Elke A.; Nosske, Dietmar; Griebel, Juergen

2009-01-01

The introduction of positron emission tomography (PET)/magnetic resonance (MR) systems into medical practice in the foreseeable future may not only lead to a gain in clinical diagnosis compared to PET/computed tomography (CT) imaging due to the superior soft-tissue contrast of the MR technology but can also substantially reduce exposure of patients to ionizing radiation. On the other hand, there are also risks and health effects associated with the use of diagnostic MR devices that have to be considered carefully. This review article summarizes biophysical and biological aspects, which are of relevance for the assessment of health effects related to the exposure of patients to both ionizing radiation in PET and magnetic and electromagnetic fields in MR. On this basis, some considerations concerning the justification and optimization of PET/MR examinations are presented - as far as this is possible at this very early stage. Current safety standards do not take into account synergistic effects of ionizing radiation and magnetic and electromagnetic fields. In the light of the developing PET/MR technology, there is an urgent need to investigate this aspect in more detail for exposure levels that will occur at PET/MR systems. (orig.)

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

Fessenheim 2: ASN's green light for continuing operation - Beginning of the works for unit 1

International Nuclear Information System (INIS)

Anon.

2013-01-01

Every 10 years a nuclear power plant operator has to make a re-assessment of the nuclear safety standard of his plant. This re-assessment is made of 2 parts: first the review of the safety conformity and secondly a thorough re-examination of the safety that takes into account today's safety standards and feedback experience from similar plants. This detailed assessment of the safety aims at checking that the consequences of the different aging phenomena are well mastered for the next 10 years at least. At the end of this re-assessment, the ASN (French Nuclear Safety Authorities) decide or not the continuation of plant activity or can prescribe safety improvements. In the case of the unit 2 of the Fessenheim plant that has just finished its third decennial safety re-assessment, the ASN has prescribed the same improvements as for the unit 1, that is to say the reinforcement of the resistance to corium of the foundation raft and an improvement on the emergency cooling system. The works on the unit 1 have begun despite contestation from anti-nuclear associations that question the cost of the safety upgrading (20 to 30 millions euros) while the unit is expected to be decommissioned by end 2016. (A.C.)

A large-scale examination of the effectiveness of anonymous marking in reducing group performance differences in higher education assessment.

Directory of Open Access Journals (Sweden)

Daniel P Hinton

Full Text Available The present research aims to more fully explore the issues of performance differences in higher education assessment, particularly in the context of a common measure taken to address them. The rationale for the study is that, while performance differences in written examinations are relatively well researched, few studies have examined the efficacy of anonymous marking in reducing these performance differences, particularly in modern student populations. By examining a large archive (N = 30674 of assessment data spanning a twelve-year period, the relationship between assessment marks and factors such as ethnic group, gender and socio-environmental background was investigated. In particular, analysis focused on the impact that the implementation of anonymous marking for assessment of written examinations and coursework has had on the magnitude of mean score differences between demographic groups of students. While group differences were found to be pervasive in higher education assessment, these differences were observed to be relatively small in practical terms. Further, it appears that the introduction of anonymous marking has had a negligible effect in reducing them. The implications of these results are discussed, focusing on two issues, firstly a defence of examinations as a fair and legitimate form of assessment in Higher Education, and, secondly, a call for the re-examination of the efficacy of anonymous marking in reducing group performance differences.

A large-scale examination of the effectiveness of anonymous marking in reducing group performance differences in higher education assessment.

Science.gov (United States)

Hinton, Daniel P; Higson, Helen

2017-01-01

The present research aims to more fully explore the issues of performance differences in higher education assessment, particularly in the context of a common measure taken to address them. The rationale for the study is that, while performance differences in written examinations are relatively well researched, few studies have examined the efficacy of anonymous marking in reducing these performance differences, particularly in modern student populations. By examining a large archive (N = 30674) of assessment data spanning a twelve-year period, the relationship between assessment marks and factors such as ethnic group, gender and socio-environmental background was investigated. In particular, analysis focused on the impact that the implementation of anonymous marking for assessment of written examinations and coursework has had on the magnitude of mean score differences between demographic groups of students. While group differences were found to be pervasive in higher education assessment, these differences were observed to be relatively small in practical terms. Further, it appears that the introduction of anonymous marking has had a negligible effect in reducing them. The implications of these results are discussed, focusing on two issues, firstly a defence of examinations as a fair and legitimate form of assessment in Higher Education, and, secondly, a call for the re-examination of the efficacy of anonymous marking in reducing group performance differences.

Knowledge representation in safety assessment: improving transparency and traceability

Energy Technology Data Exchange (ETDEWEB)

Lemos, F.L. de [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Sullivan, T. [Brookhaven National Laboratory (BNL), Upton, NY (United States); Ross, T. [University of New Mexico (UNM), Albuquerque, NM (United States); Guimaraes, L.N.F. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)

2011-07-01

Transparency and traceability are key factors for confidence building, acceptability, and quality enhancement of the safety assessment, and safety case for a radioactive waste disposal facility. In order to facilitate analysis and promote discussions, all of the information used to make decisions should be readily available to stake holders. The information should convey a good understanding of the intermediate decisions processes, allowing examination of alternatives and 'what if questions'. In an ideal situation all stake holders, including scientists and the public, should be able to follow the path of a certain parameter, from the beginning where it was defined, its assumptions and uncertainties, throughout the calculations until the final results of the safety assessment. One of the main challenges, to achieving such a transparency and traceability, is that stake holders are a very diverse audience, with very different backgrounds. This could require preparation of various versions of the same documentation, which would be impractical. While the linguistic information is of crucial importance to understanding the reasoning, it is very difficult to convey the supporting conditions, and consequent uncertainties for the selection of parameters values. Even scientists involved in the process can become confused due to the overwhelming amount of information that is used to support parameter value selection. The amount of details makes it difficult to track the decisions, which lead to the selection of a certain parameter, throughout the calculations. This paper presents a methodology to represent the linguistic information used in the safety assessment in terms of mathematical expressions by using the fuzzy sets and fuzzy logic tools. This methodology aims to help information to be readily available while keeping, as much as possible, the original meaning of the linguistic expressions and, consequently, to be available at any time as a quick reference

Knowledge representation in safety assessment: improving transparency and traceability

International Nuclear Information System (INIS)

Lemos, F.L. de; Sullivan, T.; Ross, T.; Guimaraes, L.N.F.

2011-01-01

Transparency and traceability are key factors for confidence building, acceptability, and quality enhancement of the safety assessment, and safety case for a radioactive waste disposal facility. In order to facilitate analysis and promote discussions, all of the information used to make decisions should be readily available to stake holders. The information should convey a good understanding of the intermediate decisions processes, allowing examination of alternatives and 'what if questions'. In an ideal situation all stake holders, including scientists and the public, should be able to follow the path of a certain parameter, from the beginning where it was defined, its assumptions and uncertainties, throughout the calculations until the final results of the safety assessment. One of the main challenges, to achieving such a transparency and traceability, is that stake holders are a very diverse audience, with very different backgrounds. This could require preparation of various versions of the same documentation, which would be impractical. While the linguistic information is of crucial importance to understanding the reasoning, it is very difficult to convey the supporting conditions, and consequent uncertainties for the selection of parameters values. Even scientists involved in the process can become confused due to the overwhelming amount of information that is used to support parameter value selection. The amount of details makes it difficult to track the decisions, which lead to the selection of a certain parameter, throughout the calculations. This paper presents a methodology to represent the linguistic information used in the safety assessment in terms of mathematical expressions by using the fuzzy sets and fuzzy logic tools. This methodology aims to help information to be readily available while keeping, as much as possible, the original meaning of the linguistic expressions and, consequently, to be available at any time as a quick reference. This would

Operational safety review programmes for nuclear power plants. Guidelines for assessment

International Nuclear Information System (INIS)

2002-01-01

The IAEA has been offering the Operational Safety Review Team (OSART) programme to provide advice and assistance to Member States in enhancing the operational safety of nuclear power plants (NPPs). Simultaneously, the IAEA has encouraged self-assessment and review by Member States of their own nuclear power plants to continuously improve nuclear safety. Currently, some utilities have been implementing safety review programmes to independently review their own plants. Corporate or national operational safety review programmes may be compliance or performance based. Successful utilities have found that both techniques are necessary to provide assurance that (i) as a minimum the NPP meets specific corporate and legal requirements and (ii) management at the NPP is encouraged to pursue continuous improvement principles. These programmes can bring nuclear safety benefits to the plants and utilities. The IAEA has conducted two pilot missions to assess the effectiveness of the operational review programme. Based on these missions and on the experience gained during OSART missions, this document has been developed to provide guidance on and broaden national/corporate safety review programmes in Member States, and to assist in maximizing their benefits. These guidelines are intended primarily for the IAEA team to conduct assessment of a national/corporate safety review programme. However, this report may also be used by a country or utility to establish its own national/corporate safety review programme. The guidelines may likewise be used for self-assessment or for establishing a baseline when benchmarking other safety review programmes. This report consists of four parts. Section 2 addresses the planning and preparation of an IAEA assessment mission and Sections 3 and 4 deal with specific guidelines for conducting the assessment mission itself

[Quality assessment of microscopic examination in tuberculosis diagnostic laboratories: a preliminary study].

Science.gov (United States)

Simşek, Hülya; Ceyhan, Ismail; Tarhan, Gülnur; Güner, Uğur

2010-10-01

Recently, the diagnosis of pulmonary tuberculosis (TB) has based on smear microscopy in the Direct Observed Treatment Strategy (DOTS) programme which provides the basis of treatment worldwide. Microscopic detection of AFB (Acid-Fast Bacilli) is one of the main components in the National TB Control Programmes (NTCP). Precision level in microscopy procedures and evaluations are the most important steps for accurate diagnosis of the disease and to initiate proper treatment. Therefore, the external quality assessment (EQA) is the most important implement to provide the reliability and validity of tests. In countries where NTCP are performed, this task is fulfilled by the National Reference Laboratories (NRL) according to the guidelines of the World Health Organization (WHO). For this purpose a pilot study was initiated by the central NRL of Turkey for EQA of AFB smear microscopy as part of the NTCP on January 1, 2005. A total of 5 laboratories of which 2 were district TB laboratories (A, B), 2 were tuberculosis control dispensaries (C, D), 1 was a national reference laboratory (E), participated in this study. Blind re-checking method (re-examination of randomly selected slides) was used for the evaluation, and the slides were sent to the central NRL with 3 months interval, four times a year, selected according to LQAS (Lot Quality Assurance Sampling) guides. In the re-evaluation of the slides, false positivity (FP), false negativity (FN) and quantification errors (QE) were noted. Laboratory A, sent totally 525 slides between January 1, 2005 and April 1, 2008. In the result of re-checking, 514 (97.9%) slides were found concordant, and 11 (2.1%) were discordant (10 FP, 1 FN). Laboratory B, participated in the study between October 1, 2005 and July 1, 2006 and of the 67 re-examined slides, 60 (89.5%) were concordant and 7 (10.5%) were discordant (2 FP, 0 FN, 5 QE). Laboratory C, sent 235 slides between January 1, 2005 and April 1, 2006; of them 218 (92.8%) were detected

Flamanville 3 EPR, Safety Assessment and On-site Inspections

International Nuclear Information System (INIS)

Piedagnel, Corinne; Tarallo, Francois; Monnot, Bernard

2011-01-01

As a Technical Support Organisation of the French Safety Authority (ASN), the IRSN carries out the safety assessment of EPR project design and participates in the ASN inspections performed at the construction site and in factories. The design assessment consists in defining the safety functions which should be ensured by civil structures, evaluating the EPR Technical Code for Civil works (ETC-C) in which EdF has defined design criteria and construction rules, and carrying out a detailed assessment of a selection of safety-related structures. Those detailed assessments do not consist of a technical control but of an analysis whose objectives are to ensure that design and demonstrations are robust, in accordance with safety and regulatory rules. Most assessments led IRSN to ask EdF to provide additional justification sometimes involving significant modifications. In the light of those complementary justifications and modifications, IRSN concluded that assessments carried out on design studies were globally satisfactory. The participation of IRSN to the on-site inspections led by ASN is a part of the global control of the compliance of the reactor with its safety objectives. For that purpose IRSN has defined a methodology and an inspection program intended to ASN: based on safety functions associated with civil works (confinement and resistance to aggressions), the corresponding behaviour requirements are identified and linked to a list of main civil works elements. During the inspections, deviations to the project's technical specifications or to the rules of the art were pointed out by IRSN. Those deviations cover various items, such as concrete fabrication, concrete pouring methodology, lack of reinforcement in some structures, unadapted welding procedures of the containment leak-tight steel liner and unsatisfactory treatment of concreting joints. The analysis of those problems has revealed flaws in the organisation of the contractors teams together with an

Prospects for probabilistic safety assessment

International Nuclear Information System (INIS)

Hirschberg, S.

1992-01-01

This article provides some reflections on future developments of Probabilistic Safety Assessment (PSA) in view of the present state of the art and evaluates current trends in the use of PSA for safety management. The main emphasis is on Level 1 PSA, although Level 2 aspects are also highlighted to some extent. As a starting point, the role of PSA is outlined from a historical perspective, demonstrating the rapid expansion of the uses of PSA. In this context the wide spectrum of PSA applications and the associated benefits to the users are in focus. It should be kept in mind, however, that PSA, in spite of its merits, is not a self-standing safety tool. It complements deterministic analysis and thus improves understanding and facilitating prioritization of safety issues. Significant progress in handling PSA limitations - such as reliability data, common-cause failures, human interactions, external events, accident progression, containment performance, and source-term issues - is described. This forms a background for expected future developments of PSA. Among the most important issues on the agenda for the future are PSA scope extensions, methodological improvements and computer code advancements, and full exploitation of the potential benefits of applications to operational safety management. Many PSA uses, if properly exercised, lead to safety improvements as well as major burden reductions. The article provides, in addition, International Atomic Energy Agency (IAEA) perspective on the topics covered, as reflected in the current PSA programs of the agency. 74 refs., 6 figs., 1 tab

Examination of cadmium safety rod thermal test specimens and failure mechanism evaluation

International Nuclear Information System (INIS)

Thomas, J.K.; Peacock, H.B.; Iyer, N.C.

1992-01-01

The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of a hypothetical LOCA event. Accordingly, an experimental cadmium safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. Companion reports describe the experiments and a structural evaluation (finite element analysis) of the safety rod. This report deals primarily with the examination of the test specimens, evaluation of possible failure mechanisms, and confirmatory separate effects experiments. It is concluded that the failures observed in the cadmium safety rod thermal tests which occurred at low temperature (T 800 degrees C) with fast thermal ramp rates are concluded to be mechanical in nature without significant environmental degradation. Based on these tests, tasks were initiated to design and manufacture B 4 C safety rods to replace the cadmium safety rods. The B 4 C safety rods have been manufactured at this time and it is currently planned to charge them to the reactor in the near future. 60 refs

The practice of pre-marketing safety assessment in drug development.

Science.gov (United States)

Chuang-Stein, Christy; Xia, H Amy

2013-01-01

The last 15 years have seen a substantial increase in efforts devoted to safety assessment by statisticians in the pharmaceutical industry. While some of these efforts were driven by regulations and public demand for safer products, much of the motivation came from the realization that there is a strong need for a systematic approach to safety planning, evaluation, and reporting at the program level throughout the drug development life cycle. An efficient process can help us identify safety signals early and afford us the opportunity to develop effective risk minimization plan early in the development cycle. This awareness has led many pharmaceutical sponsors to set up internal systems and structures to effectively conduct safety assessment at all levels (patient, study, and program). In addition to process, tools have emerged that are designed to enhance data review and pattern recognition. In this paper, we describe advancements in the practice of safety assessment during the premarketing phase of drug development. In particular, we share examples of safety assessment practice at our respective companies, some of which are based on recommendations from industry-initiated working groups on best practice in recent years.

Assessing patient safety culture in hospitals across countries

NARCIS (Netherlands)

Wagner, C.; Smits, M.; Sorra, J.; Huang, C.C.

2013-01-01

Objective. It is believed that in order to reduce the number of adverse events, hospitals have to stimulate a more open culture and reflective attitude towards errors and patient safety. The objective is to examine similarities and differences in hospital patient safety culture in three countries:

Assessing patient safety culture in hospitals across countries.

NARCIS (Netherlands)

Wagner, C.; Smits, M.; Sorra, J.; Huang, C.C.

2013-01-01

Objective: It is believed that in order to reduce the number of adverse events, hospitals have to stimulate a more open culture and reflective attitude towards errors and patient safety. The objective is to examine similarities and differences in hospital patient safety culture in three countries:

Assessing patient safety culture in hospitals across countries

NARCIS (Netherlands)

Wagner, C.; Smits, M.; Sorra, J.; Huang, C.C.

2013-01-01

OBJECTIVE: It is believed that in order to reduce the number of adverse events, hospitals have to stimulate a more open culture and reflective attitude towards errors and patient safety. The objective is to examine similarities and differences in hospital patient safety culture in three countries:

Human Reliability in Probabilistic Safety Assessments

International Nuclear Information System (INIS)

Nunez Mendez, J.

1989-01-01

Nowadays a growing interest in environmental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processes and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects (This relevance has been demonstrated in the accidents happened) . However, in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a quid to carry out a Human Reliability Analysis and c) a selected overview of the techniques and methodologies currently applied in this area. (Author) 20 refs

Interim process report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Sellin, Patrick

2004-08-01

This report is a documentation of buffer processes identified as relevant to the long-term safety of a KBS-3 repository. The report is part of the interim reporting of the safety assessment SR-Can, see further the Interim main report. The final SR-Can reporting will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of this report is to document the scientific knowledge of the processes to a level required for an adequate treatment in the safety assessment. The documentation is thus from a scientific point of not exhaustive since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. The purpose is further to determine the handling of each process in the safety assessment and to demonstrate how uncertainties are taken care of, given the suggested handling. The process documentation in the SR 97 version of the Process report is a starting point for this SR-Can interim version. As further described in the Interim main report, the list of relevant processes has been reviewed and slightly extended by comparison to other databases. Furthermore, the backfill has been included as a system part of its own, rather than being described together with the buffer as in SR 97. Apart from giving an interim account of the documentation and handling of buffer processes in SR-Can, this report is meant to serve as a template for the forthcoming documentation of processes occurring in other parts of the repository system. A complete list of processes can be found in the Interim FEP report for the safety assessment SR-Can. All material presented in this document is preliminary in nature and will possibly be updated as the SR-Can project progresses

Interim process report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sellin, Patrick (ed.)

2004-08-01

This report is a documentation of buffer processes identified as relevant to the long-term safety of a KBS-3 repository. The report is part of the interim reporting of the safety assessment SR-Can, see further the Interim main report. The final SR-Can reporting will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of this report is to document the scientific knowledge of the processes to a level required for an adequate treatment in the safety assessment. The documentation is thus from a scientific point of not exhaustive since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. The purpose is further to determine the handling of each process in the safety assessment and to demonstrate how uncertainties are taken care of, given the suggested handling. The process documentation in the SR 97 version of the Process report is a starting point for this SR-Can interim version. As further described in the Interim main report, the list of relevant processes has been reviewed and slightly extended by comparison to other databases. Furthermore, the backfill has been included as a system part of its own, rather than being described together with the buffer as in SR 97. Apart from giving an interim account of the documentation and handling of buffer processes in SR-Can, this report is meant to serve as a template for the forthcoming documentation of processes occurring in other parts of the repository system. A complete list of processes can be found in the Interim FEP report for the safety assessment SR-Can. All material presented in this document is preliminary in nature and will possibly be updated as the SR-Can project progresses.

Probabilistic safety assessment goals in Canada

International Nuclear Information System (INIS)

Snell, V.G.

1986-01-01

CANDU safety philosphy, both in design and in licensing, has always had a strong bias towards quantitative probabilistically-based goals derived from comparative safety. Formal probabilistic safety assessment began in Canada as a design tool. The influence of this carried over later on into the definition of the deterministic safety guidelines used in CANDU licensing. Design goals were further developed which extended the consequence/frequency spectrum of 'acceptable' events, from the two points defined by the deterministic single/dual failure analysis, to a line passing through lower and higher frequencies. Since these were design tools, a complete risk summation was not necessary, allowing a cutoff at low event frequencies while preserving the identification of the most significant safety-related events. These goals gave a logical framework for making decisions on implementing design changes proposed as a result of the Probabilistic Safety Analysis. Performing this analysis became a regulatory requirement, and the design goals remained the framework under which this was submitted. Recently, there have been initiatives to incorporate more detailed probabilistic safety goals into the regulatory process in Canada. These range from far-reaching safety optimization across society, to initiatives aimed at the nuclear industry only. The effectiveness of the latter is minor at very low and very high event frequencies; at medium frequencies, a justification against expenditures per life saved in other industries should be part of the goal setting

Probabilistic Safety Assessment of Waste from PyroGreen Processes

International Nuclear Information System (INIS)

Ju, Hee Jae; Ham, In hye; Hwang, Il Soon

2016-01-01

The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

Probabilistic Safety Assessment of Waste from PyroGreen Processes

Energy Technology Data Exchange (ETDEWEB)

Ju, Hee Jae; Ham, In hye; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

2016-05-15

The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

Electronuclear's safety culture assessment and enhancement program

International Nuclear Information System (INIS)

Selvatici, E.; Diaz-Francisco, J.M.; Diniz de Souza, V.

2002-01-01

The present paper describes the Eletronuclear's safety culture assessment and enhancement program. The program was launched by the company's top management one year after the creation of Eletronuclear in 1997, from the merging of two companies with different organizational cultures, the design and engineering company Nuclen and the nuclear directorate of the Utility Furnas, Operator of the Angra1 NPP. The program consisted of an assessment performed internally in 1999 with the support and advice of the IAEA. This assessment, performed with the help of a survey, pooled about 80% of the company's employees. The overall result of the assessment was that a satisfactory level of safety culture existed; however, a number of points with a considerable margin for improvement were also identified. These points were mostly related with behavioural matters such as motivation, stress in the workplace, view of mistakes, handling of conflicts, and last but not least a view by a considerable number of employees that a conflict between safety and production might exist. An Action Plan was established by the company managers to tackle these weak points. This Plan was issued as company guideline by the company's Directorate. The subsequent step was to detail and implement the different actions of the Plan, which is the phase that we are at present. In the detailing of the Action Plan, special care was taken to sum up efforts, avoiding duplication of work or competition with already existing programs. In this process it was identified that the company had a considerable number of initiatives directly related to organizational and safety culture improvement, already operational. These initiatives have been integrated in the detailed Action Plan. A new assessment, for checking the effectiveness of the undertaken actions, is planned for 2003. (author)

Distracted walking: Examining the extent to pedestrian safety problems

Directory of Open Access Journals (Sweden)

Judith Mwakalonge

2015-10-01

Full Text Available Pedestrians, much like drivers, have always been engaged in multi-tasking like using hand-held devices, listening to music, snacking, or reading while walking. The effects are similar to those experienced by distracted drivers. However, distracted walking has not received similar policies and effective interventions as distracted driving to improve pedestrian safety. This study reviewed the state-of-practice on policies, campaigns, available data, identified research needs, and opportunities pertaining to distracted walking. A comprehensive review of literature revealed that some of the agencies/organizations disseminate useful information about certain distracting activities that pedestrians should avoid while walking to improve their safety. Various walking safety rules/tips have been given, such as not wearing headphones or talking on a cell phone while crossing a street, keeping the volume down, hanging up the phone while walking, being aware of traffic, and avoiding distractions like walking with texting. The majority of the past observational-based and experimental-based studies reviewed in this study on distracted walking is in agreement that there is a positive correlation between distraction and unsafe walking behavior. However, limitations of the existing crash data suggest that distracted walking may not be a severe threat to the public health. Current pedestrian crash data provide insufficient information for researchers to examine the extent to which distracted walking causes and/or contributes to actual pedestrian safety problems.

Development of a quality assurance safety assessment database for near surface radioactive waste disposal

International Nuclear Information System (INIS)

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

2003-01-01

A quality assurance safety assessment database, called QUARK (QUality Assurance program for Radioactive waste management in Korea), has been developed to manage both analysis information and parameter database for safety assessment of Low- and Intermediate-Level radioactive Waste (LILW) disposal facility in Korea. QUARK is such a tool that serves QA purposes for managing safety assessment information properly and securely. In QUARK, the information is organized and linked to maximize the integrity of information and traceability. QUARK provides guidance to conduct safety assessment analysis, from scenario generation to result analysis, and provides a window to inspect and trace previous safety assessment analysis and parameter values. QUARK also provides default database for safety assessment staff who construct input data files using SAGE(Safety Assessment Groundwater Evaluation), a safety assessment computer code

Safety assessments for deep geological disposal of radioactive wastes

International Nuclear Information System (INIS)

Lyon, R.B.

1984-01-01

The objective of safety assessment for deep geological disposal of radioactive wastes is to evaluate how well the engineered barriers and geological setting inhibit radionuclide migration and prevent radiation dose to man. Safety assessment is influenced through interaction with the regulatory agencies, research groups, the public and the various levels of government. Under the auspices of the IAEA, a generic disposal system description has been developed to facilitate international exchange and comparison of data and results, and to enable development and comparison of performance for all components of the disposal system. It is generally accepted that a systems modelling approach is required and that safety assessment can be considered on two levels. At the systems level, all components of the system are taken into account to evaluate the risk to man. At the systems level, critical review and quality assurance on software provide the major validation techniques. Risk is a combination of dose estimate and probability of that dose. For analysis of the total system to be practical, the components are usually represented by simplified models. Recently, assessments have been taking uncertainties in the input data into account. At the detailed level, large-scale, complex computer programs model components of the system in sufficient detail that validation by comparison with field and laboratory measurements is possible. For example, three-dimensional fluid-flow, heat-transport and solute-transport computer programs have been used. Approaches to safety assessment are described, with illustrations from safety assessments performed in a number of countries. (author)

New safety performance indicators for safety assessment of radioactive waste disposal facilities. Cuban experience

International Nuclear Information System (INIS)

Peralta Vital, J.L.; Castillo, R.G.; Olivera, J.

2002-01-01

The paper shows the Cuban experience on implementing geological disposal of radioactive waste and the necessity for identifying new safety performance indicators for the safety assessment (SA) of radioactive waste disposal facilities. The selected indicator was the concentration of natural radioactive elements (U, Ra, Th, K) in the Cuban geologic environment. We have carried out a group of investigations, which have allowed characterising the concentration for the whole Country, creating a wide database where this indicator is associated with the lithology. The main lithologies in Cuba are: the sedimentary rocks (70 percent of national occurrence), which are present in the three regions (limestone and lutite), and finally the igneous and metamorphic rocks. The results show the concentrations ranges of the natural radionuclides associated fundamentally to the variation in the lithology and geographical area of the Country. In Cuba, the higher concentration (ppm) of Uranium and Radium are referenced to the Central region associated to Skarn, while for Thorium (ppm) and Potassium (%), in the East region the concentration peaks in Tuffs have been found. The concentrations ranges obtained are preliminary, they characterise the behaviour of this parameter for the Cuban geology, but they do not represent limits for safety assessment purposes yet. Also other factors should be taken into account as the assessment context, time scales and others assumptions before establishing the final concentration limits for the natural radionuclides as a radiological and nuclear safety performance indicator complementary to dose and risk for safety assessment for radiological and nuclear facilities. (author)

Development of the JNC geological disposal technical information integration system subjected for repository design and safety assessment

International Nuclear Information System (INIS)

Ishihara, Yoshinao; Ito, Takashi; Kobayashi, Shigeki; Neyama, Atsushi

2004-02-01

On this work, system manufacture about disposal technology and safety assessment field was performed towards construction of the JNC Geological Disposal Technical Information Integration System which systematized three fields of technical information acquired in investigation (site characteristic investigation) of geology environmental conditions, disposal technology (design of deep repository), and performance/safety assessment. The technical information database managed focusing on the technical information concerning individual research of an examination, analysis, etc. and the parameter set database managed focusing on the set up data set used in case of comprehensive evaluation are examined. In order to support and promote share and use of the technical information registered and managed by the database, utility functions, such as a technical information registration function, technical information search/browse function, analysis support function, and visualization function, are considered, and the system realized in these functions is built. The built system is installed in the server of JNC, and the functional check examination is carried out. (author)

Examination of the bases for proposed innovations in reactor safety technology

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

This paper employs the criteria for evaluations from the Nuclear Power Option Viability Study to examine the bases for proposed innovations in light water reactor safety technology. These bases for innovation fall into four broad categories as follows: (1) virtually exclusive reliance on passive safety features to preclude core damage in all situations, (2) design simplification using some passive safety features to reduce the frequency of core damage to less than about 10 -6 per reactor-year, (3) passive containment to preclude releases from any accident, and (4) designing to limit licensing attention to one or at least a few systems. Of these, only the first two, and perhaps only the second, hold significant promise for providing for the viability of advanced light water reactors

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

OpenAIRE

Nganje, William E.

1999-01-01

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

A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-04-15

To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

International Nuclear Information System (INIS)

1990-04-01

To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

Value-impact assessment of safety-related modifications

International Nuclear Information System (INIS)

Knowles, W.M.C.; Dinnie, K.S.; Gordon, C.W.

1992-01-01

Like other nuclear utilities, Ontario Hydro, as part of its risk management activities, continually assesses the safety of its nuclear operations. In addition, new regulatory requirements are being applied to the older nuclear power plants. Both of these result in proposed plant modifications designed to reduce the risk to the public. However, modifications to an operating plant can have serious economic effects, and the resources, both financial and personnel, required for the implementation of these modifications are limited. Thus, all potential benefits and effects of a proposed modification must be thoroughly investigated to judge whether the modification is beneficial. Ontario Hydro has begun to use comprehensive value-impact assessments, utilizing plant-specific probabilistic risk assessments (PRAs), as tools to provide an informed basis for judgments on the benefit of safety-related modifications. The results from value-impact assessments can also be used to prioritize the implementation of these modifications

Principles governing heart failure therapy re-examined relative to standard evidence-based medicine-driven guidelines.

Science.gov (United States)

Tan, Lip-Bun; Chinnappa, Shanmugakumar; Tan, David K H; Hall, Alistair S

2011-09-01

Although all aspects of clinical work nowadays are modified by the pervading influence of evidence-based medicine (EBM) and multiplicative guidelines, not many clinicians realize that the underlying premise of EBM-driven guidelines is a particular strain of consequentialist ideology. Subservience to this ideology has transformed modern medical practice, but there is a real risk of distorting good medical practice, of belittling clinical judgement, of disempowering clinicians, and subjecting patients to skewed medical reality and treatment options. With so many heart failure (HF) guidelines issued by various august bodies, it is therefore timely to reappraise principles governing modern HF therapy with a fresh examination of the hierarchy of medical imperatives, the role of alternatives to consequentialism including deontological principles in HF therapy. In addition, other ideology worth re-examining, aside from EBM, are the principle of appropriate definition of HF underlying therapeutic goals and the principle of prioritizing objectives of HF therapy. Even within standard EBM, there are many questions to reconsider: about what types of evidence are admissible, different interpretations of available evidence, emphasizing patient-centered outcome measures instead of randomized controlled trials quantifiable therapeutic outcomes, how to prescribe drugs for prognostic versus symptomatic benefits, and how to deliver HF therapy based on pathophysiological features through mechanistic considerations and not just confined to randomized controlled trials or meta-analytical statistical imperatives. Through re-examination of these fundamental principles of HF therapy, it is hoped that clinicians will be empowered to manage HF patients more holistically and better deliver HF therapies in the best interest of each individual patient.

Formal safety assessment based on relative risks model in ship navigation

Energy Technology Data Exchange (ETDEWEB)

Hu Shenping [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: sphu@mmc.shmtu.edu.cn; Fang Quangen [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: qgfang@mmc.shmtu.edu.cn; Xia Haibo [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: hbxia@mmc.shmtu.edu.cn; Xi Yongtao [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: xiyt@mmc.shmtu.edu.cn

2007-03-15

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice.

Formal safety assessment based on relative risks model in ship navigation

International Nuclear Information System (INIS)

Hu Shenping; Fang Quangen; Xia Haibo; Xi Yongtao

2007-01-01

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice

Safety indicators: an efficient tool for a better safety

International Nuclear Information System (INIS)

Aufort, P.; Lars, R.

1993-01-01

Safety indicators based on the examination of the Operating Technical Specifications have been defined with the aim of following the in-operation safety level of French nuclear power plants. These safety indicators are operation feedback tools which permit the a posteriori justification and the adjustment of actual procedures. They would allow detection of an abnormal unavailability occurrence rate or a situation revealing a potential safety problem. So, data acquisition, processing, analysis and display software allowing trend analysis of these indicators has been developed so far as: a reflexion tool for the power plant operators about the safety instructions and the adjustment of preventive maintenance, and a help for decision making at a national level for the examination and the improvement of Operating Technical Specifications. This paper presents the objectives of these safety indicators, the processing tool associated, the preliminary results obtained and more elaborate processing of these indicators. These safety indicators may be very useful in framing probabilistic safety assessments. (author)

Integrated safety assessment report: Integrated Safety Assessment Program: Millstone Nuclear Power Station, Unit 1 (Docket No. 50-245): Draft report

International Nuclear Information System (INIS)

1987-04-01

The Integrated Safety Assessment Program (ISAP) was initiated in November 1984, by the US Nuclear Regulatory Commission to conduct integrated assessments for operating nuclear power reactors. The integrated assessment is conducted in a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. In addition, procedures will be established to allow for a periodic updating of the schedules to account for licensing issues that arise in the future. This report documents the review of Millstone Nuclear Power Station, Unit No. 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit No. 1, is one of two plants being reviewed under the pilot program for ISAP. This report indicates how 85 topics selected for review were addressed. This report presents the staff's recommendations regarding the corrective actions to resolve the 85 topics and other actions to enhance plant safety. The report is being issued in draft form to obtain comments from the licensee, nuclear safety experts, and the Advisory Committee for Reactor Safeguards (ACRS). Once those comments have been resolved, the staff will present its positions, along with a long-term implementation schedule from the licensee, in the final version of this report

Assessing progress in the development of safety culture

International Nuclear Information System (INIS)

Rotaru, Ioan; Ghita, Sorin

1999-01-01

The concept of safety culture was introduced by the International Nuclear Safety Advisory Group (INSAG) in the Summary Report on the Post-Accident Meeting on the Chernobyl Accident in 1986. The concept was further expanded in the 1988 INSAG-3 report, Basic Safety Principles for Nuclear Power Plants, and again in 1991 in the INSAG-4 report. Recognizing the increasing role that safety culture is expected to play in nuclear installations worldwide, the Convention on Nuclear Safety states the Contracting Parties' desire 'to promote an effective nuclear safety culture'. The concept of safety culture is defined in INSAG-4 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. Safety culture is also an amalgamation of values, standards, morals and norms of acceptable behaviour. These are aimed at maintaining a self disciplined approach to the enhancement of safety beyond legislative and regulatory requirements. Therefore, the safety culture has to be inherent in the thoughts and actions of all the individuals at every level in an organization. The leadership provided by top management is crucial. Safety culture applies to conventional and personal safety as well as nuclear safety. All safety consideration are affected by common points of beliefs, attitudes, behaviour, and cultural differences, closely linked to a shared system of values and standards. The paper poses questions and tries to find answers relative to issues like: - how to assess progress; - specific organizational indicators of a progressive safety culture; - detection of incipient weaknesses in safety culture (organizational issues, employee issues, technology issues); - revitalizing a weakened safety culture; - overall assesment of safety culture; - general evaluation model. In conclusion, there is no consistent and

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

Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities

International Nuclear Information System (INIS)

Batandjieva, B.; Torres-Vidal, C.

2002-01-01

The International Atomic Energy Agency (IAEA) Coordinated research program ''Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities'' (ISAM) has developed improved safety assessment methodology for near surface disposal facilities. The program has been underway for three years and has included around 75 active participants from 40 countries. It has also provided examples for application to three safety cases--vault, Radon type and borehole radioactive waste disposal facilities. The program has served as an excellent forum for exchange of information and good practices on safety assessment approaches and methodologies used worldwide. It also provided an opportunity for reaching broad consensus on the safety assessment methodologies to be applied to near surface low and intermediate level waste repositories. The methodology has found widespread acceptance and the need for its application on real waste disposal facilities has been clearly identified. The ISAM was finalized by the end of 2000, working material documents are available and an IAEA report will be published in 2002 summarizing the work performed during the three years of the program. The outcome of the ISAM program provides a sound basis for moving forward to a new IAEA program, which will focus on practical application of the safety assessment methodologies to different purposes, such as licensing radioactive waste repositories, development of design concepts, upgrading existing facilities, reassessment of operating repositories, etc. The new program will also provide an opportunity for development of guidance on application of the methodology that will be of assistance to both safety assessors and regulators

[Agricultural biotechnology safety assessment].

Science.gov (United States)

McClain, Scott; Jones, Wendelyn; He, Xiaoyun; Ladics, Gregory; Bartholomaeus, Andrew; Raybould, Alan; Lutter, Petra; Xu, Haibin; Wang, Xue

2015-01-01

Genetically modified (GM) crops were first introduced to farmers in 1995 with the intent to provide better crop yield and meet the increasing demand for food and feed. GM crops have evolved to include a thorough safety evaluation for their use in human food and animal feed. Safety considerations begin at the level of DNA whereby the inserted GM DNA is evaluated for its content, position and stability once placed into the crop genome. The safety of the proteins coded by the inserted DNA and potential effects on the crop are considered, and the purpose is to ensure that the transgenic novel proteins are safe from a toxicity, allergy, and environmental perspective. In addition, the grain that provides the processed food or animal feed is also tested to evaluate its nutritional content and identify unintended effects to the plant composition when warranted. To provide a platform for the safety assessment, the GM crop is compared to non-GM comparators in what is typically referred to as composition equivalence testing. New technologies, such as mass spectrometry and well-designed antibody-based methods, allow better analytical measurements of crop composition, including endogenous allergens. Many of the analytical methods and their intended uses are based on regulatory guidance documents, some of which are outlined in globally recognized documents such as Codex Alimentarius. In certain cases, animal models are recommended by some regulatory agencies in specific countries, but there is typically no hypothesis or justification of their use in testing the safety of GM crops. The quality and standardization of testing methods can be supported, in some cases, by employing good laboratory practices (GLP) and is recognized in China as important to ensure quality data. Although the number of recommended, in some cases, required methods for safety testing are increasing in some regulatory agencies, it should be noted that GM crops registered to date have been shown to be

Assessing propensity to learn from safety-related events

NARCIS (Netherlands)

Drupsteen, L.; Wybo, J.L.

2015-01-01

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

Healthcare professionals’ views of feedback on patient safety culture assessment.

NARCIS (Netherlands)

Zwijnenberg, N.C.; Hendriks, M.; Hoogervorst-Schilp, J.; Wagner, C.

2016-01-01

Background: By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals’ views on the

The Victim-Offender Overlap and Fear of In-School Victimization: A Longitudinal Examination of Risk Assessment Models

Science.gov (United States)

Melde, Chris; Esbensen, Finn-Aage

2009-01-01

Reports of serious violence in schools have raised general awareness and concern about safety in America's schools. In this article, the authors examine the extent to which in-school victimization is associated with students' perceived risk and fear of victimization. By expanding on Ferraro's risk assessment framework, the current study explores…

Application of the Integrated Safety Assessment methodology to safety margins. Dynamic Event Trees, Damage Domains and Risk Assessment

International Nuclear Information System (INIS)

Ibánez, L.; Hortal, J.; Queral, C.; Gómez-Magán, J.; Sánchez-Perea, M.; Fernández, I.; Meléndez, E.; Expósito, A.; Izquierdo, J.M.; Gil, J.; Marrao, H.; Villalba-Jabonero, E.

2016-01-01

The Integrated Safety Assessment (ISA) methodology, developed by the Consejo de Seguridad Nuclear, has been applied to an analysis of Zion NPP for sequences with Loss of the Component Cooling Water System (CCWS). The ISA methodology proposal starts from the unfolding of the Dynamic Event Tree (DET). Results from this first step allow assessing the sequence delineation of standard Probabilistic Safety Analysis results. For some sequences of interest of the outlined DET, ISA then identifies the Damage Domain (DD). This is the region of uncertain times and/or parameters where a safety limit is exceeded, which indicates the occurrence of certain damage situation. This paper illustrates application of this concept obtained simulating sequences with MAAP and with TRACE. From information of simulation results of sequence transients belonging to the DD and the time-density probability distributions of the manual actions and of occurrence of stochastic phenomena, ISA integrates the dynamic reliability equations proposed to obtain the sequence contribution to the global Damage Exceedance Frequency (DEF). Reported results show a slight increase in the DEF for sequences investigated following a power uprate from 100% to 110%. This demonstrates the potential use of the method to help in the assessment of design modifications. - Highlights: • This paper illustrates an application of the ISA methodology to safety margins. • Dynamic Event Trees are useful tool for verifying the standard PSA Event Trees. • The ISA methodology takes into account the uncertainties in human action times. • The ISA methodology shows the Damage Exceedance Frequency increase in power uprates.

Digital Intraoral Imaging Re-Exposure Rates of Dental Students.

Science.gov (United States)

Senior, Anthea; Winand, Curtis; Ganatra, Seema; Lai, Hollis; Alsulfyani, Noura; Pachêco-Pereira, Camila

2018-01-01

A guiding principle of radiation safety is ensuring that radiation dosage is as low as possible while yielding the necessary diagnostic information. Intraoral images taken with conventional dental film have a higher re-exposure rate when taken by dental students compared to experienced staff. The aim of this study was to examine the prevalence of and reasons for re-exposure of digital intraoral images taken by third- and fourth-year dental students in a dental school clinic. At one dental school in Canada, the total number of intraoral images taken by third- and fourth-year dental students, re-exposures, and error descriptions were extracted from patient clinical records for an eight-month period (September 2015 to April 2016). The data were categorized to distinguish between digital images taken with solid-state sensors or photostimulable phosphor plates (PSP). The results showed that 9,397 intraoral images were made, and 1,064 required re-exposure. The most common error requiring re-exposure for bitewing images was an error in placement of the receptor too far mesially or distally (29% for sensors and 18% for PSP). The most common error requiring re-exposure for periapical images was inadequate capture of the periapical area (37% for sensors and 6% for PSP). A retake rate of 11% was calculated, and the common technique errors causing image deficiencies were identified. Educational intervention can now be specifically designed to reduce the retake rate and radiation dose for future patients.

Guidance on the safety assessment methodology for storage of radioactive waste

International Nuclear Information System (INIS)

Kinyanjui, M.N.

2014-04-01

This project on safety assessment on storage was carried out with the main objective of ensuring safety of human life and our environment. This is the fundamental principle of radiation protection. Safety assessment has been evaluated as a tool in the safety case in the pre-construction, operational and the post closure phase of storage. In particular the iterative process of evaluating and predicting safety scenarios at each stage of the process has proved to be prudent. It is important that this concept be adopted for this type of facility to ensure safety of mankind and the environment now and in the future.

Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 1 - guideword applicability and method description.

Science.gov (United States)

Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

2008-12-15

Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design.

Guideline for examination concerning the evaluation of safety in light water power reactor installations

International Nuclear Information System (INIS)

1978-01-01

This guideline was drawn up as the guide for examination when the safety evaluation of nuclear reactors is carried out at the time of approving the installation of light water power reactors. Accordingly in case of the examination of safety, it must be confirmed that the contents of application are in conformity with this guideline. If they are in conformity, it is judged that the safety evaluation of the policy in the basic design of a reactor facility is adequate, and also that the evaluation concerning the separation from the public in surroundings is adequate as the condition of location of the reactor facility. This guideline is concerned with light water power reactors now in use, but the basic concept may be the reference for the examination of the other types of reactors. If such a case occurs that the safety evaluation does not conform to this guideline, it is not excluded when the appropriate reason is clarified. The purpose of safety evaluation, the scope to be evaluated, the selection of the events to be evaluated, the criteria for judgement, the matters taken into consideration at the time of analysis, the concrete events of abnormal transient change and accident in operation, and the concrete events of serious accident and hypothetic accident are stipulated. The explanation and two appendices are attached. (Kako, I.)

Intrusion resistant underground structure (IRUS) - safety assessment and licensing

International Nuclear Information System (INIS)

Lange, B. A.

1997-01-01

This paper describes the safety goals, human exposure scenarios and critical groups, the syvac-nsure performance assessment code, groundwater pathway safety results, and inadvertent human intrusion of the IRUS. 2 tabs

Safety assessment of complex engineered and natural systems: radioactive waste disposal

International Nuclear Information System (INIS)

McNeish, J.A.; Vallikat, V.; Atkins, J.; Balady, M.A.

1997-01-01

Evaluation of deep, geologic disposal of nuclear waste requires the probabilistic safety assessment of a complex system from the coupling of various processes and sub-systems, parameter and model uncertainties, spatial and temporal variabilities, and the multiplicity of designs and scenarios. Both the engineered and natural system are included in the evaluation. Each system has aspects with considerable uncertainty both in important parameters and in overall conceptual models. The study represented herein provides a probabilistic safety assessment of a potential respository system for multiple engineered barrier system (EBS) design and conceptual model configurations (CRWMS M and O, 1996a) and considers the effects of uncertainty on the overall results. The assessment is based on data and process models available at the time of the study and doesnt necessarily represent the current safety evaluation. In fact, the percolation flux through the repository system is now expected to be higher than the estimate used for this study. The potential effects of higher percolation fluxes are currently under study. The safety of the system was assessed for both 10,000 and 1,000,000 years. Use of alternative conceptual models also produced major improvement in safety. For example, use of a more realistic engineered system release model produced improvement of over an order of magnitude in safety. Alternative measurement locations for the safety assessment produced substantial increases in safety, through the results are based on uncertain dilution factors in the transporting groundwater. (Author)

Safety Culture Perceptions in a Collegiate Aviation Program: A Systematic Assessment