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Sample records for safety assessment diagram

  1. Safety- barrier diagrams

    DEFF Research Database (Denmark)

    Duijm, Nijs Jan

    2008-01-01

    Safety-barrier diagrams and the related so-called 'bow-tie' diagrams have become popular methods in risk analysis. This paper describes the syntax and principles for constructing consistent and valid safety-barrier diagrams. The relation of safety-barrier diagrams to other methods such as fault...... trees and Bayesian networks is discussed. A simple method for quantification of safety-barrier diagrams is proposed. It is concluded that safety-barrier diagrams provide a useful framework for an electronic data structure that integrates information from risk analysis with operational safety management....

  2. Safety-barrier diagrams

    DEFF Research Database (Denmark)

    Duijm, Nijs Jan

    2007-01-01

    Safety-barrier diagrams and the related so-called "bow-tie" diagrams have become popular methods in risk analysis. This paper describes the syntax and principles for constructing consistent and valid safety-barrier diagrams. The relation with other methods such as fault trees and Bayesian networks...... are discussed. A simple method for quantification of safety-barrier diagrams is proposed, including situations where safety barriers depend on shared common elements. It is concluded that safety-barrier diagrams provide a useful framework for an electronic data structure that integrates information from risk...... analysis with operational safety management....

  3. Safety-barrier diagrams as a safety management tool

    DEFF Research Database (Denmark)

    Duijm, Nijs Jan

    2009-01-01

    Safety-barrier diagrams and “bow-tie” diagrams have become popular methods in risk analysis and safety management. This paper describes the syntax and principles for constructing consistent and valid safety-barrier diagrams. The latter's relation to other methods such as fault trees and Bayesian...

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

    DEFF Research Database (Denmark)

    Duijm, Nijs Jan; Markert, Frank

    2009-01-01

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

  5. 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

  6. Probabilistic safety assessment model in consideration of human factors based on object-oriented bayesian networks

    International Nuclear Information System (INIS)

    Zhou Zhongbao; Zhou Jinglun; Sun Quan

    2007-01-01

    Effect of Human factors on system safety is increasingly serious, which is often ignored in traditional probabilistic safety assessment methods however. A new probabilistic safety assessment model based on object-oriented Bayesian networks is proposed in this paper. Human factors are integrated into the existed event sequence diagrams. Then the classes of the object-oriented Bayesian networks are constructed which are converted to latent Bayesian networks for inference. Finally, the inference results are integrated into event sequence diagrams for probabilistic safety assessment. The new method is applied to the accident of loss of coolant in a nuclear power plant. the results show that the model is not only applicable to real-time situation assessment, but also applicable to situation assessment based certain amount of information. The modeling complexity is kept down and the new method is appropriate to large complex systems due to the thoughts of object-oriented. (authors)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  8. Use of Influence Diagrams and Fuzzy Theory to Develop Assessment Method of Organizational Influences on Component Maintenance

    International Nuclear Information System (INIS)

    Yoonik Kim; Kwang-Won Ahn; Chang-Hyun Chung; Kil Yoo Kim; Joon-Eon Yang

    2002-01-01

    Organization can make influences on all the systems. Especially in case of nuclear power plants in which safety is established to be one of the most important operating goals, there have been a lot of research efforts for the hardware advancement. However in recent years, it has been widely recognized that organizational factors in nuclear power plants have an important influence on the safety attitudes and the safe behavior of individuals. Until now, any means to include assessments of organizational structure in probabilistic risk assessments have not been universally accepted. The objective of this work is to develop a method to assess organizational influences on component maintenance. Influence diagrams are introduced in this method as a decision making tool and fuzzy theory is used to reflect the vagueness in considering relevance of human activities in maintenance tasks. Introducing fuzzy theory to assess the organizational factors is deemed to a somewhat new trial, which makes it possible to convert linguistic vague descriptions into mathematical ones. Fuzzy linguistic descriptions offer an alternative and often complementary language to conventional, i.e., analytic approaches to modeling systems. Among the existing methodologies to assess organizational factors, the concept of the ω-factor model is utilized and the mechanism that organizational factors have influences on component maintenance is evaluated through composing influence diagrams. These influences go to failure rates and eventually affect component unavailability. Further study will make it possible that the influences of organizational factors on human error probabilities are incorporated into human reliability analysis and furthermore probabilistic safety assessment. (authors)

  9. A simple reliability block diagram method for safety integrity verification

    International Nuclear Information System (INIS)

    Guo Haitao; Yang Xianhui

    2007-01-01

    IEC 61508 requires safety integrity verification for safety related systems to be a necessary procedure in safety life cycle. PFD avg must be calculated to verify the safety integrity level (SIL). Since IEC 61508-6 does not give detailed explanations of the definitions and PFD avg calculations for its examples, it is difficult for common reliability or safety engineers to understand when they use the standard as guidance in practice. A method using reliability block diagram is investigated in this study in order to provide a clear and feasible way of PFD avg calculation and help those who take IEC 61508-6 as their guidance. The method finds mean down times (MDTs) of both channel and voted group first and then PFD avg . The calculated results of various voted groups are compared with those in IEC61508 part 6 and Ref. [Zhang T, Long W, Sato Y. Availability of systems with self-diagnostic components-applying Markov model to IEC 61508-6. Reliab Eng System Saf 2003;80(2):133-41]. An interesting outcome can be realized from the comparison. Furthermore, although differences in MDT of voted groups exist between IEC 61508-6 and this paper, PFD avg of voted groups are comparatively close. With detailed description, the method of RBD presented can be applied to the quantitative SIL verification, showing a similarity of the method in IEC 61508-6

  10. The shape of a strain-based failure assessment diagram

    International Nuclear Information System (INIS)

    Budden, P.J.; Ainsworth, R.A.

    2012-01-01

    There have been a number of recent developments of strain-based fracture assessment approaches, including proposals by Budden [Engng Frac Mech 2006;73:537–52] for a strain-based failure assessment diagram (FAD) related to the conventional stress-based FAD. However, recent comparisons with finite element (FE) data have shown that this proposed strain-based FAD can be non-conservative in some cases, particularly for deeper cracks and materials with little strain-hardening capacity. Therefore, this paper re-examines the shape of the strain-based FAD, guided by these FE analyses and some theoretical analysis. On this basis, modified proposals for the shape of the strain-based FAD are given, including simplified and more detailed options in line with the options available for stress-based FADs in existing fitness-for-service procedures. The proposals are then illustrated by a worked example and by comparison with FE data, which demonstrate that the new proposals are generally conservative. - Highlights: ► The strain-based failure assessment diagram approach to fracture is developed. ► The new approach modifies earlier proposals by Budden. ► A new generic Option 1 strain-based failure assessment diagram is proposed. ► Validation based on finite element J data for plates and cylinders is presented. ► The new approach is generally conservative compared with the finite element data.

  11. Human factors in safety assessment. Safety culture assessment

    International Nuclear Information System (INIS)

    Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

    1996-01-01

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

  12. Safe disposal of radioactive waste. Post-closure safety assessment of permanent repository in Novi han

    International Nuclear Information System (INIS)

    Mateeva, M.

    2007-01-01

    A presented material is the third part of the monograph with title 'Safe disposal of radioactive waste. Post-closure safety assessment of the permanent repository in Novi Han'. This part deals with review of the scenario selection procedure. The process system of permanent repository for radioactive waste is describing in details for different levels. Preliminary screening process of features, events and processes is presented here. Interaction matrixes for basic disposal system components are constructed. Final selection and grouping between the included features, events and processes is done. Selected and defined scenarios for post-closure safety assessment are presented too. Key words: post-closure safety assessment, scenario generation procedure, process system, process influence diagram, and interaction matrix

  13. Risk assessment of computer-controlled safety systems for fusion reactors

    International Nuclear Information System (INIS)

    Fryer, M.O.; Bruske, S.Z.

    1983-01-01

    The complexity of fusion reactor systems and the need to display, analyze, and react promptly to large amounts of information during reactor operation will require a number of safety systems in the fusion facilities to be computer controlled. Computer software, therefore, must be included in the reactor safety analyses. Unfortunately, the science of integrating computer software into safety analyses is in its infancy. Combined plant hardware and computer software systems are often treated by making simple assumptions about software performance. This method is not acceptable for assessing risks in the complex fusion systems, and a new technique for risk assessment of combined plant hardware and computer software systems has been developed. This technique is an extension of the traditional fault tree analysis and uses structured flow charts of the software in a manner analogous to wiring or piping diagrams of hardware. The software logic determines the form of much of the fault trees

  14. ARAMIS project: a more explicit demonstration of risk control through the use of bow-tie diagrams and the evaluation of safety barrier performance.

    Science.gov (United States)

    de Dianous, Valérie; Fiévez, Cécile

    2006-03-31

    Over the last two decades a growing interest for risk analysis has been noted in the industries. The ARAMIS project has defined a methodology for risk assessment. This methodology has been built to help the industrialist to demonstrate that they have a sufficient risk control on their site. Risk analysis consists first in the identification of all the major accidents, assuming that safety functions in place are inefficient. This step of identification of the major accidents uses bow-tie diagrams. Secondly, the safety barriers really implemented on the site are taken into account. The barriers are identified on the bow-ties. An evaluation of their performance (response time, efficiency, and level of confidence) is performed to validate that they are relevant for the expected safety function. At last, the evaluation of their probability of failure enables to assess the frequency of occurrence of the accident. The demonstration of the risk control based on a couple gravity/frequency of occurrence is also possible for all the accident scenarios. During the risk analysis, a practical tool called risk graph is used to assess if the number and the reliability of the safety functions for a given cause are sufficient to reach a good risk control.

  15. AADL Fault Modeling and Analysis Within an ARP4761 Safety Assessment

    Science.gov (United States)

    2014-10-01

    Analysis Generator 27 3.2.3 Mapping to OpenFTA Format File 27 3.2.4 Mapping to Generic XML Format 28 3.2.5 AADL and FTA Mapping Rules 28 3.2.6 Issues...PSSA), System Safety Assessment (SSA), Common Cause Analysis (CCA), Fault Tree Analysis ( FTA ), Failure Modes and Effects Analysis (FMEA), Failure...Modes and Effects Summary, Mar - kov Analysis (MA), and Dependence Diagrams (DDs), also referred to as Reliability Block Dia- grams (RBDs). The

  16. Analysis of leak and break behavior in a failure assessment diagram for carbon steel pipes

    International Nuclear Information System (INIS)

    Kanno, Satoshi; Hasegawa, Kunio; Shimizu, Tasuku; Saitoh, Takashi; Gotoh, Nobuho

    1992-01-01

    The leak and break behavior of a cracked coolant pipe subjected to an internal pressure and a bending moment was analyzed with a failure assessment diagram using the R6 approach. This paper examines the conditions of the detectable coolant leakage without breakage. A leakage assessment curve, a locus of assessment point for detectable coolant leakage, was defined in the failure assessment diagram. The region between the leak assessment and failure assessment curves satisfies the condition of detectable leakage without breakage. In this region, a crack can be safely inspected by a coolant leak detector. (orig.)

  17. Top-down versus bottom-up processing of influence diagrams in probabilistic analysis

    International Nuclear Information System (INIS)

    Timmerman, R.D.; Burns, T.J.; Dodds, H.L. Jr.

    1986-01-01

    Recent work by Phillips et al and Selby et al has shown that influence diagram methodology can be a useful analytical tool in reactor safety studies. In some instances, an influence diagram can be used as a graphical representation of probabilistic dependence within a system or event sequence. Under these circumstances, Bayesian statistics is employed to transform the relationships depicted in the influence diagram into the correct expression for a desired marginal probability (e.g., the top node). In the references cited above, the authors demonstrated the usefulness of influence diagrams for assessing the reliability of operator performance during pressurized thermal shock transients. In addition, the use of influence diagrams identified the critical variables that had the greatest impact on operator reliability for a particular scenario (e.g., control room design, procedures, etc.). Top-down and bottom-up algorithms have emerged as the dominant methods for quantifying influence diagrams. The purpose of this paper is to demonstrate a potential error in employing the bottom-up algorithm when dealing with interdependencies

  18. A new model of Ishikawa diagram for quality assessment

    Science.gov (United States)

    Liliana, Luca

    2016-11-01

    The paper presents the results of a study concerning the use of the Ishikawa diagram in analyzing the causes that determine errors in the evaluation of theparts precision in the machine construction field. The studied problem was"errors in the evaluation of partsprecision” and this constitutes the head of the Ishikawa diagram skeleton.All the possible, main and secondary causes that could generate the studied problem were identified. The most known Ishikawa models are 4M, 5M, 6M, the initials being in order: materials, methods, man, machines, mother nature, measurement. The paper shows the potential causes of the studied problem, which were firstly grouped in three categories, as follows: causes that lead to errors in assessing the dimensional accuracy, causes that determine errors in the evaluation of shape and position abnormalities and causes for errors in roughness evaluation. We took into account the main components of parts precision in the machine construction field. For each of the three categories of causes there were distributed potential secondary causes on groups of M (man, methods, machines, materials, environment/ medio ambiente-sp.). We opted for a new model of Ishikawa diagram, resulting from the composition of three fish skeletons corresponding to the main categories of parts accuracy.

  19. Application of failure assessment diagram methods to cracked straight pipes and elbows

    International Nuclear Information System (INIS)

    Ainsworth, R.A.; Gintalas, M.; Sahu, M.K.; Chattopadhyay, J.; Dutta, B.K.

    2016-01-01

    This paper reports fracture assessments of large-scale straight pipes and elbows of various pipe diameters and crack sizes. The assessments estimate the load for ductile fracture initiation using the failure assessment diagram method. Recent solutions in the literature for stress intensity factor and limit load provide the analysis inputs. An assessment of constraint effects is also performed using recent solutions for elastic T-stress. It is found that predictions of initiation load are close to the experimental values for straight pipes under pure bending. For elbows, there is generally increased conservatism in the sense that the experimental loads are greater than those predicted. The effects of constraint are found not to be a major contributor to the initiation fracture assessments but may have some influence on the ductile crack extension. - Highlights: • This paper presents assessments of the loads for ductile fracture initiation in 21 large-scale piping tests. • Modern stress intensity factor and limit load solutions were used for standard failure assessment diagram methods. • This leads to generally accurate assessments of the loads for ductile crack initiation. • The effects of constraint are found not to be a major contributor to the initiation fracture assessments.

  20. Dynamic safety assessment of natural gas stations using Bayesian network

    Energy Technology Data Exchange (ETDEWEB)

    Zarei, Esmaeil, E-mail: smlzarei65@gmail.com [Center of Excellence for Occupational Health Engineering, Research Center for Health Sciences, Faculty of Health, Hamadan University of Medical Sciences, Hamadan (Iran, Islamic Republic of); Azadeh, Ali [School of Industrial and Systems Engineering, Center of Excellence for Intelligent-Based Experimental Mechanic, College of Engineering, University of Tehran (Iran, Islamic Republic of); Khakzad, Nima [Safety and Security Science Section, Delft University of Technology, Delft (Netherlands); Aliabadi, Mostafa Mirzaei [Center of Excellence for Occupational Health Engineering, Research Center for Health Sciences, Faculty of Health, Hamadan University of Medical Sciences, Hamadan (Iran, Islamic Republic of); Mohammadfam, Iraj, E-mail: mohammadfam@umsha.ac.ir [Center of Excellence for Occupational Health Engineering, Research Center for Health Sciences, Faculty of Health, Hamadan University of Medical Sciences, Hamadan (Iran, Islamic Republic of)

    2017-01-05

    Graphical abstract: Dynamic cause-consequence analysis of the regulator system failure using BN. - Highlights: • A dynamic and comprehensive QRA (DCQRA) framework is proposed for safety assessment of CGSs. • Bow-tie diagram and Bayesian network are employed for accident scenario modeling. • Critical basic events and minimal cut sets are identified using probability updating. - Abstract: Pipelines are one of the most popular and effective ways of transporting hazardous materials, especially natural gas. However, the rapid development of gas pipelines and stations in urban areas has introduced a serious threat to public safety and assets. Although different methods have been developed for risk analysis of gas transportation systems, a comprehensive methodology for risk analysis is still lacking, especially in natural gas stations. The present work is aimed at developing a dynamic and comprehensive quantitative risk analysis (DCQRA) approach for accident scenario and risk modeling of natural gas stations. In this approach, a FMEA is used for hazard analysis while a Bow-tie diagram and Bayesian network are employed to model the worst-case accident scenario and to assess the risks. The results have indicated that the failure of the regulator system was the worst-case accident scenario with the human error as the most contributing factor. Thus, in risk management plan of natural gas stations, priority should be given to the most probable root events and main contribution factors, which have identified in the present study, in order to reduce the occurrence probability of the accident scenarios and thus alleviate the risks.

  1. Dynamic safety assessment of natural gas stations using Bayesian network

    International Nuclear Information System (INIS)

    Zarei, Esmaeil; Azadeh, Ali; Khakzad, Nima; Aliabadi, Mostafa Mirzaei; Mohammadfam, Iraj

    2017-01-01

    Graphical abstract: Dynamic cause-consequence analysis of the regulator system failure using BN. - Highlights: • A dynamic and comprehensive QRA (DCQRA) framework is proposed for safety assessment of CGSs. • Bow-tie diagram and Bayesian network are employed for accident scenario modeling. • Critical basic events and minimal cut sets are identified using probability updating. - Abstract: Pipelines are one of the most popular and effective ways of transporting hazardous materials, especially natural gas. However, the rapid development of gas pipelines and stations in urban areas has introduced a serious threat to public safety and assets. Although different methods have been developed for risk analysis of gas transportation systems, a comprehensive methodology for risk analysis is still lacking, especially in natural gas stations. The present work is aimed at developing a dynamic and comprehensive quantitative risk analysis (DCQRA) approach for accident scenario and risk modeling of natural gas stations. In this approach, a FMEA is used for hazard analysis while a Bow-tie diagram and Bayesian network are employed to model the worst-case accident scenario and to assess the risks. The results have indicated that the failure of the regulator system was the worst-case accident scenario with the human error as the most contributing factor. Thus, in risk management plan of natural gas stations, priority should be given to the most probable root events and main contribution factors, which have identified in the present study, in order to reduce the occurrence probability of the accident scenarios and thus alleviate the risks.

  2. ASSESSING PHYSICAL DEVELOPMENT OF CHILDREN WITH PERCENTILE DIAGRAMS

    Directory of Open Access Journals (Sweden)

    Rita R. Kildiyarova

    2017-01-01

    Full Text Available The results of the analysis of methods assessing anthropometric measures in children are presented. A method for visual examination of physical development using author's percentile diagrams for height, body weight, and the harmony of development of children of different age groups is offered. The method can be quickly performed, it is recommended for mass screening examination of children under outpatient treatment. To monitor the health of a specific child, a monitoring assessment of physical development is possible. The analysis of Z-score is of great clinical importance when determining anthropometric measures below the 3rd percentile, for the assessment of premature infants with congenital malformations and other diseases, in the presence of obesity. Graphical curves of body weight, height to age, body weight according to the height of boys and girls can be used by pediatricians. 

  3. Safety assurance logic techniques for evaluation of accident prevention and mitigation

    International Nuclear Information System (INIS)

    McWethy, L.M.; Hagan, J.W.

    1976-01-01

    Safety assurance methods have been developed and applied in reactor safety assessments of FFTF. These methods promote visibility of the total safety provided by the plant, both in prevention of off-normal or accident conditions as well as provision of various features which terminate conditions within acceptable bounds if such conditions should occur. One of the primary techniques applied in safety assurance is the development of safety assurance diagrams. These diagrams explicitly identify the multiple lines of defense which prevent accident progression. The diagrams graphically demonstrate the defense-in-depth provided by the plant for each postulated occurrence. Lines of defense are shown against ever having an occurrence in the first place; thus giving appropriate emphasis on accident prevention, and visibility to the designer's role in promoting this level of safety. These diagrams, or accident process trees, also show graphically the various paths of postulated accident progression to their logical termination. Evaluation of the importance and strength of each line-of-defense assures fulfillment of the safety objectives of the overall plant system

  4. Extended block diagram method for a multi-state system reliability assessment

    International Nuclear Information System (INIS)

    Lisnianski, Anatoly

    2007-01-01

    The presented method extends the classical reliability block diagram method to a repairable multi-state system. It is very suitable for engineering applications since the procedure is well formalized and based on the natural decomposition of the entire multi-state system (the system is represented as a collection of its elements). Until now, the classical block diagram method did not provide the reliability assessment for the repairable multi-state system. The straightforward stochastic process methods are very difficult for engineering application in such cases due to the 'dimension damnation'-huge number of system states. The suggested method is based on the combined random processes and the universal generating function technique and drastically reduces the number of states in the multi-state model

  5. Evaluation of flaws in ferritic piping: ASME Code Appendix J, Deformation Plasticity Failure Assessment Diagram (DPFAD)

    International Nuclear Information System (INIS)

    Bloom, J.M.

    1991-08-01

    This report summarizes the methods and bases used by an ASME Code procedure for the evaluation of flaws in ferritic piping. The procedure is currently under consideration by the ASME Boiler and Pressure Vessel Code Committee of Section 11. The procedure was initially proposed in 1985 for the evaluation of the acceptability of flaws detected in piping during in-service inspection for certain materials, identified in Article IWB-3640 of the ASME Boiler and Pressure Vessel Code Section 11 ''Rules for In-service Inspection of Nuclear Power Plant Components.'' for which the fracture toughness is not sufficiently high to justify acceptance based solely on the plastic limit load evaluation methodology of Appendix C and IWB-3641. The procedure, referred to as Appendix J, originally included two approaches: a J-integral based tearing instability (J-T) analysis and the deformation plasticity failure assessment diagram (DPFAD) methodology. In Appendix J, a general DPFAD approach was simplified for application to part-through wall flows in ferritic piping through the use of a single DPFAD curve for circumferential flaws. Axial flaws are handled using two DPFAD curves where the ratio of flaw depth to wall thickness is used to determine the appropriate DPFAD curve. Flaws are evaluated in Appendix J by comparing the actual pipe applied stress with the allowable stress with the appropriate safety factors for the flaw size at the end of the evaluation period. Assessment points for circumferential and axial flaws are plotted on the appropriate failure assessment diagram. In addition, this report summarizes the experimental test predictions of the results of the Battelle Columbus Laboratory experiments, the Eiber experiments, and the JAERI tests using the Appendix J DPFAD methodology. Lastly, this report also provides guidelines for handling residual stresses in the evaluation procedure. 22 refs., 13 figs., 5 tabs

  6. Use of reliability engineering tools in safety and risk assessment of nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias, E-mail: amandaraso@hotmail.com, E-mail: vasconv@cdtn.br, E-mail: raissaomarques@gmail.com, E-mail: soaresw@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Serviço de Tecnologia de Reatores

    2017-07-01

    Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

  7. Use of reliability engineering tools in safety and risk assessment of nuclear facilities

    International Nuclear Information System (INIS)

    Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias

    2017-01-01

    Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

  8. Pembuatan Kakas Bantu untuk Mendeteksi Ketidaksesuaian Diagram Urutan (Sequence Diagram dengan Diagram Kasus Penggunaan (Use Case Diagram

    Directory of Open Access Journals (Sweden)

    Andrias Meisyal Yuwantoko

    2017-03-01

    Full Text Available Sebuah diagram urutan dibuat  berdasarkan alur yang ada pada deskripsi kasus penggunaan. Alur tersebut dire- presentasikan dalam  bentuk  interaksi antara aktor  dan  sistem. Pemeriksaan rancangan diagram urutan perlu dilakukan untuk mengetahui ketidaksesuaian urutan alur  kasus penggunaan dengan urutan pesan yang dikirimkan oleh objek-objek pada diagram urutan. Rancangan diagram yang sesuai merupakan kunci ketepatan (correctness implementasi  perangkat lunak. Namun, pemeriksaan ketidaksesuaian masih dilakukan secara manual. Hal ini menjadi masalah apabila sebuah proyek perangkat lunak memiliki banyak  rancangan diagram dan sumber daya manusia tidak  mencukupi. Pemeriksaan membutuhkan waktu yang lama dan memiliki dampak pada waktu pengembangan perangkat lunak. Penelitian ini mengusulkan pembuatan kakas bantu  untuk mendeteksi ketidaksesuaian diagram urutan dengan diagram kasus penggunaan. Ketidaksesuaian dilihat dari kemiripan semantik kalimat antara alur pada deskripsi kasus penggunaan dan triplet. Dari hasil pembuatan kakas bantu, kakas bantu yang dibuat dapat mendeteksi ketidaksesuaian diagram urutan dengan diagram kasus penggunaan. Kakas  bantu ini diharapkan tidak hanya membantu pemeriksaan rancangan diagram akan tetapi mempercepat waktu pengembangan perangkat lunak.

  9. From State Diagram to Class Diagram

    DEFF Research Database (Denmark)

    Borch, Ole; Madsen, Per Printz

    2009-01-01

    UML class diagram and Java source code are interrelated and Java code is a kind of interchange format. Working with UML state diagram in CASE tools, a corresponding xml file is maintained. Designing state diagrams is mostly performed manually using design patterns and coding templates - a time...... consuming process. This article demonstrates how to compile such a diagram into Java code and later, by reverse engineering, produce a class diagram. The process from state diagram via intermediate SAX parsed xml file to Apache Velocity generated Java code is described. The result is a fast reproducible...

  10. 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

  11. [Identification of meridian-acupoint diagrams and meridian diagrams].

    Science.gov (United States)

    Shen, Wei-hong

    2008-08-01

    In acu-moxibustion literature, there are two kinds of diagrams, meridian-acupoint diagrams and meridian diagrams. Because they are very similar in outline, and people now have seldom seen the typical ancient meridian diagrams, meridian-acupoint diagrams have been being incorrectly considered to be the meridian diagrams for a long time. It results in confusion in acu-moxibustion academia. The present paper stresses its importance in academic research and introduces some methods for identifying them correctly. The key points for identification of meridian-acupoint diagrams and meridian diagrams are: the legend of diagrams and the drawing style of the ancient charts. In addition, the author makes a detailed explanation about some acu-moxibustion charts which are easily confused. In order to distinguish meridian-acupoint diagrams and meridian diagrams correctly, he or she shoulnd understand the diagrams' intrinsic information as much as possible and make a comprehensive analysis about them.

  12. Reliability model for helicopter main gearbox lubrication system using influence diagrams

    International Nuclear Information System (INIS)

    Rashid, H.S.J.; Place, C.S.; Mba, D.; Keong, R.L.C.; Healey, A.; Kleine-Beek, W.; Romano, M.

    2015-01-01

    The loss of oil from a helicopter main gearbox (MGB) leads to increased friction between components, a rise in component surface temperatures, and subsequent mechanical failure of gearbox components. A number of significant helicopter accidents have been caused due to such loss of lubrication. This paper presents a model to assess the reliability of helicopter MGB lubricating systems. Safety risk modeling was conducted for MGB oil system related accidents in order to analyse key failure mechanisms and the contributory factors. Thus, the dominant failure modes for lubrication systems and key contributing components were identified. The Influence Diagram (ID) approach was then employed to investigate reliability issues of the MGB lubrication systems at the level of primary causal factors, thus systematically investigating a complex context of events, conditions, and influences that are direct triggers of the helicopter MGB lubrication system failures. The interrelationships between MGB lubrication system failure types were thus identified, and the influence of each of these factors on the overall MGB lubrication system reliability was assessed. This paper highlights parts of the HELMGOP project, sponsored by the European Aviation Safety Agency to improve helicopter main gearbox reliability. - Highlights: • We investigated methods to optimize helicopter MGB oil system run-dry capability. • Used Influence Diagram to assess design and maintenance factors of MGB oil system. • Factors influencing overall MGB lubrication system reliability were identified. • This globally influences current and future helicopter MGB designs

  13. Choice and complexation of techniques and tools for assessment of NPP I and C systems safety

    International Nuclear Information System (INIS)

    Illiashenko, Oleg; Babeshko, Eugene

    2011-01-01

    There are a lot of techniques to analyze and assess reliability and safety of NPP Instrumentation and Control (I and C) systems (e.g. FMEA - Failure Modes and Effects Analysis and its modifications, FTA - Fault Tree Analysis, HAZOP - Hazard and Operability Analysis, RBD - Reliability Block Diagram, Markov Models, etc.) and quantity of tools based on these techniques is constantly increasing. Known ways of safety assessment, as well as problems of their choice and complexation are analyzed. Objective of the paper is the development of general 'technique of techniques choosing' and tool for support of such technique. The following criteria are used for analysis and comparison and their features are described: compliance to normative documents; experience of application in industry; methods used for assessment of system NPP I and C safety; tool architecture/framework; reporting; vendor support, etc. Comparative analysis results of existing T and T - Tools and Techniques for safety analysis are presented in matrix form ('Tools-Criterion') with example. Features of complexation of different safety assessment techniques (FMECA, FTA, RBD, Markov Models) are described. The proposed technique is implemented as special tool for decision-making. The proposed technique was used for development of RPC Radiy company standard CS 66. This guide contains requirements and procedures of FMECA analysis of developed and produced NPP I and C systems based on RADIY platform. (author)

  14. Safety assessment for the IS process in a hydrogen production facility

    International Nuclear Information System (INIS)

    Cho, Nam Chul

    2005-08-01

    A substitute energy development have been required due to the dry up of the fossil fuel and an environmental problem. Consequently, among substitute energy to be discussed, producing hydrogen from water which does not release carbon is a very promising technology. Also, Iodine-Sulfur(IS) thermochemical water decomposition is one of the promising process which is used to produce hydrogen efficiently using the high temperature gas-cooled reactor(HTGR) as an energy source that is possible to supply heat over 1000 .deg. C. In this study, to make a safety assessment of the hydrogen production using the IS process, an initiating events analysis and an accident scenario modeling considering the relief system were carried out. A method for initiating event identification used the Master Logic Diagram(MLD) that is logical and deductive. As a result, 9 initiating events that cause a leakage of the chemical material were identified. 6 accident scenario based on the initiating event are identified and quantified to the event trees. The frequency of the chemical material leakage produced by IS process is estimated relatively high to the value of 1.22x10 -4 /y. Therefore, it requires more effort on safety of the hydrogen production which can be considered as a part of the nuclear system and safety management research to increase social acceptability. Moreover, these methods will be helpful to the safety assessment of the hydrogen production system of the IS process in general

  15. 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)

  16. 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

  17. Influence diagrams and decision trees for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Goetz, W.W.J.

    1996-09-01

    A review of relevant methodologies based on Influence Diagrams (IDs), Decision Trees (DTs), and Containment Event Trees (CETs) was conducted to assess the practicality of these methods for the selection of effective strategies for Severe Accident Management (SAM). The review included an evaluation of some software packages for these methods. The emphasis was on possible pitfalls of using IDs and on practical aspects, the latter by performance of a case study that was based on an existing Level 2 Probabilistic Safety Assessment (PSA). The study showed that the use of a combined ID/DT model has advantages over CET models, in particular when conservatisms in the Level 2 PSA have been identified and replaced by fair assessments of the uncertainties involved. It is recommended to use ID/DT models complementary to CET models. (orig.).

  18. Influence diagrams and decision trees for severe accident management

    International Nuclear Information System (INIS)

    Goetz, W.W.J.; Seebregts, A.J.; Bedford, T.J.

    1996-08-01

    A review of relevent methodologies based on Influence Diagrams (IDs), Decision Trees (DTs), and Containment Event Trees (CETs) was conducted to assess the practicality of these methods for the selection of effective strategies for Severe Accident Management (SAM). The review included an evaluation of some software packages for these methods. The emphasis was on possible pitfalls of using IDs and on practical aspects, the latter by performance of a case study that was based on an existing Level 2 Probabilistic Safety Assessment (PSA). The study showed that the use of a combined ID/DT model has advantages over CET models, in particular when conservatisms in the Level 2 PSA have been identified and replaced by fair assessments of the uncertainties involved. It is recommended to use ID/DT models as complementary to CET models. (orig.)

  19. Influence diagrams and decision trees for severe accident management

    International Nuclear Information System (INIS)

    Goetz, W.W.J.

    1996-09-01

    A review of relevant methodologies based on Influence Diagrams (IDs), Decision Trees (DTs), and Containment Event Trees (CETs) was conducted to assess the practicality of these methods for the selection of effective strategies for Severe Accident Management (SAM). The review included an evaluation of some software packages for these methods. The emphasis was on possible pitfalls of using IDs and on practical aspects, the latter by performance of a case study that was based on an existing Level 2 Probabilistic Safety Assessment (PSA). The study showed that the use of a combined ID/DT model has advantages over CET models, in particular when conservatisms in the Level 2 PSA have been identified and replaced by fair assessments of the uncertainties involved. It is recommended to use ID/DT models complementary to CET models. (orig.)

  20. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

    Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...

  1. The role of perceptual cues in matrix diagrams

    NARCIS (Netherlands)

    van der Meij, Jan; van Amelsvoort, Marije; Anjewierden, A.

    An experiment was conducted to assess whether the design of a matrix diagram influences how people study the diagram and whether this has an effect on recall of the presented information. We compared four versions of a matrix diagram on antisocial personality disorder. It consisted of four header

  2. The role of perceptual cues in matrix diagrams

    NARCIS (Netherlands)

    van der Meij, Jan; Amelsvoort, Marije; Anjewierden, Anjo Allert

    2015-01-01

    An experiment was conducted to assess whether the design of a matrix diagram influences how people study the diagram and whether this has an effect on recall of the presented information. We compared four versions of a matrix diagram on antisocial personality disorder. It consisted of four header

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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)

  8. Top-down versus bottom-up processing of influence diagrams in probabilistic analysis

    International Nuclear Information System (INIS)

    Timmerman, R.D.; Burns, T.J.; Dodds, H.L. Jr.

    1986-01-01

    Recent work by Phillips and Selby has shown that influence diagram methodology can be a useful analytical tool in reactor safety studies. In some instances an influence diagram can be used as a graphical representation of probabilistic dependence within a system or event sequence. Under these circumstances, Bayesian statistics is employed to transform the relationships depicted in the influence diagram into the correct expression for a desired marginal probability (e.g. the top node). Top-down and bottom-up algorithms have emerged as the dominant methods for quantifying influence diagrams. The purpose of this paper is to demonstrate a potential error in employing the bottom-up algorithm when dealing with interdependencies

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

    OpenAIRE

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

    2015-01-01

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

  10. 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

  11. 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

  12. 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.

  13. Top-down versus bottom-up processing of influence diagrams in probabilistic analysis

    International Nuclear Information System (INIS)

    Timmerman, R.D.; Burns, T.J.; Dodds, H.L. Jr.

    1984-01-01

    Recent work by Phillips et al., and Selby et al., has shown that influence diagram methodology can be a useful analytical tool in reactor safety studies. An influence diagram is a graphical representation of probabilistic dependence within a system or event sequence. Bayesian statistics are employed to transform the relationships depicted in the influence diagram into the correct expression for a desired marginal probability (e.g. the top event). As with fault trees, top-down and bottom-up algorithms have emerged as the dominant methods for quantifying influence diagrams. Purpose of this paper is to demonstrate a potential error in employing the bottom-up algorithm when dealing with interdependencies. In addition, the computing efficiency of both methods is discussed

  14. 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)

  15. Diagram Size vs. Layout Flaws: Understanding Quality Factors of UML Diagrams

    DEFF Research Database (Denmark)

    Störrle, Harald

    2016-01-01

    , though, is our third goal of extending our analysis aspects of diagram quality. Method: We improve our definition of diagram size and add a (provisional) definition of diagram quality as the number of topographic layout flaws. We apply these metrics on 60 diagrams of the five most commonly used types...... of UML diagram. We carefully analyze the structure of our diagram samples to ensure representativeness. We correlate diagram size and layout quality with modeler performance data obtained in previous experiments. The data set is the largest of its kind (n-156). Results: We replicate earlier findings......, and extend them to two new diagram types. We provide an improved definition of diagram size, and provide a definition of topographic layout quality, which is one more step towards a comprehensive definition of diagram quality as such. Both metrics are shown to be objectively applicable. We quantify...

  16. 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)

  17. A variable ordering heuristic for risk monitors based on zero-suppressed binary decision diagram

    International Nuclear Information System (INIS)

    Wang Jin; Wang Fang; Wang Jiaqun; Gu Xiaohui; Yuan Run; Li Yazhou; Hu Liqin; Wu Yican; Yin Yuan; FDS Team

    2010-01-01

    The probabilistic safety assessment model using for determining the instantaneous risk in a risk monitor are much more complex than the operational model using for determining the average risk. Therefore the development of a fast calculation engine is indispensable and challengeable. the scale of zero-suppressed binary decision diagram. In this paper an optimized method of variables ordering, which fully utilized that features of operational model, was proposed.Not only the theoretical demonstration but also the applications were also brought out. (authors)

  18. 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)

  19. 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)

  20. 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)

  1. 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

  2. Roundhouse Diagrams.

    Science.gov (United States)

    Ward, Robin E.; Wandersee, James

    2000-01-01

    Students must understand key concepts through reasoning, searching out related concepts, and making connections within multiple systems to learn science. The Roundhouse diagram was developed to be a concise, holistic, graphic representation of a science topic, process, or activity. Includes sample Roundhouse diagrams, a diagram checklist, and…

  3. 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.

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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)

  9. Comparison of residual strength-grounding damage index diagrams for tankers produced by the ALPS/HULL ISFEM and design formula method

    Directory of Open Access Journals (Sweden)

    Do Kyun Kim

    2013-03-01

    Full Text Available This study compares the Residual ultimate longitudinal strength – grounding Damage index (R-D diagrams produced by two analysis methods: the ALPS/HULL Intelligent Supersize Finite Element Method (ISFEM and the design formula (modified Paik and Mansour method – used to assess the safety of damaged ships. The comparison includes four types of double-hull oil tankers: Panamax, Aframax, Suezmax and VLCC. The R-D diagrams were calculated for a series of 50 grounding scenarios. The diagrams were efficiently sampled using the Latin Hypercube Sampling (LHS technique and comprehensively analysed based on ship size. Finally, the two methods were compared by statistically analysing the differences between their grounding damage indices and ultimate longitudinal strength predictions. The findings provide a useful example of how to apply the ultimate longitudinal strength analysis method to grounded ships.

  10. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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.

  15. 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

  16. 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)

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. An Application of Mosaic Diagrams to the Visualization of Set Relationships

    OpenAIRE

    Luz, Saturnino; Masoodian, Masood

    2017-01-01

    We present an application of mosaic diagrams to the visualisation of set relations. Venn and Euler diagrams are the best known visual representations of sets and their relationships (intersections, containment or subsets, exclusion or disjointness). In recent years, alternative forms of visualisation have been proposed. Among them, linear diagrams have been shown to compare favourably to Venn and Euler diagrams, in supporting non-interactive assessment of set relationships. Recent studies tha...

  2. 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

  3. MORT User's Manual for use with the Management Oversight and Risk Tree analytical logic diagram. [Contains a list of System Safety Development Center publications

    Energy Technology Data Exchange (ETDEWEB)

    Knox, N.W.; Eicher, R.W.

    1992-02-01

    This report contains the User's Manual for MORT (Management Oversight and Risk Tree), a logic diagram in the form of a work sheet'' that illustrates a long series of interrelated questions. MORT is a comprehensive analytical procedure that provides a disciplined method for determining the causes and contributing factors of major accidents. Alternatively, it serves as a tool to evaluate the quality of an existing system. While similar in many respects to fault tree analysis, MORT is more generalized and presents over 1,500 specific elements of an ideal universal'' management program for optimizing environment, safety and health, and other programs. This User's Manual is intended to be used with the MORT diagram dated February 1992.

  4. Master Logic Diagram: An Approach to Identify Initiating Events of HTGRs

    Science.gov (United States)

    Purba, J. H.

    2018-02-01

    Initiating events of a nuclear power plant being evaluated need to be firstly identified prior to applying probabilistic safety assessment on that plant. Various types of master logic diagrams (MLDs) have been proposedforsearching initiating events of the next generation of nuclear power plants, which have limited data and operating experiences. Those MLDs are different in the number of steps or levels and different in the basis for developing them. This study proposed another type of MLD approach to find high temperature gas cooled reactor (HTGR) initiating events. It consists of five functional steps starting from the top event representing the final objective of the safety functions to the basic event representing the goal of the MLD development, which is an initiating event. The application of the proposed approach to search for two HTGR initiating events, i.e. power turbine generator trip and loss of offsite power, is provided. The results confirmed that the proposed MLD is feasiblefor finding HTGR initiating events.

  5. 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)

  6. 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

  7. Mollier-h,x diagram for moist flue gas

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H; Hultsch, T; Suder, M

    1984-07-01

    Diagrams and formulae are presented for calculation of enthalpy and moisture content of flue gas from brown coal, heating oil, black coal and brown coal briquet combustion. The enthalpy (in kJ/kg) and moisture (g/kg) diagrams were established by computer graphics for pressure 0.1 MPa. A further diagram is provided for enthalpy and flue gas moisture, varying the combustion air supply according to coal dust and to grate firing. These thermodynamic calculations are regarded as significant for assessing methods of flue gas cooling below the moisture dew point and for waste heat recovery. 3 references.

  8. 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

  9. 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

  10. The importance of design in learning from node-link diagrams

    NARCIS (Netherlands)

    Amelsvoort, Marije; van der Meij, Jan; Anjewierden, Anjo Allert; van der Meij, Hans

    2013-01-01

    Diagrams organize by location. They give spatial cues for finding and recognizing information and for making inferences. In education, diagrams are often used to help students understand and recall information. This study assessed the influence of perceptual cues on reading behavior and subsequent

  11. Algorithmic phase diagrams

    Science.gov (United States)

    Hockney, Roger

    1987-01-01

    Algorithmic phase diagrams are a neat and compact representation of the results of comparing the execution time of several algorithms for the solution of the same problem. As an example, the recent results are shown of Gannon and Van Rosendale on the solution of multiple tridiagonal systems of equations in the form of such diagrams. The act of preparing these diagrams has revealed an unexpectedly complex relationship between the best algorithm and the number and size of the tridiagonal systems, which was not evident from the algebraic formulae in the original paper. Even so, for a particular computer, one diagram suffices to predict the best algorithm for all problems that are likely to be encountered the prediction being read directly from the diagram without complex calculation.

  12. 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

  13. 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

  14. 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.

  15. 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)

  16. 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

  17. Assessment of redundant systems with imperfect coverage by means of binary decision diagrams

    Energy Technology Data Exchange (ETDEWEB)

    Myers, Albert F. [Northrop Grumman Corporation, 1840 Century Park East, Los Angeles, CA 90067-2199 (United States)], E-mail: Al.Myers@ngc.com; Rauzy, Antoine [IML/CNRS, 163, Avenue de Luminy, 13288 Marseille Cedex 09 (France)], E-mail: arauzy@iml.univ-mrs.fr

    2008-07-15

    In this article, we study the assessment of the reliability of redundant systems with imperfect fault coverage. We term fault coverage as the ability of a system to isolate and correctly accommodate failures of redundant elements. For highly reliable systems, such as avionic and space systems, fault coverage is in general imperfect and has a significant impact on system reliability. We review here the different models of imperfect fault coverage. We propose efficient algorithms to assess them separately (as k-out-of-n selectors). We show how to implement these algorithms into a binary decision diagrams engine. Finally, we report experimental results on real life test cases that show on the one hand the importance of imperfect coverage and on the other hand the efficiency of the proposed approach.

  18. 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)

  19. Electronic diagrams

    CERN Document Server

    Colwell, Morris A

    1976-01-01

    Electronic Diagrams is a ready reference and general guide to systems and circuit planning and in the preparation of diagrams for both newcomers and the more experienced. This book presents guidelines and logical procedures that the reader can follow and then be equipped to tackle large complex diagrams by recognition of characteristic 'building blocks' or 'black boxes'. The goal is to break down many of the barriers that often seem to deter students and laymen in learning the art of electronics, especially when they take up electronics as a spare time occupation. This text is comprised of nin

  20. 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

  1. Viral pathogenesis in diagrams

    National Research Council Canada - National Science Library

    Tremblay, Michel; Berthiaume, Laurent; Ackermann, Hans-Wolfgang

    2001-01-01

    .... The 268 diagrams in Viral Pathogenesis in Diagrams were selected from over 800 diagrams of English and French virological literature, including one derived from a famous drawing by Leonardo da Vinci...

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. 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

  7. Methodology used in the integrated assessment of PIUS-600 safety

    International Nuclear Information System (INIS)

    Fullwood, R.; Higgins, J.; Kroegar, P.

    1993-01-01

    The revolutionary reactor design, PIUS-600 as described in the Preliminary Safety Analysis Report (PSID) was subjected to analysis consisting of Failure Modes, Effects and Criticality Analysis (FMECA), Hazards and Operability (HAZOP) analysis, and conventional engineering review of the stress, neutronics, thermal hydraulics, and corrosion. These results were integrated in the PIUS Intermediate Table (PIT) from which accident initiators and mitigators were identified and categorized into seven estimated frequency intervals. Accident consequences were classified as: CC-1, minor radiological release, CC-2, clad release, CC-3, major release. The systems were analyzed using event sequence diagrams (ESDs) and event trees (ETs). The resulting accident sequences of the ET, were categorized into Event conditions (ECs) based on initiator frequency and combinations of failures. System interactions were considered in the FMECAs, ESDs, ETs and in an interaction table that also identified system safety classifications

  8. Methodology used in the integrated assessment of PIUS-600 safety

    International Nuclear Information System (INIS)

    Fullwood, R.; Higgins, J.; Kroeger, P.

    1993-01-01

    The revolutionary reactor design, PIUS-600 as described in the Preliminary Safety Analysis Report (PSID) was subject to analyses consisting of Failure Modes. Effects and Criticality Analysis (FMECA), Hazards and Operability (HAZOP) analysis, and conventional engineering review of the stress, neutronics, thermal hydraulics, and corrosion. These results were integrated in the PIUS Intermediate Table (PIT) from which accident initiators and mitigators were identified and categorized into seven estimated frequency intervals. Accident consequences were classified as: CC-1, minor radiological release, CC-2, clad release, CC-3, major release. The systems were analyzed using event sequence diagrams (ESDs) and event trees (ETs). The resulting accident sequences of the ET, were categorized into Event conditions (ECs) based on initiator frequency and combinations of failures. System interactions were considered in the FMECAs, ESDs, ETs and in an interaction table that also identified system safety classifications

  9. 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

  10. 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)

  11. 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)

  12. Diagram, a Learning Environment for Initiation to Object-Oriented Modeling with UML Class Diagrams

    Science.gov (United States)

    Py, Dominique; Auxepaules, Ludovic; Alonso, Mathilde

    2013-01-01

    This paper presents Diagram, a learning environment for object-oriented modelling (OOM) with UML class diagrams. Diagram an open environment, in which the teacher can add new exercises without constraints on the vocabulary or the size of the diagram. The interface includes methodological help, encourages self-correcting and self-monitoring, and…

  13. VennDiagram: a package for the generation of highly-customizable Venn and Euler diagrams in R

    Directory of Open Access Journals (Sweden)

    Boutros Paul C

    2011-01-01

    Full Text Available Abstract Background Visualization of orthogonal (disjoint or overlapping datasets is a common task in bioinformatics. Few tools exist to automate the generation of extensively-customizable, high-resolution Venn and Euler diagrams in the R statistical environment. To fill this gap we introduce VennDiagram, an R package that enables the automated generation of highly-customizable, high-resolution Venn diagrams with up to four sets and Euler diagrams with up to three sets. Results The VennDiagram package offers the user the ability to customize essentially all aspects of the generated diagrams, including font sizes, label styles and locations, and the overall rotation of the diagram. We have implemented scaled Venn and Euler diagrams, which increase graphical accuracy and visual appeal. Diagrams are generated as high-definition TIFF files, simplifying the process of creating publication-quality figures and easing integration with established analysis pipelines. Conclusions The VennDiagram package allows the creation of high quality Venn and Euler diagrams in the R statistical environment.

  14. VennDiagram: a package for the generation of highly-customizable Venn and Euler diagrams in R.

    Science.gov (United States)

    Chen, Hanbo; Boutros, Paul C

    2011-01-26

    Visualization of orthogonal (disjoint) or overlapping datasets is a common task in bioinformatics. Few tools exist to automate the generation of extensively-customizable, high-resolution Venn and Euler diagrams in the R statistical environment. To fill this gap we introduce VennDiagram, an R package that enables the automated generation of highly-customizable, high-resolution Venn diagrams with up to four sets and Euler diagrams with up to three sets. The VennDiagram package offers the user the ability to customize essentially all aspects of the generated diagrams, including font sizes, label styles and locations, and the overall rotation of the diagram. We have implemented scaled Venn and Euler diagrams, which increase graphical accuracy and visual appeal. Diagrams are generated as high-definition TIFF files, simplifying the process of creating publication-quality figures and easing integration with established analysis pipelines. The VennDiagram package allows the creation of high quality Venn and Euler diagrams in the R statistical environment.

  15. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  16. 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

  17. Introduction to Feynman diagrams

    CERN Document Server

    Bilenky, Samoil Mikhelevich

    1974-01-01

    Introduction to Feynman Diagrams provides Feynman diagram techniques and methods for calculating quantities measured experimentally. The book discusses topics Feynman diagrams intended for experimental physicists. Topics presented include methods for calculating the matrix elements (by perturbation theory) and the basic rules for constructing Feynman diagrams; techniques for calculating cross sections and polarizations; processes in which both leptons and hadrons take part; and the electromagnetic and weak form factors of nucleons. Experimental physicists and graduate students of physics will

  18. Study on Safety Assessment for TINT- Pre disposal Radioactive Waste Management Facilities by the Application of SAFRAN Software

    International Nuclear Information System (INIS)

    Ya-anant, Nanthavan

    2011-06-01

    Full text: The Radioactive Waste Management Center, Thailand Institute of Nuclear Technology (TINT) provides a centralized radioactive waste management (RWM) service in the country. The pre disposal RWM facilities are composed of low and intermediate level waste treatment and storage facilities. The benefits of this study are (1) to improve the safety of pre disposal RWM facilities (2) to experience with the SAFRAN software tool for the safety assessment of pre disposal RWM facilities, which has been developed following to the methodology from International Atomic Energy Agency (IAEA). The work was performed on collecting all waste management data, the diagram of facilities, buildings, location, procedure, waste classification, waste form, radiological/chemical/physical properties including scenarios in normal and accidental conditions. The result of normal condition is that the effective dose per year of worker and public is less than 20 mSv and 1 mSv respectively. So the TINT-RWM operation is safe, as referred to the regulation

  19. Pourbaix Diagrams at Elevated Temperatures A Study of Zinc and Tin

    Science.gov (United States)

    Palazhchenko, Olga

    Metals in industrial settings such as power plants are often subjected to high temperature and pressure aqueous environments, where failure to control corrosion compromises worker and environment safety. For instance, zircaloy (1.2-1.7 wt.% Sn) fuel rods are exposed to aqueous 250-310 °C coolant in CANDU reactors. The Pourbaix (EH-pH) diagram is a plot of electrochemical potential versus pH, which shows the domains of various metal species and by inference, corrosion susceptibility. Elevated temperature data for tin +II and tin +IV species were obtained using solid-aqueous phase equilibria with the respective oxides, in a batch vessel with in-situ pH measurement. Solubilities, determined via spectroscopic techniques, were used to calculate equilibrium constants and the Gibbs energies of Sn complexes for E-pH diagram construction. The SnOH3+ and Sn(OH )-5 species were incorporated, for the first time, into the 298.15 K and 358.15 K diagrams, with novel Go values determined at 358.15 K. Key words: Pourbaix diagrams, EH-pH, elevated temperatures, solubility, equilibrium, metal oxides, hydrolysis, redox potential, pH, thermochemical data, tin, zinc, zircaloy, corrosion, passivity.

  20. 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

  1. Criteria for evaluating programme theory diagrams in quality improvement initiatives: a structured method for appraisal.

    Science.gov (United States)

    Issen, Laurel; Woodcock, Thomas; McNicholas, Christopher; Lennox, Laura; Reed, Julie E

    2018-04-09

    Despite criticisms that many quality improvement (QI) initiatives fail due to incomplete programme theory, there is no defined way to evaluate how programme theory has been articulated. The objective of this research was to develop, and assess the usability and reliability of scoring criteria to evaluate programme theory diagrams. Criteria development was informed by published literature and QI experts. Inter-rater reliability was tested between two evaluators. About 63 programme theory diagrams (42 driver diagrams and 21 action-effect diagrams) were reviewed to establish whether the criteria could support comparative analysis of different approaches to constructing diagrams. Components of the scoring criteria include: assessment of overall aim, logical overview, clarity of components, cause-effect relationships, evidence and measurement. Independent reviewers had 78% inter-rater reliability. Scoring enabled direct comparison of different approaches to developing programme theory; action-effect diagrams were found to have had a statistically significant but moderate improvement in programme theory quality over driver diagrams; no significant differences were observed based on the setting in which driver diagrams were developed. The scoring criteria summarise the necessary components of programme theory that are thought to contribute to successful QI projects. The viability of the scoring criteria for practical application was demonstrated. Future uses include assessment of individual programme theory diagrams and comparison of different approaches (e.g. methodological, teaching or other QI support) to produce programme theory. The criteria can be used as a tool to guide the production of better programme theory diagrams, and also highlights where additional support for QI teams could be needed.

  2. 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

  3. Diagrams benefit symbolic problem-solving.

    Science.gov (United States)

    Chu, Junyi; Rittle-Johnson, Bethany; Fyfe, Emily R

    2017-06-01

    The format of a mathematics problem often influences students' problem-solving performance. For example, providing diagrams in conjunction with story problems can benefit students' understanding, choice of strategy, and accuracy on story problems. However, it remains unclear whether providing diagrams in conjunction with symbolic equations can benefit problem-solving performance as well. We tested the impact of diagram presence on students' performance on algebra equation problems to determine whether diagrams increase problem-solving success. We also examined the influence of item- and student-level factors to test the robustness of the diagram effect. We worked with 61 seventh-grade students who had received 2 months of pre-algebra instruction. Students participated in an experimenter-led classroom session. Using a within-subjects design, students solved algebra problems in two matched formats (equation and equation-with-diagram). The presence of diagrams increased equation-solving accuracy and the use of informal strategies. This diagram benefit was independent of student ability and item complexity. The benefits of diagrams found previously for story problems generalized to symbolic problems. The findings are consistent with cognitive models of problem-solving and suggest that diagrams may be a useful additional representation of symbolic problems. © 2017 The British Psychological Society.

  4. Development and application of α-hull and Voronoi diagrams in the assessment of roundness error

    International Nuclear Information System (INIS)

    Li, Xiuming; Liu, Hongqi; Li, Wei

    2011-01-01

    Computational geometry has been used to select effective data points from the measured data points for evaluating the roundness error to improve the computational complexity. However, for precision parts most of the measured points are on the vertices of the convex hull; it cannot have any effect on improving the computational complexity with the Voronoi diagrams. In this paper the roundness error is evaluated with α-hull and the Voronoi diagram instead of convex hull. An approach for constructing α-hull with the minimum radius separation is presented to determine the vertices of the Voronoi diagram. The experimental results showed that the roundness error of the minimum zone circle could be solved efficiently with α-hull and the Voronoi diagram

  5. 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)

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. Optimizing UML Class Diagrams

    Directory of Open Access Journals (Sweden)

    Sergievskiy Maxim

    2018-01-01

    Full Text Available Most of object-oriented development technologies rely on the use of the universal modeling language UML; class diagrams play a very important role in the design process play, used to build a software system model. Modern CASE tools, which are the basic tools for object-oriented development, can’t be used to optimize UML diagrams. In this manuscript we will explain how, based on the use of design patterns and anti-patterns, class diagrams could be verified and optimized. Certain transformations can be carried out automatically; in other cases, potential inefficiencies will be indicated and recommendations given. This study also discusses additional CASE tools for validating and optimizing of UML class diagrams. For this purpose, a plugin has been developed that analyzes an XMI file containing a description of class diagrams.

  13. 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)

  14. Qualitative Risk Assessment for an LNG Refueling Station and Review of Relevant Safety Issues

    Energy Technology Data Exchange (ETDEWEB)

    Siu, N.; Herring, J.S.; Cadwallader, L.; Reece, W.; Byers, J.

    1998-02-01

    This report is a qualitative assessment of the public and worker risk involved with the operation of a liquefied natural gas (LNG) vehicle refueling facility. This study includes facility maintenance and operations, tank truck deliveries, and end-use vehicle fueling; it does not treat the risks of LNG vehicles on roadways. Accident initiating events are identified by using a Master Logic Diagram, a Failure Modes and Effects Analysis, and historical operating experiences. The event trees were drawn to depict possible sequences of mitigating events following the initiating events. The phenomenology of LNG and other vehicle fuels is discussed to characterize the hazard posed by LNG usage. Based on the risk modeling and analysis, recommendations are given to improve the safety of LNG refueling stations in the areas of procedures and training, station design, and the dissemination of ``best practice`` information throughout the LNG community.

  15. Interim qualitative risk assessment for an LNG refueling station and review of relevant safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Siu, N.; Herring, S.; Cadwallader, L.; Reece, W.; Byers, J.

    1997-07-01

    This report is a qualitative assessment of the public and worker risk involved with the operation of a liquefied natural (LNG) vehicle refueling facility. This study includes facility maintenance and operations, tanker truck delivers and end-use vehicle fueling; it does not treat the risks of LNG vehicles on roadways. Accident initiating events are identified by using a Master Logic Diagram, a Failure Modes and Effects analysis and historical operating experiences. The event trees were drawn to depict possible sequences of mitigating events following the initiating events. The phenomenology of LNG and other vehicle fuels is discussed to characterize the hazard posed by LNG usage. Based on the risk modeling and analysis, recommendations are given to improve the safety of LNG refueling stations in the areas of procedures and training, station design, and the dissemination of best practice information throughout the LNG community.

  16. 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)

  17. Study on structural seismic margin and probabilistic seismic risk. Development of a structural capacity-seismic risk diagram

    International Nuclear Information System (INIS)

    Nakajima, Masato; Ohtori, Yasuki; Hirata, Kazuta

    2010-01-01

    Seismic margin is extremely important index and information when we evaluate and account seismic safety of critical structures, systems and components quantitatively. Therefore, it is required that electric power companies evaluate the seismic margin of each plant in back-check of nuclear power plants in Japan. The seismic margin of structures is usually defined as a structural capacity margin corresponding to design earthquake ground motion. However, there is little agreement as to the definition of the seismic margin and we have no knowledge about a relationship between the seismic margin and seismic risk (annual failure probability) which is obtained in PSA (Probabilistic Safety Assessment). The purpose of this report is to discuss a definition of structural seismic margin and to develop a diagram which can identify a relation between seismic margin and seismic risk. The main results of this paper are described as follows: (1) We develop seismic margin which is defined based on the fact that intensity of earthquake ground motion is more appropriate than the conventional definition (i.e., the response-based seismic margin) for the following reasons: -seismic margin based on earthquake ground motion is invariant where different typed structures are considered, -stakeholders can understand the seismic margin based on the earthquake ground motion better than the response-based one. (2) The developed seismic margin-risk diagram facilitates us to judge easily whether we need to perform detailed probabilistic risk analysis or only deterministic analysis, given that the reference risk level although information on the uncertainty parameter beta is not obtained. (3) We have performed numerical simulations based on the developed method for four sites in Japan. The structural capacity-risk diagram differs depending on each location because the diagram is greatly influenced by seismic hazard information for a target site. Furthermore, the required structural capacity

  18. 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)

  19. 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

  20. Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

    Energy Technology Data Exchange (ETDEWEB)

    Groth, Katrina, E-mail: kgroth@umd.ed [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States); Wang Chengdong; Mosleh, Ali [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States)

    2010-12-15

    This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

  1. Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

    International Nuclear Information System (INIS)

    Groth, Katrina; Wang Chengdong; Mosleh, Ali

    2010-01-01

    This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

  2. 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

  3. 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)

  4. 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)

  5. Stage line diagram: An age-conditional reference diagram for tracking development

    NARCIS (Netherlands)

    Buuren, S. van; Ooms, J.C.L.

    2009-01-01

    This paper presents a method for calculating stage line diagrams, a novel type of reference diagram useful for tracking developmental processes over time. Potential fields of applications include: dentistry (tooth eruption), oncology (tumor grading, cancer staging), virology (HIV infection and

  6. Stage line diagram: an age-conditional reference diagram for tracking development.

    NARCIS (Netherlands)

    Van Buuren, S.; Ooms, J.C.L.

    2009-01-01

    This paper presents a method for calculating stage line diagrams, a novel type of reference diagram useful for tracking developmental processes over time. Potential fields of applications include: dentistry (tooth eruption), oncology (tumor grading, cancer staging), virology (HIV infection and

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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)

  12. CERPHASE: Computer-generated phase diagrams

    International Nuclear Information System (INIS)

    Ruys, A.J.; Sorrell, C.C.; Scott, F.H.

    1990-01-01

    CERPHASE is a collection of computer programs written in the programming language basic and developed for the purpose of teaching the principles of phase diagram generation from the ideal solution model of thermodynamics. Two approaches are used in the generation of the phase diagrams: freezing point depression and minimization of the free energy of mixing. Binary and ternary phase diagrams can be generated as can diagrams containing the ideal solution parameters used to generate the actual phase diagrams. Since the diagrams generated utilize the ideal solution model, data input required from the operator is minimal: only the heat of fusion and melting point of each component. CERPHASE is menu-driven and user-friendly, containing simple instructions in the form of screen prompts as well as a HELP file to guide the operator. A second purpose of CERPHASE is in the prediction of phase diagrams in systems for which no experimentally determined phase diagrams are available, enabling the estimation of suitable firing or sintering temperatures for otherwise unknown systems. Since CERPHASE utilizes ideal solution theory, there are certain limitations imposed on the types of systems that can be predicted reliably. 6 refs., 13 refs

  13. The TbBr3–LiBr binary system: Experimental thermodynamic investigation and assessment of phase diagram

    International Nuclear Information System (INIS)

    Rycerz, L.; Gong, W.; Gaune-Escard, M.

    2013-01-01

    Highlights: ► DSC measurements for the (LiBr + TbBr 3 ) system. ► congruently Li3TbBr 6 and incongruently melting Li5TbBr 8 compounds. ► Thermodynamic description of the liquid phase in the (LiBr + TbBr 3 ) system. ► Assessment with a two-sublattice ionic solution model. - Abstract: DSC was used to study the phase equilibrium in the TbBr 3 –LiBr binary system. The results obtained provided a basis for constructing the phase diagram of this system. It exhibits two compounds: Li 5 TbBr 8 , which decomposes in the solid state at 611 K, and Li 3 TbBr 6 , which melts congruently at 785 K with the related enthalpy 59.1 kJ·mol −1 . The binary LiBr–TbBr 3 system was then optimized using the available experimental information on phase diagram and thermodynamic properties. A two-sub-lattice ionic solution model (Li + ) P :(Br − , TbBr 6 −3 , TbBr 3 ) Q was adopted to describe the liquid phase. The present assessment of the binary LiBr–TbBr 3 system was in good agreement with the corresponding experimental data and confirmed their consistency.

  14. 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

  15. 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

  16. ASSESSING STUDENTS’ COMPETENCE IN DEVELOPING CHOROPLETH MAPS COMBINED WITH DIAGRAM MAPS

    Directory of Open Access Journals (Sweden)

    GABRIELA OSACI-COSTACHE

    2015-01-01

    Full Text Available Choropleth maps combined with diagram maps are frequently used in geography. For this reason, based on the maps developed by students, the study aims at the following: identifying and analyzing the errors made by the students; establishing and analyzing the competence level of the students; identifying the causes that led to these errors; and finding the best solutions to improve both the educational process aiming at the formation of this kind of competences and the students’ results. The map assessment was accomplished during two academic years (2013-2014 and 2014-2015, in the aftermath of the activities meant to train the competence. We assessed 105 maps prepared by the students in Cartography (Faculty of Geography, University of Bucharest based on an analytical evaluation grid, with dichotomous scale, comprising 15 criteria. This tool helped us identify the errors made by the students, as well as their competence level. By applying a questionnaire, we identified the source of the errors from the students’ perspective, while by comparing the errors and the competence levels at the end of the two academic years we were able to come up with potential solutions for the improvement of the teaching and learning process.

  17. 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

  18. Para-equilibrium phase diagrams

    International Nuclear Information System (INIS)

    Pelton, Arthur D.; Koukkari, Pertti; Pajarre, Risto; Eriksson, Gunnar

    2014-01-01

    Highlights: • A rapidly cooled system may attain a state of para-equilibrium. • In this state rapidly diffusing elements reach equilibrium but others are immobile. • Application of the Phase Rule to para-equilibrium phase diagrams is discussed. • A general algorithm to calculate para-equilibrium phase diagrams is described. - Abstract: If an initially homogeneous system at high temperature is rapidly cooled, a temporary para-equilibrium state may result in which rapidly diffusing elements have reached equilibrium but more slowly diffusing elements have remained essentially immobile. The best known example occurs when homogeneous austenite is quenched. A para-equilibrium phase assemblage may be calculated thermodynamically by Gibbs free energy minimization under the constraint that the ratios of the slowly diffusing elements are the same in all phases. Several examples of calculated para-equilibrium phase diagram sections are presented and the application of the Phase Rule is discussed. Although the rules governing the geometry of these diagrams may appear at first to be somewhat different from those for full equilibrium phase diagrams, it is shown that in fact they obey exactly the same rules with the following provision. Since the molar ratios of non-diffusing elements are the same in all phases at para-equilibrium, these ratios act, as far as the geometry of the diagram is concerned, like “potential” variables (such as T, pressure or chemical potentials) rather than like “normal” composition variables which need not be the same in all phases. A general algorithm to calculate para-equilibrium phase diagrams is presented. In the limit, if a para-equilibrium calculation is performed under the constraint that no elements diffuse, then the resultant phase diagram shows the single phase with the minimum Gibbs free energy at any point on the diagram; such calculations are of interest in physical vapor deposition when deposition is so rapid that phase

  19. 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

  20. 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.

  1. 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...

  2. 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

  3. 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.

  4. 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)

  5. Sintering diagrams of UO2

    International Nuclear Information System (INIS)

    Mohan, A.; Soni, N.C.; Moorthy, V.K.

    1979-01-01

    Ashby's method (see Acta Met., vol. 22, p. 275, 1974) of constructing sintering diagrams has been modified to obtain contribution diagrams directly from the computer. The interplay of sintering variables and mechanisms are studied and the factors that affect the participation of mechanisms in UO 2 are determined. By studying the physical properties, it emerges that the order of inaccuracies is small in most cases and do not affect the diagrams. On the other hand, even a 10% error in activation energies, which is quite plausible, would make a significant difference to the diagram. The main criticism of Ashby's approach is that the numerous properties and equations used, communicate their inaccuracies to the diagrams and make them unreliable. The present study has considerably reduced the number of factors that need to be refined to make the sintering diagrams more meaningful. (Auth.)

  6. 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)

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. Feynman diagram drawing made easy

    International Nuclear Information System (INIS)

    Baillargeon, M.

    1997-01-01

    We present a drawing package optimised for Feynman diagrams. These can be constructed interactively with a mouse-driven graphical interface or from a script file, more suitable to work with a diagram generator. It provides most features encountered in Feynman diagrams and allows to modify every part of a diagram after its creation. Special attention has been paid to obtain a high quality printout as easily as possible. This package is written in Tcl/Tk and in C. (orig.)

  17. 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

  18. 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

  19. 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.

  20. Family history assessment for colorectal cancer (CRC) risk analysis - comparison of diagram- and questionnaire-based web interfaces.

    Science.gov (United States)

    Schultz, Michael; Seo, Steven Bohwan; Holt, Alec; Regenbrecht, Holger

    2015-11-18

    Colorectal cancer (CRC) has a high incidence, especially in New Zealand. The reasons for this are unknown. While most cancers develop sporadically, a positive family history, determined by the number and age at diagnosis of affected first and second degree relatives with CRC is one of the major factors, which may increase an individual's lifetime risk. Before a patient can be enrolled in a surveillance program a detailed assessment and documentation of the family history is important but time consuming and often inaccurate. The documentation is usually paper-based. Our aim was therefore to develop and validate the usability and efficacy of a web-based family history assessment tool for CRC suitable for the general population. The tool was also to calculate the risk and make a recommendation for surveillance. Two versions of an electronic assessment tool, diagram-based and questionnaire-based, were developed with the risk analysis and recommendations for surveillance based on the New Zealand Guidelines Group recommendations. Accuracy of our tool was tested prior to the study by comparing risk calculations based on family history by experienced gastroenterologists with the electronic assessment. The general public, visiting a local science fair were asked to use and comment on the usability of the two interfaces. Ninety people assessed and commented on the two interfaces. Both interfaces were effective in assessing the risk to develop CRC through their familial history for CRC. However, the questionnaire-based interface performed with significantly better satisfaction (p = 0.001) than the diagram-based interface. There was no difference in efficacy though. We conclude that a web-based questionnaire tool can assist in the accurate documentation and analysis of the family history relevant to determine the individual risk of CRC based on local guidelines. The calculator is now implemented and assessable through the web-page of a local charity for colorectal cancer

  1. The Semiotic Structure of Geometry Diagrams: How Textbook Diagrams Convey Meaning

    Science.gov (United States)

    Dimmel, Justin K.; Herbst, Patricio G.

    2015-01-01

    Geometry diagrams use the visual features of specific drawn objects to convey meaning about generic mathematical entities. We examine the semiotic structure of these visual features in two parts. One, we conduct a semiotic inquiry to conceptualize geometry diagrams as mathematical texts that comprise choices from different semiotic systems. Two,…

  2. 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

  3. Stage line diagram: an age-conditional reference diagram for tracking development.

    Science.gov (United States)

    van Buuren, Stef; Ooms, Jeroen C L

    2009-05-15

    This paper presents a method for calculating stage line diagrams, a novel type of reference diagram useful for tracking developmental processes over time. Potential fields of applications include: dentistry (tooth eruption), oncology (tumor grading, cancer staging), virology (HIV infection and disease staging), psychology (stages of cognitive development), human development (pubertal stages) and chronic diseases (stages of dementia). Transition probabilities between successive stages are modeled as smoothly varying functions of age. Age-conditional references are calculated from the modeled probabilities by the mid-P value. It is possible to eliminate the influence of age by calculating standard deviation scores (SDS). The method is applied to the empirical data to produce reference charts on secondary sexual maturation. The mean of the empirical SDS in the reference population is close to zero, whereas the variance depends on age. The stage line diagram provides quick insight into both status (in SDS) and tempo (in SDS/year) of development of an individual child. Other measures (e.g. height SDS, body mass index SDS) from the same child can be added to the chart. Diagrams for sexual maturation are available as a web application at http://vps.stefvanbuuren.nl/puberty. The stage line diagram expresses status and tempo of discrete changes on a continuous scale. Wider application of these measures scores opens up new analytic possibilities. (c) 2009 John Wiley & Sons, Ltd.

  4. An influence diagram for urban flood risk assessment through pluvial flood hazards under non-stationary conditions

    DEFF Research Database (Denmark)

    Åström, Helena Lisa Alexandra; Friis Hansen, P.; Garrè, Luca

    2014-01-01

    Urban flooding introduces significant risk to society. Non-stationarity leads to increased uncertainty and this is challenging to include in actual decision-making. The primary objective of this study was to develop a risk assessment and decision support framework for pluvial urban flood risk under...... non-stationary conditions using an influence diagram (ID) which is a Bayesian network (BN) extended with decision and utility nodes. Non-stationarity is considered to be the influence of climate change where extreme precipitation patterns change over time. The overall risk is quantified in monetary...... terms expressed as expected annual damage. The network is dynamic in as much as it assesses risk at different points in time. The framework provides means for decision-makers to assess how different decisions on flood adaptation affect the risk now and in the future. The result from the ID was extended...

  5. 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.

  6. Ring diagrams and phase transitions

    International Nuclear Information System (INIS)

    Takahashi, K.

    1986-01-01

    Ring diagrams at finite temperatures carry most infrared-singular parts among Feynman diagrams. Their effect to effective potentials are in general so significant that one must incorporate them as well as 1-loop diagrams. The author expresses these circumstances in some examples of supercooled phase transitions

  7. Web-Enabled Mechanistic Case Diagramming: A Novel Tool for Assessing Students' Ability to Integrate Foundational and Clinical Sciences.

    Science.gov (United States)

    Ferguson, Kristi J; Kreiter, Clarence D; Haugen, Thomas H; Dee, Fred R

    2018-02-20

    As medical schools move from discipline-based courses to more integrated approaches, identifying assessment tools that parallel this change is an important goal. The authors describe the use of test item statistics to assess the reliability and validity of web-enabled mechanistic case diagrams (MCDs) as a potential tool to assess students' ability to integrate basic science and clinical information. Students review a narrative clinical case and construct an MCD using items provided by the case author. Students identify the relationships among underlying risk factors, etiology, pathogenesis and pathophysiology, and the patients' signs and symptoms. They receive one point for each correctly-identified link. In 2014-15 and 2015-16, case diagrams were implemented in consecutive classes of 150 medical students. The alpha reliability coefficient for the overall score, constructed using each student's mean proportion correct across all cases, was 0.82. Discrimination indices for each of the case scores with the overall score ranged from 0.23 to 0.51. In a G study using those students with complete data (n = 251) on all 16 cases, 10% of the variance was true score variance, and systematic case variance was large. Using 16 cases generated a G coefficient (relative score reliability) equal to .72 and a Phi equal to .65. The next phase of the project will involve deploying MCDs in higher-stakes settings to determine whether similar results can be achieved. Further analyses will determine whether these assessments correlate with other measures of higher-order thinking skills.

  8. 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

  9. 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.

  10. Stereo 3D spatial phase diagrams

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jinwu, E-mail: kangjw@tsinghua.edu.cn; Liu, Baicheng, E-mail: liubc@tsinghua.edu.cn

    2016-07-15

    Phase diagrams serve as the fundamental guidance in materials science and engineering. Binary P-T-X (pressure–temperature–composition) and multi-component phase diagrams are of complex spatial geometry, which brings difficulty for understanding. The authors constructed 3D stereo binary P-T-X, typical ternary and some quaternary phase diagrams. A phase diagram construction algorithm based on the calculated phase reaction data in PandaT was developed. And the 3D stereo phase diagram of Al-Cu-Mg ternary system is presented. These phase diagrams can be illustrated by wireframe, surface, solid or their mixture, isotherms and isopleths can be generated. All of these can be displayed by the three typical display ways: electronic shutter, polarization and anaglyph (for example red-cyan glasses). Especially, they can be printed out with 3D stereo effect on paper, and watched by the aid of anaglyph glasses, which makes 3D stereo book of phase diagrams come to reality. Compared with the traditional illustration way, the front of phase diagrams protrude from the screen and the back stretches far behind of the screen under 3D stereo display, the spatial structure can be clearly and immediately perceived. These 3D stereo phase diagrams are useful in teaching and research. - Highlights: • Stereo 3D phase diagram database was constructed, including binary P-T-X, ternary, some quaternary and real ternary systems. • The phase diagrams can be watched by active shutter or polarized or anaglyph glasses. • The print phase diagrams retains 3D stereo effect which can be achieved by the aid of anaglyph glasses.

  11. 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

  12. 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

  13. Towards Diagram Understanding: A Pilot Study Measuring Cognitive Workload Through Eye-Tracking

    DEFF Research Database (Denmark)

    Maier, Anja; Baltsen, Nick; Christoffersen, Henrik

    2014-01-01

    We investigate model understanding, in particular , how the quality of the UML diagram layout impacts cognitive load. We hypothesize that this w ill have a significant impact on the structure and effectiveness of engineers’ communication. In previous work, we have studied task performance...... measurements and subjective assessments; here, we also investigate behavioral indicators such as fixation and pupillary dilation. We use such indicators to explore diagram understanding- and reading strategies and how such strategies are impacted, e.g. by diagram type and expertise level. In the pilot eye...

  14. 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

  15. 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

  16. Importance diagrams - a novel presentation of the response of a material to neutron irradiation

    International Nuclear Information System (INIS)

    Forrest, R.A.

    1998-01-01

    Activation of fusion materials following neutron irradiation is of great technological importance, especially in the study of safety and environmental impacts. Currently, activation calculations are performed for a particular neutron spectrum, appropriate to a region in a particular fusion device, which makes it difficult to extract generic information. The present work gives details of a method to present the dominant nuclides for the radiological responses or an irradiated material in a fashion that is independent of the neutron spectrum and almost independent of the flux. The importance diagrams show regions in the decay time versus neutron energy space where a nuclide contributes >50% of the response. The importance diagrams for pure iron and SS316 are described, and it is noted that the shapes of the various regions vary very little with the total neutron flux. Variation of the diagrams with irradiation time occurs at short decay times in a systematic fashion. The use of the diagrams in a realistic spectrum relies on an expansion, which while not generally true, does hold approximately for many of the nuclides of interest. The diagrams are therefore a valuable summary of the universal, device-independent, response of the materials, and when combined with pathway information give a comprehensive description of activation for that material. (orig.)

  17. 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.

  18. 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.

  19. A study on the influence diagrams for the application to containment performance analysis

    International Nuclear Information System (INIS)

    Park, Joon Won

    1995-02-01

    Influence diagrams have been applied to containment performance analysis of Young-Gwang 3 and 4 in an effort to explicitly display the dependencies between events and to treat operator intervention more generally. This study has been initiated to remove the three major drawbacks of the current event tree methodology: 1) Event tree cannot express dependency between events explicitly. 2) Accident Progression Event Tree (APET) cannot represent entire containment system. 3) It is difficult to consider operator intervention with event tree. To resolve these problems, a new approach, i.e., influence diagrams, are proposed. In the present work, the applicability of the influence diagrams have been demonstrated to YGN 3 and 4 containment performance analysis and an assessment of accident management strategies. To show that the results of the application of influence diagrams are reasonable, results are compared with that of YGN 3 and 4 IPE. Both results are in good agreement. In addition, influence diagrams are used to assess two accident management strategies: 1) RCS depressurization, 2) cavity flooding. Cavity flooding has a favorable effect to late containment failure and basemat melt-through, and depressurization of RCS is good for steam generator tube rupture. However, early containment failure probability is worse in both cases. As a result of the present study, it is shown that influence diagrams can be applied to the containment performance analysis

  20. 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

  1. 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)

  2. 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)

  3. 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

  4. 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

  5. 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

  6. Automation of Feynman diagram evaluations

    International Nuclear Information System (INIS)

    Tentyukov, M.N.

    1998-01-01

    A C-program DIANA (DIagram ANAlyser) for the automation of Feynman diagram evaluations is presented. It consists of two parts: the analyzer of diagrams and the interpreter of a special text manipulating language. This language can be used to create a source code for analytical or numerical evaluations and to keep the control of the process in general

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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...

  12. 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

  13. Effect of a Diagram on Primary Students' Understanding About Electric Circuits

    Science.gov (United States)

    Preston, Christine Margaret

    2017-09-01

    This article reports on the effect of using a diagram to develop primary students' conceptual understanding about electric circuits. Diagrammatic representations of electric circuits are used for teaching and assessment despite the absence of research on their pedagogical effectiveness with young learners. Individual interviews were used to closely analyse Years 3 and 5 (8-11-year-old) students' explanations about electric circuits. Data was collected from 20 students in the same school providing pre-, post- and delayed post-test dialogue. Students' thinking about electric circuits and changes in their explanations provide insights into the role of diagrams in understanding science concepts. Findings indicate that diagram interaction positively enhanced understanding, challenged non-scientific views and promoted scientific models of electric circuits. Differences in students' understanding about electric circuits were influenced by prior knowledge, meta-conceptual awareness and diagram conventions including a stylistic feature of the diagram used. A significant finding that students' conceptual models of electric circuits were energy rather than current based has implications for electricity instruction at the primary level.

  14. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A2: Scenarios

    International Nuclear Information System (INIS)

    Skagius, K.; Wiborgh, M.

    1998-01-01

    This appendix gives a short description of the scenario methodology adopted in the previous safety assessment of SFR. Since then new methodologies for developing structured descriptions of how processes and interactions between processes affect the evolution of a repository system. Two such methods are briefly described. These methods are very similar, but they differ in the way the system is graphically structured. One of the methods is based on Process Influence Diagrams, PID, and the other on Interaction matrices. It is proposed that the method based on Interaction matrices is used for the scenario work in project SAFE. The main reason for this is that the method already has been applied by SKB, which means that it will be possible to use already existing procedures and documentation systems. The proposed procedure involves the development of Interaction matrices for a defined Reference scenario and the use of these matrices to illustrate the effect of different Scenario initiating FEPs. The proposed procedure is described in this appendix

  15. 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

  16. 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.

  17. 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.

  18. 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

  19. 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)

  20. VennDiagramWeb: a web application for the generation of highly customizable Venn and Euler diagrams.

    Science.gov (United States)

    Lam, Felix; Lalansingh, Christopher M; Babaran, Holly E; Wang, Zhiyuan; Prokopec, Stephenie D; Fox, Natalie S; Boutros, Paul C

    2016-10-03

    Visualization of data generated by high-throughput, high-dimensionality experiments is rapidly becoming a rate-limiting step in computational biology. There is an ongoing need to quickly develop high-quality visualizations that can be easily customized or incorporated into automated pipelines. This often requires an interface for manual plot modification, rapid cycles of tweaking visualization parameters, and the generation of graphics code. To facilitate this process for the generation of highly-customizable, high-resolution Venn and Euler diagrams, we introduce VennDiagramWeb: a web application for the widely used VennDiagram R package. VennDiagramWeb is hosted at http://venndiagram.res.oicr.on.ca/ . VennDiagramWeb allows real-time modification of Venn and Euler diagrams, with parameter setting through a web interface and immediate visualization of results. It allows customization of essentially all aspects of figures, but also supports integration into computational pipelines via download of R code. Users can upload data and download figures in a range of formats, and there is exhaustive support documentation. VennDiagramWeb allows the easy creation of Venn and Euler diagrams for computational biologists, and indeed many other fields. Its ability to support real-time graphics changes that are linked to downloadable code that can be integrated into automated pipelines will greatly facilitate the improved visualization of complex datasets. For application support please contact Paul.Boutros@oicr.on.ca.

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. Genus Ranges of Chord Diagrams.

    Science.gov (United States)

    Burns, Jonathan; Jonoska, Nataša; Saito, Masahico

    2015-04-01

    A chord diagram consists of a circle, called the backbone, with line segments, called chords, whose endpoints are attached to distinct points on the circle. The genus of a chord diagram is the genus of the orientable surface obtained by thickening the backbone to an annulus and attaching bands to the inner boundary circle at the ends of each chord. Variations of this construction are considered here, where bands are possibly attached to the outer boundary circle of the annulus. The genus range of a chord diagram is the genus values over all such variations of surfaces thus obtained from a given chord diagram. Genus ranges of chord diagrams for a fixed number of chords are studied. Integer intervals that can be, and those that cannot be, realized as genus ranges are investigated. Computer calculations are presented, and play a key role in discovering and proving the properties of genus ranges.

  6. 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

  7. 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/

  8. 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/

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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...

  15. 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...

  16. 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.

  17. Status of Safety and Environmental Activities for Inertial Fusion Energy

    International Nuclear Information System (INIS)

    Latkowski, J.F.; Reyes, S.; Cadwallader, L.C.; Sharpe, J.P.; Marshall, T.D.; Merrill, B.J.; Moore, R.L.; Petti, D.A.; Falquina, R.; Rodriguez, A.; Sanz, J.; Cabellos, O.

    2003-01-01

    Over the past several years, significant progress has been made in the analysis of safety and environmental (S and E) issues for inertial fusion energy (IFE). Detailed safety assessments have been performed for the baseline power plant concepts, as well as for a conceptual target fabrication facility. Safety analysis results are helping to drive the agenda for experiments. A survey of the S and E characteristics - both radiological and chemical - of candidate target materials has been completed. Accident initiating events have been identified and incorporated into master logic diagrams, which will be essential to the detailed safety analyses that will be needed in the future. Studies of aerosol generation and transport will have important safety implications. A Monte Carlo-based uncertainty analysis procedure has been developed for use in neutron activation calculations. Finally, waste management issues are receiving increased attention and are deserving of further discussion

  18. 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!

  19. Collaborative diagramming during problem based learning in medical education: Do computerized diagrams support basic science knowledge construction?

    Science.gov (United States)

    De Leng, Bas; Gijlers, Hannie

    2015-05-01

    To examine how collaborative diagramming affects discussion and knowledge construction when learning complex basic science topics in medical education, including its effectiveness in the reformulation phase of problem-based learning. Opinions and perceptions of students (n = 70) and tutors (n = 4) who used collaborative diagramming in tutorial groups were collected with a questionnaire and focus group discussions. A framework derived from the analysis of discourse in computer-supported collaborative leaning was used to construct the questionnaire. Video observations were used during the focus group discussions. Both students and tutors felt that collaborative diagramming positively affected discussion and knowledge construction. Students particularly appreciated that diagrams helped them to structure knowledge, to develop an overview of topics, and stimulated them to find relationships between topics. Tutors emphasized that diagramming increased interaction and enhanced the focus and detail of the discussion. Favourable conditions were the following: working with a shared whiteboard, using a diagram format that facilitated distribution, and applying half filled-in diagrams for non-content expert tutors and\\or for heterogeneous groups with low achieving students. The empirical findings in this study support the findings of earlier more descriptive studies that diagramming in a collaborative setting is valuable for learning complex knowledge in medicine.

  20. 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

  1. 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

  2. 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

  3. 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.)

  4. Integrated system of safety features for spent fuel interim storage

    International Nuclear Information System (INIS)

    Pantazi, Doina; Stanciu, Marcela; Mateescu, Silvia; Marin, Ion

    1999-01-01

    The design of the spent fuel interim storage facility (SFISF) must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility. To elaborate the safety documentation necessary for licensing, we were trying to chose the most appropriate approach related to safety features for SFISF, based on national and international regulations, standards and recommendations, as well as on the experience of other countries with similar facilities and finally, on our own experience in designing other nuclear objectives in Romania. The paper presents the issues that we consider important for the safety evaluation and are developed as a detailed diagram. The diagram contains in a logical succession the following issues: - fundamental principles of radioprotection; - fundamental safety principles of radioactive waste management; - safety objectives of SFISF; - safety criteria for SFISF; - safety requirements for SFISF; - siting criteria for SFISF; - siting requirements for SFISF. (authors)

  5. 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

  6. Asymptotic laws for random knot diagrams

    Science.gov (United States)

    Chapman, Harrison

    2017-06-01

    We study random knotting by considering knot and link diagrams as decorated, (rooted) topological maps on spheres and pulling them uniformly from among sets of a given number of vertices n, as first established in recent work with Cantarella and Mastin. The knot diagram model is an exciting new model which captures both the random geometry of space curve models of knotting as well as the ease of computing invariants from diagrams. We prove that unknot diagrams are asymptotically exponentially rare, an analogue of Sumners and Whittington’s landmark result for self-avoiding polygons. Our proof uses the same key idea: we first show that knot diagrams obey a pattern theorem, which describes their fractal structure. We examine how quickly this behavior occurs in practice. As a consequence, almost all diagrams are asymmetric, simplifying sampling from this model. We conclude with experimental data on knotting in this model. This model of random knotting is similar to those studied by Diao et al, and Dunfield et al.

  7. A Critical Appraisal of the `Day' Diagram

    Science.gov (United States)

    Roberts, A. P.; Tauxe, L.; Heslop, D.

    2017-12-01

    The `Day' diagram [Day et al., 1977; doi:10.1016/0031-9201(77)90108-X] is used widely to infer the mean domain state of magnetic mineral assemblages. The Day plot coordinates are the ratios of the saturation remanent magnetization to saturation magnetization (Mrs/Ms) and the coercivity of remanence to coercivity (Bcr/Bc), as determined from a major hysteresis loop and a backfield demagnetization curve. Based on theoretical and empirical arguments, Day plots are typically demarcated into stable single domain (SD), `pseudosingle domain' (`PSD'), and multidomain (MD) zones. It is a simple task to determine Mrs/Ms and Bcr/Bc for a sample and to assign a mean domain state based on the boundaries defined by Day et al. [1977]. Many other parameters contribute to variability in a Day diagram, including surface oxidation, mineral stoichiometry, stress state, magnetostatic interactions, and mixtures of magnetic particles with different sizes and shapes. Bulk magnetic measurements usually lack detailed independent evidence to constrain each free parameter, which makes the Day diagram fundamentally ambiguous. This raises questions about its usefulness for diagnosing magnetic particle size variations. The Day diagram is also used to make inferences about binary mixing of magnetic particles, where, for example, mixtures of SD and MD particles give rise to a bulk `PSD' response even though the concentration of `PSD' grains could be zero. In our assessment of thousands of hysteresis measurements of geological samples, binary mixing occurs in a tiny number of cases. Ternary, quaternary, and higher order mixing are usually observed. Also, uniaxial SD and MD end-members are nearly always inappropriate for considering mixing because uniaxial SD particles are virtually non-existent in igneous rocks. Thus, use of mixing lines in Day diagrams routinely provides unsatisfactory representations of particle size variations. We critically appraise the Day diagram and argue that its many

  8. 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)

  9. 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)

  10. Safety Concepts in Structural Glass Engineering : Towards an Integrated Approach

    NARCIS (Netherlands)

    Bos, F.P.

    2009-01-01

    This dissertation proposes the Integrated Approach to Structural Glass Safety, based on four clearly defined element safety properties, damage sensitivity, relative resistance, redundancy, and fracture mode. The Element Safety Diagram (ESD) is introduced to provide an easy-to-read graphical

  11. Drawing Euler Diagrams with Circles

    OpenAIRE

    Stapleton, Gem; Zhang, Leishi; Howse, John; Rodgers, Peter

    2010-01-01

    Euler diagrams are a popular and intuitive visualization tool which are used in a wide variety of application areas, including biological and medical data analysis. As with other data visualization methods, such as graphs, bar charts, or pie charts, the automated generation of an Euler diagram from a suitable data set would be advantageous, removing the burden of manual data analysis and the subsequent task of drawing an appropriate diagram. Various methods have emerged that automatically dra...

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. 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

  20. Application of the cause-consequence diagram method to static systems

    International Nuclear Information System (INIS)

    Andrews, J.D.; Ridley, L.M.

    2002-01-01

    In the last 30 years, various mathematical models have been used to identify the effect of component failures on the performance of a system. The most frequently used technique for system reliability assessment is Fault Tree Analysis (FTA) and a large proportion of its popularity can be attributed to the fact that it provides a very good documentation of the way that the system failure logic was developed. Exact quantification of the fault tree, however, can be problematic for very large systems and in such situations, approximations can be used. Alternatively, an exact result can be obtained via the conversion of the fault tree into a binary decision diagram (BDD). The BDD, however, loses all failure logic documentation during the conversion process. This paper outlines the use of the cause-consequence diagram method as a tool for system risk and reliability analysis. As with the FTA method, the cause-consequence diagram documents the failure logic of the system. In addition to this the cause-consequence diagram produces the exact failure probability in a very efficient calculation procedure. The cause-consequence diagram technique has been applied to a static system and shown to yield the same result as those produced by the solution of the equivalent fault tree and BDD. On the basis of this general rules have been devised for the correct construction of the cause-consequence diagram given a static system. The use of the cause-consequence method in this manner has significant implications in terms of efficiency of the reliability analysis and can be shown to have benefits for static systems

  1. 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).

  2. 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

  3. Network and system diagrams revisited: Satisfying CEA requirements for causality analysis

    International Nuclear Information System (INIS)

    Perdicoulis, Anastassios; Piper, Jake

    2008-01-01

    Published guidelines for Cumulative Effects Assessment (CEA) have called for the identification of cause-and-effect relationships, or causality, challenging researchers to identify methods that can possibly meet CEA's specific requirements. Together with an outline of these requirements from CEA key literature, the various definitions of cumulative effects point to the direction of a method for causality analysis that is visually-oriented and qualitative. This article consequently revisits network and system diagrams, resolves their reported shortcomings, and extends their capabilities with causal loop diagramming methodology. The application of the resulting composite causality analysis method to three Environmental Impact Assessment (EIA) case studies appears to satisfy the specific requirements of CEA regarding causality. Three 'moments' are envisaged for the use of the proposed method: during the scoping stage, during the assessment process, and during the stakeholder participation process

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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)

  16. 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

  17. 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)

  18. Application of ISO22000, failure mode, and effect analysis (FMEA) cause and effect diagrams and pareto in conjunction with HACCP and risk assessment for processing of pastry products.

    Science.gov (United States)

    Varzakas, Theodoros H

    2011-09-01

    The Failure Mode and Effect Analysis (FMEA) model has been applied for the risk assessment of pastry processing. A tentative approach of FMEA application to the pastry industry was attempted in conjunction with ISO22000. Preliminary Hazard Analysis was used to analyze and predict the occurring failure modes in a food chain system (pastry processing plant), based on the functions, characteristics, and/or interactions of the ingredients or the processes, upon which the system depends. Critical Control points have been identified and implemented in the cause and effect diagram (also known as Ishikawa, tree diagram, and fishbone diagram). In this work a comparison of ISO22000 analysis with HACCP is carried out over pastry processing and packaging. However, the main emphasis was put on the quantification of risk assessment by determining the Risk Priority Number (RPN) per identified processing hazard. Storage of raw materials and storage of final products at -18°C followed by freezing were the processes identified as the ones with the highest RPN (225, 225, and 144 respectively) and corrective actions were undertaken. Following the application of corrective actions, a second calculation of RPN values was carried out leading to considerably lower values (below the upper acceptable limit of 130). It is noteworthy that the application of Ishikawa (Cause and Effect or Tree diagram) led to converging results thus corroborating the validity of conclusions derived from risk assessment and FMEA. Therefore, the incorporation of FMEA analysis within the ISO22000 system of a pastry processing industry is considered imperative.

  19. Model-Driven Development of Safety Architectures

    Science.gov (United States)

    Denney, Ewen; Pai, Ganesh; Whiteside, Iain

    2017-01-01

    We describe the use of model-driven development for safety assurance of a pioneering NASA flight operation involving a fleet of small unmanned aircraft systems (sUAS) flying beyond visual line of sight. The central idea is to develop a safety architecture that provides the basis for risk assessment and visualization within a safety case, the formal justification of acceptable safety required by the aviation regulatory authority. A safety architecture is composed from a collection of bow tie diagrams (BTDs), a practical approach to manage safety risk by linking the identified hazards to the appropriate mitigation measures. The safety justification for a given unmanned aircraft system (UAS) operation can have many related BTDs. In practice, however, each BTD is independently developed, which poses challenges with respect to incremental development, maintaining consistency across different safety artifacts when changes occur, and in extracting and presenting stakeholder specific information relevant for decision making. We show how a safety architecture reconciles the various BTDs of a system, and, collectively, provide an overarching picture of system safety, by considering them as views of a unified model. We also show how it enables model-driven development of BTDs, replete with validations, transformations, and a range of views. Our approach, which we have implemented in our toolset, AdvoCATE, is illustrated with a running example drawn from a real UAS safety case. The models and some of the innovations described here were instrumental in successfully obtaining regulatory flight approval.

  20. 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)

  1. TIME-TEMPERATURE-TRANSFORMATION (TTT) DIAGRAMS FOR FUTURE WASTE COMPOSITIONS

    International Nuclear Information System (INIS)

    Billings, A.; Edwards, T.

    2010-01-01

    As a part of the Waste Acceptance Product Specifications (WAPS) for Vitrified High-Level Waste Forms defined by the Department of Energy - Office of Environmental Management, the waste form stability must be determined for each of the projected high-level waste (HLW) types at the Savannah River Site (SRS). Specifically, WAPS 1.4.1 requires the glass transition temperature (T g ) to be defined and time-temperature-transformation (TTT) diagrams to be developed. The T g of a glass is an indicator of the approximate temperature where the supercooled liquid converts to a solid on cooling or conversely, where the solid begins to behave as a viscoelastic solid on heating. A TTT diagram identifies the crystalline phases that can form as a function of time and temperature for a given waste type or more specifically, the borosilicate glass waste form. In order to assess durability, the Product Consistency Test (PCT) was used and the durability results compared to the Environmental Assessment (EA) glass. The measurement of glass transition temperature and the development of TTT diagrams have already been performed for the seven Defense Waste Processing Facility (DWPF) projected compositions as defined in the Waste Form Compliance Plan (WCP) and in SRNL-STI-2009-00025. Additional phase transformation information exists for other projected compositions, but overall these compositions did not cover composition regions estimated for future waste processing. To develop TTT diagrams for future waste types, the Savannah River National Laboratory (SRNL) fabricated two caches of glass from reagent grade oxides to simulate glass compositions which would be likely processed with and without Al dissolution. These were used for glass transition temperature measurement and TTT diagram development. The glass transition temperatures of both glasses were measured using differential scanning calorimetry (DSC) and were recorded to be 448 C and 452 C. Using the previous TTT diagrams as guidance

  2. 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...

  3. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

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

  4. Use of the Ishikawa diagram in a case-control analysis to assess the causes of a diffuse lamellar keratitis outbreak

    Directory of Open Access Journals (Sweden)

    Luis Henrique Lira

    Full Text Available ABSTRACT Purpose: To identify the causes of a diffuse lamellar keratitis (DLK outbreak using a systematic search tool in a case-control analysis. Methods: An Ishikawa diagram was used to guide physicians to determine the potential risk factors involved in this outbreak. Coherence between the occurrences and each possible cause listed in the diagram was verified, and the total number of eyes at risk was used to calculate the proportion of affected eyes. Multivariate analysis was performed using logistic regression to determine the independent effect of the risk factors, after controlling for confounders and test interactions. Results: All DLK cases were reported in 2007 between June 13 and December 21; during this period, 3,698 procedures were performed. Of the 1,682 flap-related procedures, 204 eyes of 141 individuals presented with DLK. No direct relationship was observed between the occurrence of DLK and the presence of any specific factors; however, flap-lifting enhancements, procedures performed during the morning shift, and non-use of therapeutic contact lenses after the surgery were significantly related to higher occurrence percentages of this condition. Conclusions: The Ishikawa diagram, like most quality tools, is a visualization and knowledge organization tool. This systematization allowed the investigators to thoroughly assess all the possible causes of DLK outbreak. A clear view of the entire surgical logistics permitted even more rigid management of the main factors involved in the process and, as a result, highlighted factors that deserved attention. The case-control analysis on every factor raised by the Ishikawa diagram indicated that the commonly suspected factors such as biofilm contamination of the water reservoir in autoclaves, the air-conditioning filter system, glove powder, microkeratome motor oil, and gentian violet markers were not related to the outbreak.

  5. Use of the Ishikawa diagram in a case-control analysis to assess the causes of a diffuse lamellar keratitis outbreak.

    Science.gov (United States)

    Lira, Luis Henrique; Hirai, Flávio E; Oliveira, Marivaldo; Portellinha, Waldir; Nakano, Eliane Mayumi

    2017-01-01

    To identify the causes of a diffuse lamellar keratitis (DLK) outbreak using a systematic search tool in a case-control analysis. An Ishikawa diagram was used to guide physicians to determine the potential risk factors involved in this outbreak. Coherence between the occurrences and each possible cause listed in the diagram was verified, and the total number of eyes at risk was used to calculate the proportion of affected eyes. Multivariate analysis was performed using logistic regression to determine the independent effect of the risk factors, after controlling for confounders and test interactions. All DLK cases were reported in 2007 between June 13 and December 21; during this period, 3,698 procedures were performed. Of the 1,682 flap-related procedures, 204 eyes of 141 individuals presented with DLK. No direct relationship was observed between the occurrence of DLK and the presence of any specific factors; however, flap-lifting enhancements, procedures performed during the morning shift, and non-use of therapeutic contact lenses after the surgery were significantly related to higher occurrence percentages of this condition. The Ishikawa diagram, like most quality tools, is a visualization and knowledge organization tool. This systematization allowed the investigators to thoroughly assess all the possible causes of DLK outbreak. A clear view of the entire surgical logistics permitted even more rigid management of the main factors involved in the process and, as a result, highlighted factors that deserved attention. The case-control analysis on every factor raised by the Ishikawa diagram indicated that the commonly suspected factors such as biofilm contamination of the water reservoir in autoclaves, the air-conditioning filter system, glove powder, microkeratome motor oil, and gentian violet markers were not related to the outbreak.

  6. 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

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. SafetyBarrierManager, a software tool to perform risk analysis using ARAMIS's principles

    DEFF Research Database (Denmark)

    Duijm, Nijs Jan

    2017-01-01

    of the ARAMIS project, Risø National Laboratory started developing a tool that could implement these methodologies, leading to SafetyBarrierManager. The tool is based on the principles of “safety‐barrier diagrams”, which are very similar to “bowties”, with the possibility of performing quantitative analysis......The ARAMIS project resulted in a number of methodologies, dealing with among others: the development of standard fault trees and “bowties”; the identification and classification of safety barriers; and including the quality of safety management into the quantified risk assessment. After conclusion....... The tool allows constructing comprehensive fault trees, event trees and safety‐barrier diagrams. The tool implements the ARAMIS idea of a set of safety barrier types, to which a number of safety management issues can be linked. By rating the quality of these management issues, the operational probability...

  12. 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

  13. 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

  14. 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

  15. [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

  16. 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

  17. 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

  18. Risk analysis using AS/NZS 4360:2004, Bow-Tie diagram and ALARP on construction projects of Banyumanik Hospital

    Science.gov (United States)

    Sari, Diana Puspita; Pujotomo, Darminto; Wardani, Nadira Kusuma

    2017-11-01

    The Determination of risk is an uncertain event. Risks can have negative or positive impacts on project objectives. A project was defined as a series of activities and tasks that have a purpose, specifications, and limits of cost. Banyumanik Hospital Development Project is one of the construction projects in Semarang which have experienced some problems. The first problem is project delays on building stake. The second problem is delay of material supply. Finally, the problem that occurs is less management attention to health safety as evidenced by the unavailability of PPE for the workers. These problems will pose a risk to be a very important risk management performed by contractors at the Banyumanik Hospital Development Project to reduce the impact that would be caused by the risk borne by the provider of construction services. This research aim to risk identification, risk assessment and risk mitigation. Project risk management begins with the identification of risks based on the project life cycle. The risk assessment carried out by AS I NZS 4360: 2004 to the impacts of cost, time and quality. The results obtained from the method of AS I NZS 4360: 2004 is the risk that requires the handling of mitigation. Mitigated risk is the risk that had significant and high level. There are four risks that require risk mitigation with Bow-Tie diagrams which is work accidents, contract delays, material delays and design changes. Bow-Tie diagram method is a method for identifying causal and preventive action and recovery of a risk. Results obtained from Bow-Tie diagram method is a preventive action and recovery. This action is used as input to the ALARP method. ALARP method is used to determine the priority of the strategy proposed in the category broadly acceptable, tolerable, and unacceptable.

  19. 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.

  20. 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.

  1. 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.

  2. The reliability of measuring pain distribution and location using body pain diagrams in patients with acute whiplash-associated disorders.

    Science.gov (United States)

    Southerst, Danielle; Stupar, Maja; Côté, Pierre; Mior, Silvano; Stern, Paula

    2013-09-01

    The objective of this study was to measure the interexaminer reliability of scoring pain distribution using paper and electronic body pain diagrams in patients with acute whiplash-associated disorder and to assess the intermethod reliability of measuring pain distribution and location using paper and electronic diagrams. We conducted an interexaminer reliability study on 80 participants recruited from a randomized controlled trial on the conservative management of acute grade I/II whiplash-associated disorder. Participants were assessed for inclusion/exclusion criteria by an experienced clinician. As part of the baseline assessment, participants independently completed paper and electronic pain diagrams. Diagrams were scored independently by 2 examiners using the body region method. Interexaminer and intermethod reliability was computed using intraclass correlation coefficients (ICCs) for pain distribution and κ coefficient for pain location. We used Bland-Altman plots to compute limits of agreement. The interexaminer reliability was ICC = 0.925 for paper and ICC = 0.997 for the electronic body pain diagram. The intermethod reliability for measuring pain distribution ranged from ICC = 0.63 to ICC = 0.93. For pain location, the intermethod reliability varied from κ = 0.23 (posterior neck) to κ = 0.90 (right side of the face). We found good to excellent interexaminer reliability for scoring 2 versions of the body pain diagram. Pain distribution and pain location were reliably and consistently measured on body pain diagrams using paper and electronic methods; therefore, clinicians and researchers may choose either medium when using body pain diagrams. Copyright © 2013 National University of Health Sciences. Published by Mosby, Inc. All rights reserved.

  3. On-shell diagrams for N=8 supergravity amplitudes

    Energy Technology Data Exchange (ETDEWEB)

    Heslop, Paul; Lipstein, Arthur E. [Department of Mathematical Sciences, Durham University,Lower Mountjoy, Stockton Road, Durham, DH1 3LE (United Kingdom)

    2016-06-10

    We define recursion relations for N=8 supergravity amplitudes using a generalization of the on-shell diagrams developed for planar N=4 super-Yang-Mills. Although the recursion relations generically give rise to non-planar on-shell diagrams, we show that at tree-level the recursion can be chosen to yield only planar diagrams, the same diagrams occurring in the planar N=4 theory. This implies non-trivial identities for non-planar diagrams as well as interesting relations between the N=4 and N=8 theories. We show that the on-shell diagrams of N=8 supergravity obey equivalence relations analogous to those of N=4 super-Yang-Mills, and we develop a systematic algorithm for reading off Grassmannian integral formulae directly from the on-shell diagrams. We also show that the 1-loop 4-point amplitude of N=8 supergravity can be obtained from on-shell diagrams.

  4. Phase diagram of ammonium nitrate

    Energy Technology Data Exchange (ETDEWEB)

    Dunuwille, Mihindra; Yoo, Choong-Shik, E-mail: csyoo@wsu.edu [Department of Chemistry and Institute for Shock Physics, Washington State University, Pullman, Washington 99164 (United States)

    2013-12-07

    Ammonium Nitrate (AN) is a fertilizer, yet becomes an explosive upon a small addition of chemical impurities. The origin of enhanced chemical sensitivity in impure AN (or AN mixtures) is not well understood, posing significant safety issues in using AN even today. To remedy the situation, we have carried out an extensive study to investigate the phase stability of AN and its mixtures with hexane (ANFO–AN mixed with fuel oil) and Aluminum (Ammonal) at high pressures and temperatures, using diamond anvil cells (DAC) and micro-Raman spectroscopy. The results indicate that pure AN decomposes to N{sub 2}, N{sub 2}O, and H{sub 2}O at the onset of the melt, whereas the mixtures, ANFO and Ammonal, decompose at substantially lower temperatures. The present results also confirm the recently proposed phase IV-IV{sup ′} transition above 17 GPa and provide new constraints for the melting and phase diagram of AN to 40 GPa and 400°C.

  5. Phase diagram of ammonium nitrate

    International Nuclear Information System (INIS)

    Dunuwille, Mihindra; Yoo, Choong-Shik

    2013-01-01

    Ammonium Nitrate (AN) is a fertilizer, yet becomes an explosive upon a small addition of chemical impurities. The origin of enhanced chemical sensitivity in impure AN (or AN mixtures) is not well understood, posing significant safety issues in using AN even today. To remedy the situation, we have carried out an extensive study to investigate the phase stability of AN and its mixtures with hexane (ANFO–AN mixed with fuel oil) and Aluminum (Ammonal) at high pressures and temperatures, using diamond anvil cells (DAC) and micro-Raman spectroscopy. The results indicate that pure AN decomposes to N 2 , N 2 O, and H 2 O at the onset of the melt, whereas the mixtures, ANFO and Ammonal, decompose at substantially lower temperatures. The present results also confirm the recently proposed phase IV-IV ′ transition above 17 GPa and provide new constraints for the melting and phase diagram of AN to 40 GPa and 400°C

  6. 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

  7. 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)

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

    OpenAIRE

    Adjekum, Daniel Kwasi

    2014-01-01

    An assessment of the perceptions of respondents on the safety culture at an accredited Part 141 four year collegiate aviation program was conducted as part of the implementation of a safety management system (SMS). The Collegiate Aviation Program Safety Culture Assessment Survey (CAPSCAS), which was modified and revalidated from the existing Commercial Aviation Safety Survey (CASS), was used. Participants were drawn from flight students and certified flight instructors in the program. The sur...

  9. Role and meaning of safety assessment from the point of view of IAEA

    International Nuclear Information System (INIS)

    Lyubarskiy, A.

    2012-01-01

    In 2006, the IAEA published its revised Safety Fundamentals. This states that the ''fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation''. This objective has to be achieved for all facilities and activities and for all stages over the lifetime of a facility by adherence to ten fundamental principles. This leads, inter alia, to the requirement for a safety assessment to be carried out. In particular, the text accompanying Principle 3 on leadership and management for safety states that: ''3.15. Safety has to be assessed for all facilities and activities, consistent with a graded approach. Safety assessment involves the systematic analysis of normal operation and its effects, of the ways in which failures might occur and of the consequences of such failures. Safety assessments cover the safety measures necessary to control the hazard, and the design and engineered safety features are assessed to demonstrate that they fulfill the safety functions required of them. Where control measures or operator actions are called on to maintain safety, an initial safety assessment has to be carried out to demonstrate that the arrangements made are robust and that they can be relied on. A facility may only be constructed and commissioned or an activity may only be commenced once it has been demonstrated to the satisfaction of the regulatory body that the proposed safety measures are adequate.'' Principle 3 further states that the process of safety assessment for facilities and activities is repeated in the conduct of operations in order to take into account changed circumstances (such as the application of new standards or scientific and technological developments), the feedback of operating experience, modifications and the effects of ageing. Continuation of operations over long periods of time requires reassessments demonstrating that the safety measures remain adequate. (orig.)

  10. Thermochemical measurements and assessment of the phase diagrams in the system Y-Ba-Cu-O

    International Nuclear Information System (INIS)

    Boudene, A.; Mohammad, A.

    1996-01-01

    The aim of this project was to provide a self-consistent set of Gibbs energy data for all phases in the system Y-Ba-Cu-O. Experimental thermochemical investigations by differential thermal analysis (DTA), thermogravimetry (TG), electromotive force measurements (EMF), oxygen coulometric titration (OCT), drop and solution calorimetry, and conventional phase analysis (annealing, quenching, and X-ray diffraction [XRD]) as well as ab initio calculations of interaction energies for the 123 phase have been carried out. The experimental information (phase equilibria, heat capacity, enthalpies of formation, oxygen partial pressures, and so forth) has been used in computer-based assessments of the Gibbs energies. These data have been employed to generate phase diagrams by way of equilibrium computations. All binary and ternary subsystems have been fully assessed. For the quaternary system a dataset covering the subsolidus range has been derived. Applications of the data to practical questions, such as the production of 123 superconductors by an MOCVD process, the producibility of metallic precursors, and the oxidation of a copper-enriched stoichiometric oxide precursor, are demonstrated

  11. The mean squared writhe of alternating random knot diagrams

    Energy Technology Data Exchange (ETDEWEB)

    Diao, Y; Hinson, K [Department of Mathematics and Statistics University of North Carolina at Charlotte, NC 28223 (United States); Ernst, C; Ziegler, U, E-mail: ydiao@uncc.ed [Department of Mathematics and Computer Science, Western Kentucky University, Bowling Green, KY 42101 (United States)

    2010-12-10

    The writhe of a knot diagram is a simple geometric measure of the complexity of the knot diagram. It plays an important role not only in knot theory itself, but also in various applications of knot theory to fields such as molecular biology and polymer physics. The mean squared writhe of any sample of knot diagrams with n crossings is n when for each diagram at each crossing one of the two strands is chosen as the overpass at random with probability one-half. However, such a diagram is usually not minimal. If we restrict ourselves to a minimal knot diagram, then the choice of which strand is the over- or under-strand at each crossing is no longer independent of the neighboring crossings and a larger mean squared writhe is expected for minimal diagrams. This paper explores the effect on the correlation between the mean squared writhe and the diagrams imposed by the condition that diagrams are minimal by studying the writhe of classes of reduced, alternating knot diagrams. We demonstrate that the behavior of the mean squared writhe heavily depends on the underlying space of diagram templates. In particular this is true when the sample space contains only diagrams of a special structure. When the sample space is large enough to contain not only diagrams of a special type, then the mean squared writhe for n crossing diagrams tends to grow linearly with n, but at a faster rate than n, indicating an intrinsic property of alternating knot diagrams. Studying the mean squared writhe of alternating random knot diagrams also provides some insight into the properties of the diagram generating methods used, which is an important area of study in the applications of random knot theory.

  12. Methodology for Safety Assessment Applied to Predisposal Waste Management. Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) 2004–2010)

    International Nuclear Information System (INIS)

    2015-12-01

    Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) (2004–2010) The IAEA’s progamme on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) focused on approaches and mechanisms for application of safety assessment methodologies for the predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts, which have since been incorporated into IAEA Safety Standards Series No. GSG-3, Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste. In 2005, an initial specification was developed for the Safety Assessment Framework (SAFRAN) software tool to apply the SADRWMS flowcharts. In 2008, an in-depth application of the SAFRAN tool and the SADRWMS methodology was carried out on the predisposal management facilities of the Thailand Institute of Nuclear Technology Radioactive Waste Management Centre (TINT Facility). This publication summarizes the content and outcomes of the SADRWMS programme. The Chairman’s Report of the SADRWMS Project and the Report of the TINT test case are provided on the CD-ROM which accompanies this report

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

    International Nuclear Information System (INIS)

    Menuet, Lise; Beauquier, Sophie

    2013-01-01

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

  14. Using Affinity Diagrams to Evaluate Interactive Prototypes

    DEFF Research Database (Denmark)

    Lucero, Andrés

    2015-01-01

    our particular use of affinity diagramming in prototype evaluations. We reflect on a decade’s experience using affinity diagramming across a number of projects, both in industry and academia. Our affinity diagramming process in interaction design has been tailored and consists of four stages: creating...

  15. The amplituhedron from momentum twistor diagrams

    International Nuclear Information System (INIS)

    Bai, Yuntao; He, Song

    2015-01-01

    We propose a new diagrammatic formulation of the all-loop scattering amplitudes/Wilson loops in planar N=4 SYM, dubbed the “momentum-twistor diagrams”. These are on-shell-diagrams obtained by gluing trivalent black and white vertices in momentum twistor space, which, in the reduced diagram case, are known to be related to diagrams in the original twistor space. The new diagrams are manifestly Yangian invariant, and they naturally represent factorization and forward-limit contributions in the all-loop BCFW recursion relations in momentum twistor space, in a fashion that is completely different from those in momentum space. We show how to construct and evaluate momentum-twistor diagrams, and how to use them to obtain tree-level amplitudes and loop-level integrands; in particular the latter involve isolated bubble-structures for loop variables arising from forward limits, or the entangled removal of particles. From each diagram, the generalized “boundary measurement” directly gives the C, D matrices, thus a cell in the amplituhedron associated with the amplitude, and we expect that our diagrammatic representations of the amplitude provide triangulations of the amplituhedron. To demonstrate the computational power of the formalism, we give explicit results for general two-loop integrands, and the cells of the amplituhedron for two-loop MHV amplitudes.

  16. Economic aspects of risk assessment in chemical safety

    Energy Technology Data Exchange (ETDEWEB)

    Drummond, M F; Shannon, H S

    1986-05-01

    This paper considers how the economic aspects of risk assessment in chemical safety can be strengthened. Its main focus is on how economic appraisal techniques, such as cost-benefit and cost-effectiveness analysis, can be adapted to the requirements of the risk-assessment process. Following a discussion of the main methodological issues raised by the use of economic appraisal, illustrated by examples from the health and safety field, a number of practical issues are discussed. These include the consideration of the distribution of costs, effects and benefits, taking account of uncertainty, risk probabilities and public perception, making the appraisal techniques useful to the early stages of the risk-assessment process and structuring the appraisal to permit continuous feedback to the participants in the risk-assessment process. It is concluded that while the way of thinking embodied in economic appraisal is highly relevant to the consideration of choices in chemical safety, the application of these principles in formal analysis of risk reduction procedures presents a more mixed picture. The main suggestions for improvement in the analyses performed are the undertaking of sensitivity analyses of study results to changes in the key assumptions, the presentation of the distribution of costs and benefits by viewpoint, the comparison of health and safety measures in terms of their incremental cost per life-year (or quality-adjusted life-year) gained and the more frequent retrospective review and revision of the economic analyses that are undertaken.

  17. The role of probabilistic safety assessment in the design

    International Nuclear Information System (INIS)

    Green, A.; Ingham, E.L.

    1989-01-01

    The use of probabilistic safety assessment (PSA) for Heysham 2 and Torness marked a major change in the design approach to nuclear safety within the U.K. Design Safety Guidelines incorporating probabilistic safety targets required that design justification would necessitate explicit consideration of the consequence of accidents in relation to their frequency. The paper discusses these safety targets and their implications, the integration of PSA into the design process and an outline of the methodology. The influence of PSA on the design is discussed together with its role in the overall demonstration of reactor safety. (author)

  18. Diagram Techniques in Group Theory

    Science.gov (United States)

    Stedman, Geoffrey E.

    2009-09-01

    Preface; 1. Elementary examples; 2. Angular momentum coupling diagram techniques; 3. Extension to compact simple phase groups; 4. Symmetric and unitary groups; 5. Lie groups and Lie algebras; 6. Polarisation dependence of multiphoton processes; 7. Quantum field theoretic diagram techniques for atomic systems; 8. Applications; Appendix; References; Indexes.

  19. 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

  20. 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.

  1. 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.

  2. 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

  3. Safety Assessment of Talc as Used in Cosmetics.

    Science.gov (United States)

    Fiume, Monice M; Boyer, Ivan; 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

    2015-01-01

    The Cosmetic Ingredient Review Expert Panel (Panel) assessed the safety of talc for use in cosmetics. The safety of talc has been the subject of much debate through the years, partly because the relationship between talc and asbestos is commonly misunderstood. Industry specifications state that cosmetic-grade talc must contain no detectable fibrous, asbestos minerals. Therefore, the large amount of available animal and clinical data the Panel relied on in assessing the safety of talc only included those studies on talc that did not contain asbestos. The Panel concluded that talc is safe for use in cosmetics in the present practices of use and concentration (some cosmetic products are entirely composed of talc). Talc should not be applied to the skin when the epidermal barrier is missing or significantly disrupted. © The Author(s) 2015.

  4. 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 10-Point Initiative to strengthen environment, safety, and health (ES ampersand H) programs, and waste management activities at DOE production, research, and testing facilities. One of the points involved conducting dent Tiger Team Assessments of DOE operating facilities. The Office of Special independent 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 manual documents the processes to be used to perform the ES ampersand H Progress Assessments. It was developed based upon the lessons learned from Tiger Team Assessments, the two pilot Progress Assessments, and Progress Assessments that have been completed. The manual will be updated periodically to reflect lessons learned or changes in policy

  5. A study on the mechanical stress relieving and safety assessment without post-weld heat treatment

    International Nuclear Information System (INIS)

    Xu Jijin; Chen Ligong; Ni Chunzhen

    2007-01-01

    For full welded body valve, the temperature of grommet cannot exceed 150 deg. C in order to prevent it from damaging and assure the tightness and the service life of valve. Therefore, post-weld heat treatment (PWHT) cannot be used to relieve the residual stresses. In this study, the effect of the mechanical stress relieving (MSR) treatment on the residual stresses was studied by the finite element method and experimental work. A pressure and time diagram of MSR treatment was established. A two-dimensional axisymmetric finite element model was used to simulate the residual stresses field. Before and after MSR treatment, the residual stresses on the outer surface were measured by the blind hole drilling method. Finally, the fracture toughness behaviors of weld zone (WZ) and heat affected zone (HAZ) were investigated in terms of crack tip opening displacement (CTOD) according to BS7448 and DNV-OS-C401 fracture toughness tests standards. The safety of the valve in active service was assessed without PWHT. Through comparison and analysis, the axial residual stresses and the hoop residual stresses on the outer surface of valve are mainly tensile. The peak value of tensile stress occurs nearer to the outer surface of the valve. MSR treatment can decrease the peak value of axial residual stresses and hoop residual stresses on the outer surface obviously and make the residual stresses distribution more uniform. The safety of the valve in active service is reliable without PWHT

  6. On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland). Nuclear Safety Dept.

    1996-12-01

    The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.).

  7. On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

    International Nuclear Information System (INIS)

    Hyvaerinen, J.

    1996-12-01

    The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.)

  8. Assessment of Safety Standards for Automotive Electronic Control Systems

    Science.gov (United States)

    2016-06-01

    This report summarizes the results of a study that assessed and compared six industry and government safety standards relevant to the safety and reliability of automotive electronic control systems. These standards include ISO 26262 (Road Vehicles - ...

  9. Plasma-safety assessment model and safety analyses of ITER

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  10. Logic verification system for power plant sequence diagrams

    International Nuclear Information System (INIS)

    Fukuda, Mitsuko; Yamada, Naoyuki; Teshima, Toshiaki; Kan, Ken-ichi; Utsunomiya, Mitsugu.

    1994-01-01

    A logic verification system for sequence diagrams of power plants has been developed. The system's main function is to verify correctness of the logic realized by sequence diagrams for power plant control systems. The verification is based on a symbolic comparison of the logic of the sequence diagrams with the logic of the corresponding IBDs (interlock Block Diagrams) in combination with reference to design knowledge. The developed system points out the sub-circuit which is responsible for any existing mismatches between the IBD logic and the logic realized by the sequence diagrams. Applications to the verification of actual sequence diagrams of power plants confirmed that the developed system is practical and effective. (author)

  11. Assessment of the thermodynamic properties and phase diagram of the Bi–Pd system

    Czech Academy of Sciences Publication Activity Database

    Vřešťál, J.; Pinkas, J.; Watson, A.; Scott, A.; Houserová, Jana; Kroupa, Aleš

    2006-01-01

    Roč. 30, č. 1 (2006), s. 14-17 ISSN 0364-5916 R&D Projects: GA MŠk(CZ) OC 531.002 Institutional research plan: CEZ:AV0Z2041904 Keywords : phase diagram * thermodynamic modelling Subject RIV: BJ - Thermodynamics Impact factor: 1.432, year: 2006

  12. Safety assessment for the above ground storage of Cadmium Safety and Control Rods at the Solid Waste Management Facility

    International Nuclear Information System (INIS)

    Shaw, K.W.

    1993-11-01

    The mission of the Savannah River Site is changing from radioisotope production to waste management and environmental restoration. As such, Reactor Engineering has recently developed a plan to transfer the safety and control rods from the C, K, L, and P reactor disassembly basin areas to the Transuranic (TRU) Waste Storage Pads for long-term, retrievable storage. The TRU pads are located within the Solid Waste Management Facilities at the Savannah River Site. An Unreviewed Safety Question (USQ) Safety Evaluation has been performed for the proposed disassembly basin operations phase of the Cadmium Safety and Control Rod Project. The USQ screening identified a required change to the authorization basis; however, the Proposed Activity does not involve a positive USQ Safety Evaluation. A Hazard Assessment for the Cadmium Safety and Control Rod Project determined that the above-ground storage of the cadmium rods results in no change in hazard level at the TRU pads. A Safety Assessment that specifically addresses the storage (at the TRU pads) phase of the Cadmium Safety and Control Rod Project has been performed. Results of the Safety Assessment support the conclusion that a positive USQ is not involved as a result of the Proposed Activity

  13. Comprehending 3D Diagrams: Sketching to Support Spatial Reasoning.

    Science.gov (United States)

    Gagnier, Kristin M; Atit, Kinnari; Ormand, Carol J; Shipley, Thomas F

    2017-10-01

    Science, technology, engineering, and mathematics (STEM) disciplines commonly illustrate 3D relationships in diagrams, yet these are often challenging for students. Failing to understand diagrams can hinder success in STEM because scientific practice requires understanding and creating diagrammatic representations. We explore a new approach to improving student understanding of diagrams that convey 3D relations that is based on students generating their own predictive diagrams. Participants' comprehension of 3D spatial diagrams was measured in a pre- and post-design where students selected the correct 2D slice through 3D geologic block diagrams. Generating sketches that predicated the internal structure of a model led to greater improvement in diagram understanding than visualizing the interior of the model without sketching, or sketching the model without attempting to predict unseen spatial relations. In addition, we found a positive correlation between sketched diagram accuracy and improvement on the diagram comprehension measure. Results suggest that generating a predictive diagram facilitates students' abilities to make inferences about spatial relationships in diagrams. Implications for use of sketching in supporting STEM learning are discussed. Copyright © 2016 Cognitive Science Society, Inc.

  14. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  15. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  16. Safety assessment guidance in the International Atomic Energy Agency RADWASS Program

    Energy Technology Data Exchange (ETDEWEB)

    Vovk, I.F.; Seitz, R.R.

    1995-12-31

    The IAEA RADWASS programme is aimed at establishing a coherent and comprehensive set of principles and standards for the safe management of waste and formulating the guidelines necessary for their application. A large portion of this programme has been devoted to safety assessments for various waste management activities. Five Safety Guides are planned to be developed to provide general guidance to enable operators and regulators to develop necessary framework for safety assessment process in accordance with international recommendations. They cover predisposal, near surface disposal, geological disposal, uranium/thorium mining and milling waste, and decommissioning and environmental restoration. The Guide on safety assessment for near surface disposal is at the most advanced stage of preparation. This draft Safety Guide contains guidance on description of the disposal system, development of a conceptual model, identification and description of relevant scenarios and pathways, consequence analysis, presentation of results and confidence building. The set of RADWASS publications is currently undergoing in-depth review to ensure a harmonized approach throughout the Safety Series.

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

    NARCIS (Netherlands)

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

    1995-01-01

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

  18. Nuclear utility self-assessment as viewed by the corporate nuclear safety committee

    International Nuclear Information System (INIS)

    Corcoran, W.R.

    1992-01-01

    This paper discusses how corporate nuclear safety committees use the principles of self-assessment to enhance nuclear power plant safety performance. Corporate nuclear safety committees function to advise the senior nuclear power executive on matters affecting nuclear safety. These committees are required by the administrative controls section of the plant technical specifications which are part of the final safety analysis report and the operating license. Committee membership includes senior utility executives, executives from sister utilities, utility senior technical experts, and outside consultants. Current corporate nuclear safety committees often have a finely tuned intuitive feel for self-assessment that they use to probe the underlying opportunities for quality and safety enhancements. The questions prompted by the self-assessment orientation enable the utility line organization members to gain better perspectives on the characteristics of the organizational systems that they manage and work in

  19. Satake diagrams of affine Kac-Moody algebras

    Energy Technology Data Exchange (ETDEWEB)

    Tripathy, L K [S B R Government Womens' College, Berhampur, Orissa 760 001 (India); Pati, K C [Department of Physics, Khallikote College, Berhampur, Orissa 760 001 (India)

    2006-02-10

    Satake diagrams of affine Kac-Moody algebras (untwisted and twisted) are obtained from their Dynkin diagrams. These diagrams give a classification of restricted root systems associated with these algebras. In the case of simple Lie algebras, these root systems and Satake diagrams correspond to symmetric spaces which have recently found many physical applications in quantum integrable systems, quantum transport problems, random matrix theories etc. We hope these types of root systems may have similar applications in theoretical physics in future and may correspond to symmetric spaces analogue of affine Kac-Moody algebras if they exist.

  20. Analytical solutions of linked fault tree probabilistic risk assessments using binary decision diagrams with emphasis on nuclear safety applications

    International Nuclear Information System (INIS)

    Nusbaumer, O. P. M.

    2007-01-01

    This study is concerned with the quantification of Probabilistic Risk Assessment (PRA) using linked Fault Tree (FT) models. Probabilistic Risk assessment (PRA) of Nuclear Power Plants (NPPs) complements traditional deterministic analysis; it is widely recognized as a comprehensive and structured approach to identify accident scenarios and to derive numerical estimates of the associated risk levels. PRA models as found in the nuclear industry have evolved rapidly. Increasingly, they have been broadly applied to support numerous applications on various operational and regulatory matters. Regulatory bodies in many countries require that a PRA be performed for licensing purposes. PRA has reached the point where it can considerably influence the design and operation of nuclear power plants. However, most of the tools available for quantifying large PRA models are unable to produce analytically correct results. The algorithms of such quantifiers are designed to neglect sequences when their likelihood decreases below a predefined cutoff limit. In addition, the rare event approximation (e.g. Moivre's equation) is typically implemented for the first order, ignoring the success paths and the possibility that two or more events can occur simultaneously. This is only justified in assessments where the probabilities of the basic events are low. When the events in question are failures, the first order rare event approximation is always conservative, resulting in wrong interpretation of risk importance measures. Advanced NPP PRA models typically include human errors, common cause failure groups, seismic and phenomenological basic events, where the failure probabilities may approach unity, leading to questionable results. It is accepted that current quantification tools have reached their limits, and that new quantification techniques should be investigated. A novel approach using the mathematical concept of Binary Decision Diagram (BDD) is proposed to overcome these deficiencies

  1. Quantitative reliability assessment for safety critical system software

    International Nuclear Information System (INIS)

    Chung, Dae Won; Kwon, Soon Man

    2005-01-01

    An essential issue in the replacement of the old analogue I and C to computer-based digital systems in nuclear power plants is the quantitative software reliability assessment. Software reliability models have been successfully applied to many industrial applications, but have the unfortunate drawback of requiring data from which one can formulate a model. Software which is developed for safety critical applications is frequently unable to produce such data for at least two reasons. First, the software is frequently one-of-a-kind, and second, it rarely fails. Safety critical software is normally expected to pass every unit test producing precious little failure data. The basic premise of the rare events approach is that well-tested software does not fail under normal routine and input signals, which means that failures must be triggered by unusual input data and computer states. The failure data found under the reasonable testing cases and testing time for these conditions should be considered for the quantitative reliability assessment. We will present the quantitative reliability assessment methodology of safety critical software for rare failure cases in this paper

  2. Safety assessment of botanicals and botanical preparations used as ingredients in food supplements: testing an European Food Safety Authority-tiered approach.

    Science.gov (United States)

    Speijers, Gerrit; Bottex, Bernard; Dusemund, Birgit; Lugasi, Andrea; Tóth, Jaroslav; Amberg-Müller, Judith; Galli, Corrado L; Silano, Vittorio; Rietjens, Ivonne M C M

    2010-02-01

    This article describes results obtained by testing the European Food Safety Authority-tiered guidance approach for safety assessment of botanicals and botanical preparations intended for use in food supplements. Main conclusions emerging are as follows. (i) Botanical ingredients must be identified by their scientific (binomial) name, in most cases down to the subspecies level or lower. (ii) Adequate characterization and description of the botanical parts and preparation methodology used is needed. Safety of a botanical ingredient cannot be assumed only relying on the long-term safe use of other preparations of the same botanical. (iii) Because of possible adulterations, misclassifications, replacements or falsifications, and restorations, establishment of adequate quality control is necessary. (iv) The strength of the evidence underlying concerns over a botanical ingredient should be included in the safety assessment. (v) The matrix effect should be taken into account in the safety assessment on a case-by-case basis. (vi) Adequate data and methods for appropriate exposure assessment are often missing. (vii) Safety regulations concerning toxic contaminants have to be complied with. The application of the guidance approach can result in the conclusion that safety can be presumed, that the botanical ingredient is of safety concern, or that further data are needed to assess safety.

  3. Measuring Best Practices for Workplace Safety, Health, and Well-Being: The Workplace Integrated Safety and Health Assessment.

    Science.gov (United States)

    Sorensen, Glorian; Sparer, Emily; Williams, Jessica A R; Gundersen, Daniel; Boden, Leslie I; Dennerlein, Jack T; Hashimoto, Dean; Katz, Jeffrey N; McLellan, Deborah L; Okechukwu, Cassandra A; Pronk, Nicolaas P; Revette, Anna; Wagner, Gregory R

    2018-05-01

    To present a measure of effective workplace organizational policies, programs, and practices that focuses on working conditions and organizational facilitators of worker safety, health and well-being: the workplace integrated safety and health (WISH) assessment. Development of this assessment used an iterative process involving a modified Delphi method, extensive literature reviews, and systematic cognitive testing. The assessment measures six core constructs identified as central to best practices for protecting and promoting worker safety, health and well-being: leadership commitment; participation; policies, programs, and practices that foster supportive working conditions; comprehensive and collaborative strategies; adherence to federal and state regulations and ethical norms; and data-driven change. The WISH Assessment holds promise as a tool that may inform organizational priority setting and guide research around causal pathways influencing implementation and outcomes related to these approaches.

  4. NASA Aviation Safety Program Systems Analysis/Program Assessment Metrics Review

    Science.gov (United States)

    Louis, Garrick E.; Anderson, Katherine; Ahmad, Tisan; Bouabid, Ali; Siriwardana, Maya; Guilbaud, Patrick

    2003-01-01

    The goal of this project is to evaluate the metrics and processes used by NASA's Aviation Safety Program in assessing technologies that contribute to NASA's aviation safety goals. There were three objectives for reaching this goal. First, NASA's main objectives for aviation safety were documented and their consistency was checked against the main objectives of the Aviation Safety Program. Next, the metrics used for technology investment by the Program Assessment function of AvSP were evaluated. Finally, other metrics that could be used by the Program Assessment Team (PAT) were identified and evaluated. This investigation revealed that the objectives are in fact consistent across organizational levels at NASA and with the FAA. Some of the major issues discussed in this study which should be further investigated, are the removal of the Cost and Return-on-Investment metrics, the lack of the metrics to measure the balance of investment and technology, the interdependencies between some of the metric risk driver categories, and the conflict between 'fatal accident rate' and 'accident rate' in the language of the Aviation Safety goal as stated in different sources.

  5. Development of Safety Culture Assessment Strategy for Korean NPP

    International Nuclear Information System (INIS)

    Park, Jung Hwan; Kim, Jong Hyun

    2014-01-01

    This paper aims at developing the requirements for a method to evaluate the operational safety culture, evaluating currently available methods based on the requirements, and suggesting a method to evaluate and improve the operational safety culture for Korean nuclear power plants. This paper reviews the widely-used methods to assess safety culture for NPPs and their basis. Then, this paper develops the requirements for the method to evaluate operational safety culture for Korean NPPs. Based on these requirements, Korean Safety Culture Indicators (KSCI) and evaluation measures are also suggested. Finally this paper proposes the guidelines to develop improvements to safety culture from the evaluation results

  6. Development of Safety Culture Assessment Strategy for Korean NPP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Hwan; Kim, Jong Hyun [KEPCO, Ulsan (Korea, Republic of)

    2014-08-15

    This paper aims at developing the requirements for a method to evaluate the operational safety culture, evaluating currently available methods based on the requirements, and suggesting a method to evaluate and improve the operational safety culture for Korean nuclear power plants. This paper reviews the widely-used methods to assess safety culture for NPPs and their basis. Then, this paper develops the requirements for the method to evaluate operational safety culture for Korean NPPs. Based on these requirements, Korean Safety Culture Indicators (KSCI) and evaluation measures are also suggested. Finally this paper proposes the guidelines to develop improvements to safety culture from the evaluation results.

  7. Institutionalization of safety re-assessment system for operating nuclear power plants

    International Nuclear Information System (INIS)

    Kim, H. J.; Cho, J. C.; Min, B. K.; Park, J. S.; Jung, H. D.; Oh, K. M.; Kim, W. K.; Lim, J. H.

    1999-01-01

    In this study, in-depth reviews of the foreign countries' experiences and practices in applications of the periodic safety review (PSR), backfitting and license renewal systems as well as the current status of nuclear power safety assurance programs and activities in Korea have been performed to investigate the necessity and feasibility of the application of the systems for the domestic operating nuclear power plants and to establish effective strategy and methodology for the institutionalization of a periodic safety re-assessment system appropriate to both the domestic and international nuclear power environments by incorporating the PSR with the backfitting and license renewal systems. For these purposes, the regulatory policy, fundamental principles and detailed requirements for the institutionalization of the safety re-assessment system and the effective measures for active implementation of the backfitting program have been developed and then a comparative study of benefits and shortcomings has been conducted for the three different models of the periodic safety re-assessment system incorporated with either the license renewal or life extension process, which have been considered as practicable ones in the domestic situation. The model chosen in this study as the most appropriate safety re-assessment system is the one that the re-assessments are performed at the interval of ten years throughout the service life of nuclear power plant and the ten-year license renewal or life extension after the expiration of design life can be permitted based on the regulatory review of the re-assessment results and follow-up measures. Finally, this paper has discussed on the details of the requirements, approach and procedures established for the institutionalization of the periodic safety re-assessment system chosen as the most appropriate one for domestic applications

  8. Comparative assessment of safety indicators for vehicle trajectories on the highway

    NARCIS (Netherlands)

    Mullakkal Babu, F.A.; Wang, M.; Farah, H.; van Arem, B.; Happee, R.

    2017-01-01

    Safety measurement and analysis have been a challenging and well-researched topic in transportation. Conventionally, surrogate safety measures have been used as safety indicators in simulation models for safety assessment, in control formulations for driver assistance systems, and in data analysis

  9. How many standard area diagram sets are needed for accurate disease severity assessment

    Science.gov (United States)

    Standard area diagram sets (SADs) are widely used in plant pathology: a rater estimates disease severity by comparing an unknown sample to actual severities in the SADs and interpolates an estimate as accurately as possible (although some SADs have been developed for categorizing disease too). Most ...

  10. Assessment of thermodynamic properties and phase diagram in the Ag–In–Pd system

    Czech Academy of Sciences Publication Activity Database

    Zemanová, A.; Semenova, O.; Kroupa, Aleš; Vřešťál, J.; Chandrasekaran, K.; Richter, K. W.; Ipser, H.

    2007-01-01

    Roč. 15, č. 1 (2007), s. 77-84 ISSN 0966-9795 R&D Projects: GA MŠk(CZ) OC 532.001 Institutional research plan: CEZ:AV0Z20410507 Keywords : phase diagrams * ternary alloy systems * prediction Subject RIV: BJ - Thermodynamics Impact factor: 2.219, year: 2007

  11. Contingency diagrams as teaching tools

    OpenAIRE

    Mattaini, Mark A.

    1995-01-01

    Contingency diagrams are particularly effective teaching tools, because they provide a means for students to view the complexities of contingency networks present in natural and laboratory settings while displaying the elementary processes that constitute those networks. This paper sketches recent developments in this visualization technology and illustrates approaches for using contingency diagrams in teaching.

  12. Delimiting diagrams

    NARCIS (Netherlands)

    Oostrom, V. van

    2004-01-01

    We introduce the unifying notion of delimiting diagram. Hitherto unrelated results such as: Minimality of the internal needed strategy for orthogonal first-order term rewriting systems, maximality of the limit strategy for orthogonal higher-order pattern rewrite systems (with maximality of the

  13. Safety assessment of Novi Han radioactive waste repository - features, problems, results and perspectives

    International Nuclear Information System (INIS)

    Mateeva, M.

    2000-01-01

    This paper summarizes the work done and the achievements reached in the Novi Han radioactive waste repository safety assessment within the IAEA Model Project 'Increasing the safety of Novi Han radioactive waste repository BUL 4/005'. The overall safety assessment has a wide context, but the work reported here relates only to some details and results concerning the development and implementation of the appropriate methodology approach, model and computer code used for the calculations. Different steps and procedures are included for a better practical understanding of the obtained results during the safety assessment performance. The methodology approach is widely based on an international experience in safety analysis and implemented for evaluation computer code AMBER, which is one of the recommended from the safety assessments experts. (author)

  14. 76 FR 74723 - New Car Assessment Program (NCAP); Safety Labeling

    Science.gov (United States)

    2011-12-01

    ... [Docket No. NHTSA 2010-0025] RIN 2127-AK51 New Car Assessment Program (NCAP); Safety Labeling AGENCY... NHTSA's regulation on vehicle labeling of safety rating information to reflect the enhanced NCAP ratings... Traffic Safety Administration under the enhanced NCAP testing and rating program. * * * * * (e) * * * (4...

  15. ROLE OF UML SEQUENCE DIAGRAM CONSTRUCTS IN OBJECT LIFECYCLE CONCEPT

    Directory of Open Access Journals (Sweden)

    Miroslav Grgec

    2007-06-01

    Full Text Available When modeling systems and using UML concepts, a real system can be viewed in several ways. The RUP (Rational Unified Process defines the "4 + 1 view": 1. Logical view (class diagram (CD, object diagram (OD, sequence diagram (SD, collaboration diagram (COD, state chart diagram (SCD, activity diagram (AD, 2.Process view (use case diagram, CD, OD, SD, COD, SCD, AD, 3. Development view (package diagram, component diagram, 4. Physical view (deployment diagram, and 5. Use case view (use case diagram, OD, SD, COD, SCD, AD which combines the four mentioned above. With sequence diagram constructs we are describing object behavior in scope of one use case and their interaction. Each object in system goes through a so called lifecycle (create, supplement object with data, use object, decommission object. The concept of the object lifecycle is used to understand and formalize the behavior of objects from creation to deletion. With help of sequence diagram concepts our paper will describe the way of interaction modeling between objects through lifeline of each of them, and their importance in software development.

  16. A novel decision diagrams extension method

    International Nuclear Information System (INIS)

    Li, Shumin; Si, Shubin; Dui, Hongyan; Cai, Zhiqiang; Sun, Shudong

    2014-01-01

    Binary decision diagram (BDD) is a graph-based representation of Boolean functions. It is a directed acyclic graph (DAG) based on Shannon's decomposition. Multi-state multi-valued decision diagram (MMDD) is a natural extension of BDD for the symbolic representation and manipulation of the multi-valued logic functions. This paper proposes a decision diagram extension method based on original BDD/MMDD while the scale of a reliability system is extended. Following a discussion of decomposition and physical meaning of BDD and MMDD, the modeling method of BDD/MMDD based on original BDD/MMDD is introduced. Three case studies are implemented to demonstrate the presented methods. Compared with traditional BDD and MMDD generation methods, the decision diagrams extension method is more computationally efficient as shown through the running time

  17. Initiation and arrest - two approaches to pressure vessel safety

    International Nuclear Information System (INIS)

    Brumovsky, M.; Filip, R.; Stepanek, S.

    1976-01-01

    The safety analysis is described of the reactor pressure vessel related to brittle fracture based on the fracture mechanics theory using two different approximations, i.e., the Crack Arrest Temperature (CAT) or Nil Ductility Temperature (NDT), and fracture toughness. The variation of CAT with stress was determined for different steel specimens of 120 to 200 mm in thickness. A diagram is shown of CAT variation with stress allowing the determination of crack arrest temperature for all types of commonly used steels independently of the NDT initial value. The diagram also shows that the difference between fracture transition elastic (FTE) and NDT depends on the type of material and determines the value of the ΔTsub(sigma) factor typical of the safety coefficient. The so-called fracture toughness reference value Ksub(IR) is recommended for the computation of pressure vessel criticality. Also shown is a defect analysis diagram which may be used for the calculation of pressure vessel safety prior to and during operation and which may also be used in making the decision on what crack sizes are critical, what cracks may be arrested and what cracks are likely to expand. The diagram is also important for the fact that it is material-independent and may be employed for the estimates of pre-operational and operational inspections and for pressure vessel life prediction. It is generally applicable to materials of greater thickness in the region where the validity of linear elastic fracture mechanics is guaranteed. (J.P.)

  18. Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    . The standards are also applied by regulatory bodies and operators around the world to enhance safety in nuclear power generation and in nuclear applications in medicine, industry, agriculture and research. Safety is not an end in itself but a prerequisite for the purpose of the protection of people in all States and of the environment - now and in the future. The risks associated with ionizing radiation must be assessed and controlled without unduly limiting the contribution of nuclear energy to equitable and sustainable development. Governments, regulatory bodies and operators everywhere must ensure that nuclear material and radiation sources are used beneficially, safely and ethically. The IAEA safety standards are designed to facilitate this, and I encourage all Member States to make use of them.

  19. Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

    International Nuclear Information System (INIS)

    2011-01-01

    . The standards are also applied by regulatory bodies and operators around the world to enhance safety in nuclear power generation and in nuclear applications in medicine, industry, agriculture and research. Safety is not an end in itself but a prerequisite for the purpose of the protection of people in all States and of the environment - now and in the future. The risks associated with ionizing radiation must be assessed and controlled without unduly limiting the contribution of nuclear energy to equitable and sustainable development. Governments, regulatory bodies and operators everywhere must ensure that nuclear material and radiation sources are used beneficially, safely and ethically. The IAEA safety standards are designed to facilitate this, and I encourage all Member States to make use of them.

  20. Validation of BS7910:2005 failure assessment diagrams for cracked square hollow section T-, Y- and K-joints

    International Nuclear Information System (INIS)

    Lie, S.T.; Yang, Z.M.; Gho, W.M.

    2009-01-01

    This paper describes the usage of finite element (FE) analyses results to validate the standard BS7910 assessment procedure for the safe design of cracked square hollow section (SHS) T-, Y- and K-joints. In the study, the actual 3D surface cracks obtained from previous fatigue tests have been included in the FE models. An automatic mesh generation program is then developed and used to produce the failure assessment diagram (FAD) through the J-integral method. The ultimate strength of uncracked SHS joints with reduced load bearing areas have been referenced to derive the plastic collapse loads of cracked SHS joints for the development of FAD. These loads have been validated against the previous experimental results. In comparison with the existing standard BS7910 Level 2A/3A FAD curve and the proposed assessment procedure for circular hollow section joints, it is found that a plastic collapse load with a penalty factor of 1.05 will be sufficient for the safe assessment of cracked SHS T, Y, and K-joints under brace end axial loading

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

    International Nuclear Information System (INIS)

    1997-01-01

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

  2. Data report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    2010-12-01

    This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

  3. Data report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

  4. Psychometric model for safety culture assessment in nuclear research facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, C.S. do, E-mail: claudio.souza@ctmsp.mar.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), Av. Professor Lineu Prestes 2468, 05508-000 São Paulo, SP (Brazil); Andrade, D.A., E-mail: delvonei@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil); Mesquita, R.N. de, E-mail: rnavarro@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil)

    2017-04-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  5. Psychometric model for safety culture assessment in nuclear research facilities

    International Nuclear Information System (INIS)

    Nascimento, C.S. do; Andrade, D.A.; Mesquita, R.N. de

    2017-01-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  6. 78 FR 14912 - International Aviation Safety Assessment (IASA) Program Change

    Science.gov (United States)

    2013-03-08

    ... Aviation Safety Assessment (IASA) Program Change AGENCY: Federal Aviation Administration (FAA), DOT. ACTION..., into the U.S., or codeshare with a U.S. air carrier, complies with international aviation safety... subject to that country's aviation safety oversight can serve the United States using its own aircraft or...

  7. Fuel and canister process report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    Werme, Lars; Lilja, Christina

    2010-12-01

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It 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/

  8. Fuel and canister process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars; Lilja, Christina (eds.)

    2010-12-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It 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/

  9. Probabilistic safety assessment in radioactive waste disposal

    International Nuclear Information System (INIS)

    Robinson, P.C.

    1987-07-01

    Probabilistic safety assessment codes are now widely used in radioactive waste disposal assessments. This report gives an overview of the current state of the field. The relationship between the codes and the regulations covering radioactive waste disposal is discussed and the characteristics of current codes is described. The problems of verification and validation are considered. (author)

  10. Assessment of Electrical Safety Beliefs and Practices: A Case Study

    Directory of Open Access Journals (Sweden)

    S. Boubaker

    2017-12-01

    Full Text Available In this paper, the electrical safety beliefs and practices in Hail region, Saudi Arabia, have been assessed. Based on legislative recommendations and rules applied in Saudi Arabia, on official statistics regarding the electricity-caused accidents and on the analysis of more than 200 photos captured in Hail (related to electrical safety, a questionnaire composed of 36 questions (10 for the respondents information, 16 for the home safety culture and 10 for the electrical devices purchasing culture has been devised and distributed to residents. 228 responses have been collected and analyzed. Using a scale similar to the one adopted for a university student GPA calculation, the electrical safety level (ESL in Hail region has been found to be 0.76 (in a scale of 4 points which is a very low score and indicates a poor electrical safety culture. Several recommendations involving different competent authorities have been proposed. Future work will concern the assessment of safety in industrial companies in Hail region.

  11. Scheil-Gulliver Constituent Diagrams

    Science.gov (United States)

    Pelton, Arthur D.; Eriksson, Gunnar; Bale, Christopher W.

    2017-06-01

    During solidification of alloys, conditions often approach those of Scheil-Gulliver cooling in which it is assumed that solid phases, once precipitated, remain unchanged. That is, they no longer react with the liquid or with each other. In the case of equilibrium solidification, equilibrium phase diagrams provide a valuable means of visualizing the effects of composition changes upon the final microstructure. In the present study, we propose for the first time the concept of Scheil-Gulliver constituent diagrams which play the same role as that in the case of Scheil-Gulliver cooling. It is shown how these diagrams can be calculated and plotted by the currently available thermodynamic database computing systems that combine Gibbs energy minimization software with large databases of optimized thermodynamic properties of solutions and compounds. Examples calculated using the FactSage system are presented for the Al-Li and Al-Mg-Zn systems, and for the Au-Bi-Sb-Pb system and its binary and ternary subsystems.

  12. Edible safety requirements and assessment standards for agricultural genetically modified organisms.

    Science.gov (United States)

    Deng, Pingjian; Zhou, Xiangyang; Zhou, Peng; Du, Zhong; Hou, Hongli; Yang, Dongyan; Tan, Jianjun; Wu, Xiaojin; Zhang, Jinzhou; Yang, Yongcun; Liu, Jin; Liu, Guihua; Li, Yonghong; Liu, Jianjun; Yu, Lei; Fang, Shisong; Yang, Xiaoke

    2008-05-01

    This paper describes the background, principles, concepts and methods of framing the technical regulation for edible safety requirement and assessment of agricultural genetically modified organisms (agri-GMOs) for Shenzhen Special Economic Zone in the People's Republic of China. It provides a set of systematic criteria for edible safety requirements and the assessment process for agri-GMOs. First, focusing on the degree of risk and impact of different agri-GMOs, we developed hazard grades for toxicity, allergenicity, anti-nutrition effects, and unintended effects and standards for the impact type of genetic manipulation. Second, for assessing edible safety, we developed indexes and standards for different hazard grades of recipient organisms, for the influence of types of genetic manipulation and hazard grades of agri-GMOs. To evaluate the applicability of these criteria and their congruency with other safety assessment systems for GMOs applied by related organizations all over the world, we selected some agri-GMOs (soybean, maize, potato, capsicum and yeast) as cases to put through our new assessment system, and compared our results with the previous assessments. It turned out that the result of each of the cases was congruent with the original assessment.

  13. Safety assessment of emergency power systems for nuclear power plants

    International Nuclear Information System (INIS)

    1992-01-01

    This publication is intended to assist the safety assessor within a regulatory body, or one working as a consultant, in assessing the safety of a given design of the emergency power systems (EPS) for a nuclear power plant. The present publication refers closely to the NUSS Safety Guide 50-SG-D7 (Rev. 1), Emergency Power Systems at Nuclear Power Plants. It covers therefore exactly the same technical subject as that Safety Guide. In view of its objective, however, it attempts to help in the evaluation of possible technical solutions which are intended to fulfill the safety requirements. Section 2 clarifies the scope further by giving an outline of the assessment steps in the licensing process. After a general outline of the assessment process in relation to the licensing of a nuclear power plant, the publication is divided into two parts. First, all safety issues are presented in the form of questions that have to be answered in order for the assessor to be confident of a safe design. The second part presents the same topics in tabulated form, listing the required documentation which the assessor has to consult and those international and national technical standards pertinent to the topics. An extensive reference list provides information on standards. 1 tab

  14. Reactor Safety Assessment System--A situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

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

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base that uses the parametric values, the known operator actions, and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant

  15. Reactor Safety Assessment System: a situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

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

    1985-04-01

    The Reactor Safety Assessment System is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base which uses the parametric values, the known operator actions and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant. 5 figs

  16. Nuclear safety chains

    International Nuclear Information System (INIS)

    Robbins, M.C.; Eames, G.F.; Mayell, J.R.

    1981-01-01

    An original scheme has been developed for expressing the complex interrelationships associated with the engineered safeguards provided for a nuclear power station. This management tool, based upon network diagrams called Nuclear Safety Chains, looks at the function required of a particular item of safety plant, defines all of the vital supplies and support features necessary for successful operation, and expresses them in visual form, to facilitate analysis and optimisation for operations and maintenance staff. The safety chains are confined to manual schemes at present, although they are designed to be compatible with modern computer techniques. Their usefulness with any routine maintenance planning application on high technology plant is already being appreciated. (author)

  17. Operations space diagram for ECRH and ECCD

    DEFF Research Database (Denmark)

    Bindslev, H.

    2004-01-01

    at the design phase. At the operations phase it may also prove useful in setting up experimental scenarios by showing operational possibilities, avoiding the need for survey type ray-tracing at the initial planning stages. The diagram may also serve the purpose of communicating operational possibilities to non......A Clemmov-Mullaly-Allis (CMA) type diagram, the ECW-CMA diagram, for representing the operational possibilities of electron cyclotron heating and current drive (ECRH/ECCD) systems for fusion plasmas is presented. In this diagram, with normalized density and normalized magnetic field coordinates......, the parameter range in which it is possible to achieve a given task (e.g. O-mode current drive for stabilizing a neoclassical tearing mode) appears as a region. With also the Greenwald density limit shown, this diagram condenses the information on operational possibilities, facilitating the overview required...

  18. [Safety assessment of foods derived from genetically modified plants].

    Science.gov (United States)

    Pöting, A; Schauzu, M

    2010-06-01

    The placing of genetically modified plants and derived food on the market falls under Regulation (EC) No. 1829/2003. According to this regulation, applicants need to perform a safety assessment according to the Guidance Document of the Scientific Panel on Genetically Modified Organisms of the European Food Safety Authority (EFSA), which is based on internationally agreed recommendations. This article gives an overview of the underlying legislation as well as the strategy and scientific criteria for the safety assessment, which should generally be based on the concept of substantial equivalence and carried out in relation to an unmodified conventional counterpart. Besides the intended genetic modification, potential unintended changes also have to be assessed with regard to potential adverse effects for the consumer. All genetically modified plants and derived food products, which have been evaluated by EFSA so far, were considered to be as safe as products derived from the respective conventional plants.

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

    International Nuclear Information System (INIS)

    2015-08-01

    Technical Volume 1 of this report has described what happened during the accident at the Fukushima Daiichi nuclear power plant (NPP). This volume begins (Section 2.1) with a review of how the design basis of the site for external events was assessed initially and then reassessed over the life of the NPP. The section also describes the physical changes that were made to the units as a result. The remainder of the volume describes the treatment of beyond design basis events in the safety assessment of the site, the accident management provisions, the effectiveness of regulatory programmes, human and organizational factors and the safety culture, and the role of operating experience. Further background information is contained in three annexes included on the CD-ROM of this Technical Volume which describe analytical investigations of the accident along with information on topics such as system performance, defence in depth and severe accident phenomena. Section 2.2 provides an assessment of the systems that failed, resulting in a failure to maintain the fundamental safety functions in Units 1–3, which were in operation at the time of the tsunami and in which the reactor pressure vessels (RPV) and containment vessels failed. The section also describes Units 4-6, which were shut down at the time of the tsunami, and the site’s central spent fuel storage facility. Section 2.3 discusses the probabilistic and deterministic safety assessments of beyond design basis accidents (BDBAs) that had been performed for the plant and the insights from these assessments that had led to changes in the plant’s design. The section pays particular attention to the assessment of extreme natural hazards, such as the one which led to the total loss of AC power supply on the site. The additional loss of DC power supply in Units 1 and 2 played a key role in the progression of the accident because it impeded the diagnosis of plant conditions and made the operators unaware of the status of

  20. Penguin-like diagrams from the standard model

    International Nuclear Information System (INIS)

    Ping, Chia Swee

    2015-01-01

    The Standard Model is highly successful in describing the interactions of leptons and quarks. There are, however, rare processes that involve higher order effects in electroweak interactions. One specific class of processes is the penguin-like diagram. Such class of diagrams involves the neutral change of quark flavours accompanied by the emission of a gluon (gluon penguin), a photon (photon penguin), a gluon and a photon (gluon-photon penguin), a Z-boson (Z penguin), or a Higgs-boson (Higgs penguin). Such diagrams do not arise at the tree level in the Standard Model. They are, however, induced by one-loop effects. In this paper, we present an exact calculation of the penguin diagram vertices in the ‘tHooft-Feynman gauge. Renormalization of the vertex is effected by a prescription by Chia and Chong which gives an expression for the counter term identical to that obtained by employing Ward-Takahashi identity. The on-shell vertex functions for the penguin diagram vertices are obtained. The various penguin diagram vertex functions are related to one another via Ward-Takahashi identity. From these, a set of relations is obtained connecting the vertex form factors of various penguin diagrams. Explicit expressions for the gluon-photon penguin vertex form factors are obtained, and their contributions to the flavor changing processes estimated

  1. Penguin-like diagrams from the standard model

    Energy Technology Data Exchange (ETDEWEB)

    Ping, Chia Swee [High Impact Research, University of Malaya, 50603 Kuala Lumpur (Malaysia)

    2015-04-24

    The Standard Model is highly successful in describing the interactions of leptons and quarks. There are, however, rare processes that involve higher order effects in electroweak interactions. One specific class of processes is the penguin-like diagram. Such class of diagrams involves the neutral change of quark flavours accompanied by the emission of a gluon (gluon penguin), a photon (photon penguin), a gluon and a photon (gluon-photon penguin), a Z-boson (Z penguin), or a Higgs-boson (Higgs penguin). Such diagrams do not arise at the tree level in the Standard Model. They are, however, induced by one-loop effects. In this paper, we present an exact calculation of the penguin diagram vertices in the ‘tHooft-Feynman gauge. Renormalization of the vertex is effected by a prescription by Chia and Chong which gives an expression for the counter term identical to that obtained by employing Ward-Takahashi identity. The on-shell vertex functions for the penguin diagram vertices are obtained. The various penguin diagram vertex functions are related to one another via Ward-Takahashi identity. From these, a set of relations is obtained connecting the vertex form factors of various penguin diagrams. Explicit expressions for the gluon-photon penguin vertex form factors are obtained, and their contributions to the flavor changing processes estimated.

  2. Safety assessment of Vitis vinifera (grape)-derived ingredients as used in cosmetics.

    Science.gov (United States)

    Fiume, Monice M; 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

    2014-01-01

    The Cosmetic Ingredient Review Expert Panel (Panel) assessed the safety of 24 Vitis vinifera (grape)-derived ingredients and found them safe in the present practices of use and concentration in cosmetics. These ingredients function in cosmetics mostly as skin-conditioning agents, but some function as antioxidants, flavoring agents, and/or colorants. The Panel reviewed the available animal and clinical data to determine the safety of these ingredients. Additionally, some constituents of grapes have been assessed previously for safety as cosmetic ingredients by the Panel, and others are compounds that have been discussed in previous Panel safety assessments. © The Author(s) 2014.

  3. Self-assessment of safety culture in nuclear installations. Highlights and good practices

    International Nuclear Information System (INIS)

    2002-11-01

    This report summarizes the findings of two IAEA Technical Committee Meetings on Safety Culture Self-Assessment Highlights and Good Practices. The meetings took place on 3-5 June 1998 and 23-25 October 2000 in Vienna, and involved an international cross-section of representatives who participated both in plenary discussions and working groups. The purpose of the meetings was to discuss the practical implications of evolutionary changes in the development of safety culture, and to share international experience, particularly on the methods used for the assessment of safety culture and good practices for its enhancement in an organization. The working groups were allocated specific topics for discussion, which included the following: organizational factors influencing the implementation of actions to improve safety culture; how to measure, effectively, progress in implementing solutions to safety culture problems; the symptoms of a weakening safety culture; the suitability of different methods for assessing safety culture; the achievement of sustainable improvements in safety culture using the results of assessment; the potential threats to the continuation of a strong safety culture in an organization from the many challenges facing the nuclear industry. The working groups, when appropriate, considered issues from both the utility's and the regulator's perspectives. This report will be of interest to all organizations who wish to assess and achieve a strong and sustainable safety culture. This includes not only nuclear power plants, but also other sectors of the nuclear industry such as uranium mines and mills, nuclear fuel fabrication facilities, nuclear waste repositories, research reactors, accelerators, radiography facilities, etc. The report specifically supplements other IAEA publications on this subject

  4. A 2-year study of patient safety competency assessment in 29 clinical laboratories.

    Science.gov (United States)

    Reed, Robyn C; Kim, Sara; Farquharson, Kara; Astion, Michael L

    2008-06-01

    Competency assessment is critical for laboratory operations and is mandated by the Clinical Laboratory Improvement Amendments of 1988. However, no previous reports describe methods for assessing competency in patient safety. We developed and implemented a Web-based tool to assess performance of 875 laboratory staff from 29 laboratories in patient safety. Question categories included workplace culture, categorizing error, prioritization of patient safety interventions, strength of specific interventions, and general patient safety concepts. The mean score was 85.0%, with individual scores ranging from 56% to 100% and scores by category from 81.3% to 88.6%. Of the most difficult questions (laboratory technologists. Computer-based competency assessments help laboratories identify topics for continuing education in patient safety.

  5. Common phase diagram for low-dimensional superconductors

    International Nuclear Information System (INIS)

    Michalak, Rudi

    2003-01-01

    A phenomenological phase diagram which has been derived for high-temperature superconductors from NMR Knight-shift measurements of the pseudogap is compared to the phase diagram that is obtained for organic superconductors and spin-ladder superconductors, both low-dimensional systems. This is contrasted to the phase diagram of some Heavy Fermion superconductors, i.e. superconductors not constrained to a low dimensionality

  6. South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

    International Nuclear Information System (INIS)

    Joubert, J.; Kohtz, N.; Coe, I.

    2008-01-01

    It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /l/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient

  7. Feynman diagrams without Feynman parameters

    International Nuclear Information System (INIS)

    Mendels, E.

    1978-01-01

    Dimensionally regularized Feynman diagrams are represented by means of products of k-functions. The infinite part of these diagrams is found very easily, also if they are overlapping, and the separation of the several kinds of divergences comes out quite naturally. Ward identities are proven in a transparent way. Series expansions in terms of the external momenta and their inner products are possible

  8. Phase diagrams of the elements

    International Nuclear Information System (INIS)

    Young, D.A.

    1975-01-01

    A summary of the pressure-temperature phase diagrams of the elements is presented, with graphs of the experimentally determined solid-solid phase boundaries and melting curves. Comments, including theoretical discussion, are provided for each diagram. The crystal structure of each solid phase is identified and discussed. This work is aimed at encouraging further experimental and theoretical research on phase transitions in the elements

  9. Recent Trends In The Methods Of Safety Assessment Of Rad Waste Treatment And Disposal

    International Nuclear Information System (INIS)

    Mahmoud, N.S.

    2012-01-01

    Radioactive waste management system involves a huge variety of processes and activities. This includes; collection and segregation, pretreatment, treatment, conditioning, storage and finally disposal. To assure the safety of the different facility of each step in the waste management system, the operator should prepare a safety analysis report to be assessed by the national regulatory body. The content of the safety analysis report must include all data about the site, facility design, operational phase, waste materials, and safety assessment methodologies. Safety assessment methodologies are iterative processes involving site-specific, prospective modeling evaluations of the pre-operational, operational, and post-closure time in case of disposal facilities. The safety assessment focuses primarily on a decision about compliance with performance objectives, rather than the much more difficult problem of predicting actual radiological impacts on the public at far future times. The recent organization processes of the safety assessment are improved by the ISAM working group from IAEA for waste disposal site. These safety assessment methodologies have been modified within SADRWMS IAEA project for the establishment of safety methodologies for the pre-disposal facilities (treatment and storage facilities) and the disposal site.

  10. Safety assessment of genetically modified foods

    NARCIS (Netherlands)

    Kleter, G.A.; Noordam, M.Y.

    2016-01-01

    The cultivation of genetically modified (GM) crops has steadily increased since their introduction to the market in the mid-1990s. Before these crops can be grown and sold they have to obtain regulatory approval in many countries, the process of which includes a pre-market safety assessment. The

  11. Qualified Presumption of Safety (QPS) is a generic risk assessment approach applied by the European Food Safety Authority (EFSA)

    DEFF Research Database (Denmark)

    Leuschner, R. G. K.; Robinson, T. P.; Hugas, M.

    2010-01-01

    Qualified Presumption of Safety (QPS) is a generic risk assessment approach applied by the European Food Safety Authority (EFSA) to notified biological agents aiming at simplifying risk assessments across different scientific Panels and Units. The aim of this review is to outline the implementation...... and value of the QPS assessment for EFSA and to explain its principles such as the unambiguous identity of a taxonomic unit, the body of knowledge including potential safety concerns and how these considerations lead to a list of biological agents recommended for QPS which EFSA keeps updated through...

  12. 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

  13. 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

  14. HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

    Energy Technology Data Exchange (ETDEWEB)

    EVANS, C B

    2004-12-21

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S&M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard.

  15. Homotopy theory of modules over diagrams of rings

    Directory of Open Access Journals (Sweden)

    J. P. C. Greenlees

    2014-09-01

    Full Text Available Given a diagram of rings, one may consider the category of modules over them. We are interested in the homotopy theory of categories of this type: given a suitable diagram of model categories ℳ( (as runs through the diagram, we consider the category of diagrams where the object ( at comes from ℳ(. We develop model structures on such categories of diagrams and Quillen adjunctions that relate categories based on different diagram shapes. Under certain conditions, cellularizations (or right Bousfield localizations of these adjunctions induce Quillen equivalences. As an application we show that a cellularization of a category of modules over a diagram of ring spectra (or differential graded rings is Quillen equivalent to modules over the associated inverse limit of the rings. Another application of the general machinery here is given in work by the authors on algebraic models of rational equivariant spectra. Some of this material originally appeared in the preprint “An algebraic model for rational torus-equivariant stable homotopy theory”, arXiv:1101.2511, but has been generalized here.

  16. The application of diagrams in architectural design

    Directory of Open Access Journals (Sweden)

    Dulić Olivera

    2014-01-01

    Full Text Available Diagrams in architecture represent the visualization of the thinking process, or selective abstraction of concepts or ideas translated into the form of drawings. In addition, they provide insight into the way of thinking about and in architecture, thus creating a balance between the visual and the conceptual. The subject of research presented in this paper are diagrams as a specific kind of architectural representation, and possibilities and importance of their application in the design process. Diagrams are almost old as architecture itself, and they are an element of some of the most important studies of architecture during all periods of history - which results in a large number of different definitions of diagrams, but also very different conceptualizations of their features, functions and applications. The diagrams become part of contemporary architectural discourse during the eighties and nineties of the twentieth century, especially through the work of architects like Bernard Tschumi, Peter Eisenman, Rem Koolhaas, SANAA and others. The use of diagrams in the design process allows unification of some of the essential aspects of the profession: architectural representation and design process, as well as the question of the concept of architectural and urban design at a time of rapid changes at all levels of contemporary society. The aim of the research is the analysis of the diagram as a specific medium for processing large amounts of information that the architect should consider and incorporate into the architectural work. On that basis, it is assumed that an architectural diagram allows the creator the identification and analysis of specific elements or ideas of physical form, thereby constantly maintaining concept of the integrity of the architectural work.

  17. The 7 basic tools of quality applied to radiological safety

    International Nuclear Information System (INIS)

    Gonzalez F, J.A.

    1991-01-01

    This work seeks to establish a series of correspondences among the search of the quality and the optimization of the doses received by the occupationally exposed personnel. There are treated about the seven basic statistic tools of the quality: the Pareto technique, Cause effect diagrams, Stratification, Verification sheet, Histograms, Dispersion diagrams and Graphics and control frames applied to the Radiological Safety

  18. Reading fitness landscape diagrams through HSAB concepts

    Energy Technology Data Exchange (ETDEWEB)

    Vigneresse, Jean-Louis, E-mail: jean-louis.vigneresse@univ-lorraine.fr

    2014-10-31

    Highlights: • Qualitative information from HSAB descriptors. • 2D–3D diagrams using chemical descriptors (χ, η, ω, α) and principles (MHP, mEP, mPP). • Estimate of the energy exchange during reaction paths. • Examples from complex systems (geochemistry). - Abstract: Fitness landscapes are conceived as range of mountains, with local peaks and valleys. In terms of potential, such topographic variations indicate places of local instability or stability. The chemical potential, or electronegativity, its value changed of sign, carries similar information. In addition to chemical descriptors defined through hard-soft acid-base (HSAB) concepts and computed through density functional theory (DFT), the principles that rule chemical reactions allow the design of such landscape diagrams. The simplest diagram uses electrophilicity and hardness as coordinates. It allows examining the influence of maximum hardness or minimum electrophilicity principles. A third dimension is introduced within such a diagram by mapping the topography of electronegativity, polarizability or charge exchange. Introducing charge exchange during chemical reactions, or mapping a third parameter (f.i. polarizability) reinforces the information carried by a simple binary diagram. Examples of such diagrams are provided, using data from Earth Sciences, simple oxides or ligands.

  19. Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

    International Nuclear Information System (INIS)

    Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A.

    1990-01-01

    FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG ampersand G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort

  20. Validity of instruments to assess students' travel and pedestrian safety

    Directory of Open Access Journals (Sweden)

    Baranowski Tom

    2010-05-01

    Full Text Available Abstract Background Safe Routes to School (SRTS programs are designed to make walking and bicycling to school safe and accessible for children. Despite their growing popularity, few validated measures exist for assessing important outcomes such as type of student transport or pedestrian safety behaviors. This research validated the SRTS school travel survey and a pedestrian safety behavior checklist. Methods Fourth grade students completed a brief written survey on how they got to school that day with set responses. Test-retest reliability was obtained 3-4 hours apart. Convergent validity of the SRTS travel survey was assessed by comparison to parents' report. For the measure of pedestrian safety behavior, 10 research assistants observed 29 students at a school intersection for completion of 8 selected pedestrian safety behaviors. Reliability was determined in two ways: correlations between the research assistants' ratings to that of the Principal Investigator (PI and intraclass correlations (ICC across research assistant ratings. Results The SRTS travel survey had high test-retest reliability (κ = 0.97, n = 96, p Conclusions These validated instruments can be used to assess SRTS programs. The pedestrian safety behavior checklist may benefit from further formative work.

  1. Assessment of safety of the nuclear installations of the world

    International Nuclear Information System (INIS)

    Thomas, B.A.; Pozniakov, N.; Banga, U.

    1992-01-01

    Incidents and accidents periodically remind us that preventive measures at nuclear installations are not fully reliable. Although sound design is widely recognized to be prerequisite for safe operation, it is not sufficient. An active management that compensates for the weak aspects of the installations design by redundant operational provisions, is the key factor to ensure safe operation. Safety of nuclear installations cannot be assessed on an emotional basis. Since 1986, accurate safety assessment techniques based on an integrated approach to operational safety have been made available by the ASSET services and are applicable to any industrial process dealing with nuclear materials. The ASSET methodology enables to eliminate in advance the Root Causes of the future accidents by introducing practical safety culture principles in the current managerial practices

  2. 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)

  3. 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)

  4. Safety assessment of genetically modified crops

    International Nuclear Information System (INIS)

    Atherton, Keith T.

    2002-01-01

    The development of genetically modified (GM) crops has prompted widespread debate regarding both human safety and environmental issues. Food crops produced by modern biotechnology using recombinant techniques usually differ from their conventional counterparts only in respect of one or a few desirable genes, as opposed to the use of traditional breeding methods which mix thousands of genes and require considerable efforts to select acceptable and robust hybrid offspring. The difficulties of applying traditional toxicological testing and risk assessment procedures to whole foods are discussed along with the evaluation strategies that are used for these new food products to ensure the safety of these products for the consumer

  5. Assessing progress in the development of safety culture

    International Nuclear Information System (INIS)

    Rotaru, I.; Ghita, S.; Biro, L.

    2002-01-01

    This paper is focussed on the organizational culture and learning processes required for the implementation of all aspects of safety culture. There is no prescriptive formula for improving safety culture. However, some common characteristics and practices are emerging that can be adopted by organizations in order to make progress. The paper refers to some approaches that have been successful in a number of countries. The experience of the international nuclear industry in the development and improvement of safety culture could be extended and found useful in other nuclear activities, irrespective of scale. The examples given of specific practice cover a wide range of activities including analysis of events, the regulatory approach on safety culture, employee participation and safety performance measures. Many of these practices may be relevant to smaller organizations and could contribute to improving safety culture, whatever the size of the organization. The most effective approach is to pursue a range of practices that can be mutually supportive in the development of a progressive safety culture, supported by professional standards, organizational and management commitment. Some guidance is also given on the assessment of safety culture and on the detection of a weakening safety culture. Few suggestions for accelerating the safety culture development and improvement process are also provided. (author)

  6. Crane Safety Assessment Method Based on Entropy and Cumulative Prospect Theory

    Directory of Open Access Journals (Sweden)

    Aihua Li

    2017-01-01

    Full Text Available Assessing the safety status of cranes is an important problem. To overcome the inaccuracies and misjudgments in such assessments, this work describes a safety assessment method for cranes that combines entropy and cumulative prospect theory. Firstly, the proposed method transforms the set of evaluation indices into an evaluation vector. Secondly, a decision matrix is then constructed from the evaluation vectors and evaluation standards, and an entropy-based technique is applied to calculate the index weights. Thirdly, positive and negative prospect value matrices are established from reference points based on the positive and negative ideal solutions. Thus, this enables the crane safety grade to be determined according to the ranked comprehensive prospect values. Finally, the safety status of four general overhead traveling crane samples is evaluated to verify the rationality and feasibility of the proposed method. The results demonstrate that the method described in this paper can precisely and reasonably reflect the safety status of a crane.

  7. An approach for risk informed safety culture assessment for Canadian nuclear power stations

    International Nuclear Information System (INIS)

    Nelson, W.R.

    2010-01-01

    One of the most important components of effective safety and risk management for nuclear power stations is a healthy safety culture. DNV has developed an approach for risk informed safety culture assessment that combines two complementary paradigms for safety and risk management: loss prevention - for preventing and intervening in accidents; and critical function management - for achieving safety and performance goals. Combining these two paradigms makes it possible to provide more robust systems for safety management and to support a healthy safety culture. This approach is being applied to safety culture assessment in partnership with a Canadian nuclear utility. (author)

  8. Safety assessment of existing highway bridges and viaducts

    NARCIS (Netherlands)

    Maljaars, J.; Steenbergen, R.; Abspoel, L.; Kolstein, M.H.

    2012-01-01

    The assessment of the structural safety of existing br idges and viaducts becomes increasingly important in many countries owing to an increase in traffic loads. Most existing standards, however, are developed for the design of new structures. For this reason, an assessment method for determining

  9. Assessment of the thermodynamic properties and phase diagram of the Bi-Pd system

    Czech Academy of Sciences Publication Activity Database

    Vřešťál, Jan; Pinkas, J.; Watson, A.; Scott, A.; Houserová, Jana; Kroupa, Aleš

    2006-01-01

    Roč. 30, č. 1 (2006), s. 14-17 ISSN 0364-5916 R&D Projects: GA MŠk OC 531.001; GA MŠk OC 531.002 Institutional research plan: CEZ:AV0Z20410507 Keywords : phase diagram * ab initio calculations * calorimetry Subject RIV: BJ - Thermodynamics Impact factor: 1.432, year: 2006

  10. Selected topics on the nonrelativistic diagram technique

    International Nuclear Information System (INIS)

    Blokhintsev, L.D.; Narodetskij, I.M.

    1983-01-01

    The construction of the diagrams describing various processes in the four-particle systems is considered. It is shown that these diagrams, in particular the diagrams corresponding to the simple mechanisms often used in nuclear and atomic reaction theory, are readily obtained from the Faddeev-Yakubovsky equations. The covariant four-dimensional formalism of nonrelativistic Feynman graphs and its connection to the three-dimensional graph technique are briefly discussed

  11. Drawing Euler Diagrams with Circles: The Theory of Piercings.

    Science.gov (United States)

    Stapleton, Gem; Leishi Zhang; Howse, John; Rodgers, Peter

    2011-07-01

    Euler diagrams are effective tools for visualizing set intersections. They have a large number of application areas ranging from statistical data analysis to software engineering. However, the automated generation of Euler diagrams has never been easy: given an abstract description of a required Euler diagram, it is computationally expensive to generate the diagram. Moreover, the generated diagrams represent sets by polygons, sometimes with quite irregular shapes that make the diagrams less comprehensible. In this paper, we address these two issues by developing the theory of piercings, where we define single piercing curves and double piercing curves. We prove that if a diagram can be built inductively by successively adding piercing curves under certain constraints, then it can be drawn with circles, which are more esthetically pleasing than arbitrary polygons. The theory of piercings is developed at the abstract level. In addition, we present a Java implementation that, given an inductively pierced abstract description, generates an Euler diagram consisting only of circles within polynomial time.

  12. Multi-currency Influence Diagrams

    DEFF Research Database (Denmark)

    Nielsen, Søren Holbech; Nielsen, Thomas Dyhre; Jensen, Finn V.

    2007-01-01

    When using the influence diagrams framework for solving a decision problem with several different quantitative utilities, the traditional approach has been to convert the utilities into one common currency. This conversion is carried out using a tacit transformation, under the assumption...... that the converted problem is equivalent to the original one. In this paper we present an extension of the influence diagram framework. The extension allows for these decision problems to be modelled in their original form. We present an algorithm that, given a linear conversion function between the currencies...

  13. Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

    International Nuclear Information System (INIS)

    Han, Sang Min; Seong, Poong Hyun

    2014-01-01

    This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness

  14. Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2014-08-15

    This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness.

  15. Overview of the ISAM safety assessment methodology

    International Nuclear Information System (INIS)

    Simeonov, G.

    2003-01-01

    The ISAM safety assessment methodology consists of the following key components: specification of the assessment context description of the disposal system development and justification of scenarios formulation and implementation of models running of computer codes and analysis and presentation of results. Common issues run through two or more of these assessment components, including: use of methodological and computer tools, collation and use of data, need to address various sources of uncertainty, building of confidence in the individual components, as well as the overall assessment. The importance of the iterative nature of the assessment should be recognised

  16. Building quality into performance and safety assessment software

    International Nuclear Information System (INIS)

    Wojciechowski, L.C.

    2011-01-01

    Quality assurance is integrated throughout the development lifecycle for performance and safety assessment software. The software used in the performance and safety assessment of a Canadian deep geological repository (DGR) follows the CSA quality assurance standard CSA-N286.7 [1], Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear Power Plants. Quality assurance activities in this standard include tasks such as verification and inspection; however, much more is involved in producing a quality software computer program. The types of errors found with different verification methods are described. The integrated quality process ensures that defects are found and corrected as early as possible. (author)

  17. Safety assessment input for site selection - the Swedish example - 59031

    International Nuclear Information System (INIS)

    Andersson, Johan

    2012-01-01

    Svensk Kaernbraenslehantering AB (SKB) has performed comprehensive investigations of two candidate sites for a final repository for Sweden's spent nuclear fuel. In March 2011 SKB decided to submit licence applications for a final repository at Forsmark. Before selection, SKB stated that the site that offers the best prospects for achieving long-term safety in practice would be selected. Based on experiences previous safety assessments, a number of issues related to long-term safety need to be considered in the context of site comparison. The factors include sensitivity to climate change such as periods of permafrost and glaciations, rock mechanics evolution including the potential for thermally induced spalling and sensitivity to potential future earthquakes, current and future groundwater flow, evolution of groundwater composition and proximity to mineral resources. Each of these factors related to long-term safety for the two candidate sites is assessed in a comparative analysis of site characteristics. The assessment also considers differences in biosphere conditions and in the confidence of the site descriptions. The comparison is concluded by an assessment on how the identified differences would affect the estimated radiological risk from a repository located at either of the sites. The assessment concludes that there are a number of safety related site characteristics for which the analyses do not show any decisive differences in terms of implications on safety, between the sites Forsmark and Laxemar. However, the frequency of water conducting fractures at repository depth is much smaller at Forsmark than at Laxemar. This difference, in turn, affects the future stability of the current favourable groundwater composition, which combined with the much higher flows at Laxemar would, for the current repository design, lead to a breach in the safety functions for the buffer and the canister for many more deposition positions at Laxemar than at Forsmark. Thereby

  18. Project SAFE. Update of the SFR-1 safety assessment. Phase 1

    International Nuclear Information System (INIS)

    Andersson, Johan; Riggare, P.; Skagius, K.

    1998-10-01

    SFR-1 is a facility for disposal of low-level radioactive operational waste from the nuclear power plants in Sweden. Low-level radioactive waste from industry, medicine, and research is also disposed in SFR-1. The facility is situated in bedrock beneath the Baltic Sea, 1 km off the coast near the Forsmark nuclear power plant. SFR-1 was built between the years 1983 and 1988. An assessment of the long-term performance of the facility was included in the vast documentation that was a part of the application for an operational license. The assessment was presented in the form of a final safety report. In the operational licence for SFR-1 it is stated that renewed safety assessments should be carried out at least each ten years. In order to meet this demand SKB has launched a special project, SAFE (Safety Assessment of Final Disposal of Operational Radioactive Waste). The aim of the project is to update the safety analysis and to prepare a safety report that will be presented to the Swedish authorities not later than year 2000. Project SAFE is divided into three phases. The first phase is a prestudy, and the results of the prestudy are given in this report. The aim of the prestudy is to identify issues where additional studies would improve the basis for the updated safety analysis as well as to suggest how these studies should be carried out. The work has been divided into six different topics, namely the inventory, the near field, the far field, the biosphere, radionuclide transport calculations and scenarios. For each topic the former safety reports and regulatory reviews are scrutinised and needs for additional work is identified. The evaluations are given in appendices covering the respective topics. The main report is a summary of the appendices with a more stringent description of the repository system and the processes that are of interest and therefore should be addressed in an updated safety assessment. However, it should be pointed out that one of the

  19. Human reliability analysis methods for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-11-01

    Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

  20. Covariant diagrams for one-loop matching

    International Nuclear Information System (INIS)

    Zhang, Zhengkang

    2016-10-01

    We present a diagrammatic formulation of recently-revived covariant functional approaches to one-loop matching from an ultraviolet (UV) theory to a low-energy effective field theory. Various terms following from a covariant derivative expansion (CDE) are represented by diagrams which, unlike conventional Feynman diagrams, involve gaugecovariant quantities and are thus dubbed ''covariant diagrams.'' The use of covariant diagrams helps organize and simplify one-loop matching calculations, which we illustrate with examples. Of particular interest is the derivation of UV model-independent universal results, which reduce matching calculations of specific UV models to applications of master formulas. We show how such derivation can be done in a more concise manner than the previous literature, and discuss how additional structures that are not directly captured by existing universal results, including mixed heavy-light loops, open covariant derivatives, and mixed statistics, can be easily accounted for.

  1. Covariant diagrams for one-loop matching

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zhengkang [Michigan Center for Theoretical Physics (MCTP), University of Michigan,450 Church Street, Ann Arbor, MI 48109 (United States); Deutsches Elektronen-Synchrotron (DESY),Notkestraße 85, 22607 Hamburg (Germany)

    2017-05-30

    We present a diagrammatic formulation of recently-revived covariant functional approaches to one-loop matching from an ultraviolet (UV) theory to a low-energy effective field theory. Various terms following from a covariant derivative expansion (CDE) are represented by diagrams which, unlike conventional Feynman diagrams, involve gauge-covariant quantities and are thus dubbed “covariant diagrams.” The use of covariant diagrams helps organize and simplify one-loop matching calculations, which we illustrate with examples. Of particular interest is the derivation of UV model-independent universal results, which reduce matching calculations of specific UV models to applications of master formulas. We show how such derivation can be done in a more concise manner than the previous literature, and discuss how additional structures that are not directly captured by existing universal results, including mixed heavy-light loops, open covariant derivatives, and mixed statistics, can be easily accounted for.

  2. Covariant diagrams for one-loop matching

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zhengkang [Michigan Univ., Ann Arbor, MI (United States). Michigan Center for Theoretical Physics; Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

    2016-10-15

    We present a diagrammatic formulation of recently-revived covariant functional approaches to one-loop matching from an ultraviolet (UV) theory to a low-energy effective field theory. Various terms following from a covariant derivative expansion (CDE) are represented by diagrams which, unlike conventional Feynman diagrams, involve gaugecovariant quantities and are thus dubbed ''covariant diagrams.'' The use of covariant diagrams helps organize and simplify one-loop matching calculations, which we illustrate with examples. Of particular interest is the derivation of UV model-independent universal results, which reduce matching calculations of specific UV models to applications of master formulas. We show how such derivation can be done in a more concise manner than the previous literature, and discuss how additional structures that are not directly captured by existing universal results, including mixed heavy-light loops, open covariant derivatives, and mixed statistics, can be easily accounted for.

  3. Covariant diagrams for one-loop matching

    International Nuclear Information System (INIS)

    Zhang, Zhengkang

    2017-01-01

    We present a diagrammatic formulation of recently-revived covariant functional approaches to one-loop matching from an ultraviolet (UV) theory to a low-energy effective field theory. Various terms following from a covariant derivative expansion (CDE) are represented by diagrams which, unlike conventional Feynman diagrams, involve gauge-covariant quantities and are thus dubbed “covariant diagrams.” The use of covariant diagrams helps organize and simplify one-loop matching calculations, which we illustrate with examples. Of particular interest is the derivation of UV model-independent universal results, which reduce matching calculations of specific UV models to applications of master formulas. We show how such derivation can be done in a more concise manner than the previous literature, and discuss how additional structures that are not directly captured by existing universal results, including mixed heavy-light loops, open covariant derivatives, and mixed statistics, can be easily accounted for.

  4. Students’ learning activities while studying biological process diagrams

    NARCIS (Netherlands)

    Kragten, M.; Admiraal, W.; Rijlaarsdam, G.

    2015-01-01

    Process diagrams describe how a system functions (e.g. photosynthesis) and are an important type of representation in Biology education. In the present study, we examined students’ learning activities while studying process diagrams, related to their resulting comprehension of these diagrams. Each

  5. Initialization of Safety Assessment Process for the Croatian Radioactive Waste repository on Trgovska gora

    International Nuclear Information System (INIS)

    Lokner, V.; Levanat, I.; Subasic, D.

    2000-01-01

    An iterative process of safety assessment, presently focusing on the site-specific evaluation of the post-closure phase for the prospective LILW repository on Trgovska gora in Croatia, has recently been initiated. The primary aim of the first assessment iterations is to provide the experts involved, the regulators and the general public with a reasonable assurance that the applicable long term performance and safety objectives can be met. Another goal is to develop a sufficient understanding of the system behavior to support decisions about the site investigation, the facility design, the waste acceptance criteria and the closure conditions. In this initial phase, the safety assessment is structured in a manner following closely methodology of the ISAM. The International Programme for Improving Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities the IAEA coordinated research program started in 1997. Results of the safety assessment first iteration will be organized and presented in the form of a preliminary safety analysis report (PSAR), expected to be completed in the second part of the year 2000. As the first report on the initiated safety assessment activities, the PSAR will describe the concept and aims of the assessment process. Particular emphasis will be placed on description of the key elements of a safety assessment approach by: a) defining the assessment context; b) providing description of the disposal system; c) developing and justifying assessment scenarios; d) formulating and implementing models; and e) interpreting the scoping calculations. (author)

  6. Development of the KINS Safety Culture Maturity Model for Self and Independent Assessment

    International Nuclear Information System (INIS)

    Sheen, C.; Choi, Y.S.

    2016-01-01

    Safety culture of an organization is cultivated and affected not only by societal and regulatory environment of the organization, but by its philosophies, policies, events and activities experienced in the process of accomplishing its mission. The safety culture would be continuously changed by the interactions between its members along with time as an organic entity. In order to perform a systematic self- or independent assessment of safety culture, a safety culture assessment model (SCAM) properly reflecting cultural characteristics should be necessary. In addition, a SCAM should be helpful not only to establish correct directions, goals, and strategies for safety culture development, but should anticipating obstacles against safety culture development in the implementation process derived from the assessment. In practical terms, a SCAM should be useful for deriving effective guidelines and implementing of corrective action programs for the evaluated organization. Korea Institute of Nuclear Safety (KINS) performed a research project for six years to develop a SCAM satisfying the above prerequisites for self- and independent assessment. The KINS SCAM was developed based on the five stage safety culture maturity model proposed by Professor Patrick Hudson and was modified into four stages to reflect existing safety culture assessment experiences at Korean nuclear power plants. In order to define the change mechanism of safety culture for development and reversion, the change model proposed by Prochaska and DiClemente was introduced into KINS SCAM and developed into the Spiral Change Model.

  7. Climate and climate-related issues for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Naeslund, Jens-Ove

    2006-11-01

    The purpose of this report is to document current scientific knowledge of the climate-related conditions and processes relevant to the long-term safety of a KBS-3 repository to a level required for an adequate treatment in the safety assessment SR-Can. The report also includes a concise background description of the climate system. The report includes three main chapters: A description of the climate system (Chapter 2); Identification and discussion of climate-related issues (Chapter 3); and, A description of the evolution of climate-related conditions for the safety assessment (Chapter 4). Chapter 2 includes an overview of present knowledge of the Earth climate system and the climate conditions that can be expected to occur in Sweden on a 100,000 year time perspective. Based on this, climate-related issues relevant for the long-term safety of a KBS-3 repository are identified. These are documented in Chapter 3 'Climate-related issues' to a level required for an adequate treatment in the safety assessment. Finally, in Chapter 4, 'Evolution of climate-related conditions for the safety assessment' an evolution for a 120,000 year period is presented, including discussions of identified climate-related issues of importance for repository safety. The documentation is from a scientific point of view not exhaustive, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of a safety assessment. As further described in the SR-Can Main Report and in the Features Events and Processes report, the content of the present report has been audited by comparison with FEP databases compiled in other assessment projects. This report follows as far as possible the template for documentation of processes regarded as internal to the repository system. However, the term processes is not used in this report, instead the term issue has been used. Each issue includes a set of processes together resulting in the behaviour of a

  8. Climate and climate-related issues for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Naeslund, Jens-Ove (comp.)

    2006-11-15

    The purpose of this report is to document current scientific knowledge of the climate-related conditions and processes relevant to the long-term safety of a KBS-3 repository to a level required for an adequate treatment in the safety assessment SR-Can. The report also includes a concise background description of the climate system. The report includes three main chapters: A description of the climate system (Chapter 2); Identification and discussion of climate-related issues (Chapter 3); and, A description of the evolution of climate-related conditions for the safety assessment (Chapter 4). Chapter 2 includes an overview of present knowledge of the Earth climate system and the climate conditions that can be expected to occur in Sweden on a 100,000 year time perspective. Based on this, climate-related issues relevant for the long-term safety of a KBS-3 repository are identified. These are documented in Chapter 3 'Climate-related issues' to a level required for an adequate treatment in the safety assessment. Finally, in Chapter 4, 'Evolution of climate-related conditions for the safety assessment' an evolution for a 120,000 year period is presented, including discussions of identified climate-related issues of importance for repository safety. The documentation is from a scientific point of view not exhaustive, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of a safety assessment. As further described in the SR-Can Main Report and in the Features Events and Processes report, the content of the present report has been audited by comparison with FEP databases compiled in other assessment projects. This report follows as far as possible the template for documentation of processes regarded as internal to the repository system. However, the term processes is not used in this report, instead the term issue has been used. Each issue includes a set of processes together resulting in the

  9. Integrated safety assessment report, Haddam Neck Plant (Docket No. 50-213): Integrated Safety Assessment Program: Draft report

    International Nuclear Information System (INIS)

    1987-07-01

    The integrated assessment is conducted on 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. Procedures allow for a periodic updating of the schedules to account for licensing issues that arise in the future. The Haddam Neck Plant is one of two plants being reviewed under the pilot program. This report indicates how 82 topics selected for review were addressed, and presents the staff's recommendations regarding the corrective actions to resolve the 82 topics and other actions to enhance plant safety. 135 refs., 4 figs., 5 tabs

  10. Safety Assessment Methodologies and Their Application in Development of Near Surface Waste Disposal Facilities--ASAM Project

    International Nuclear Information System (INIS)

    Batandjieva, B.; Metcalf, P.

    2003-01-01

    Safety of near surface disposal facilities is a primary focus and objective of stakeholders involved in radioactive waste management of low and intermediate level waste and safety assessment is an important tool contributing to the evaluation and demonstration of the overall safety of these facilities. It plays significant role in different stages of development of these facilities (site characterization, design, operation, closure) and especially for those facilities for which safety assessment has not been performed or safety has not been demonstrated yet and the future has not been decided. Safety assessments also create the basis for the safety arguments presented to nuclear regulators, public and other interested parties in respect of the safety of existing facilities, the measures to upgrade existing facilities and development of new facilities. The International Atomic Energy Agency (IAEA) has initiated a number of research coordinated projects in the field of development and improvement of approaches to safety assessment and methodologies for safety assessment of near surface disposal facilities, such as NSARS (Near Surface Radioactive Waste Disposal Safety Assessment Reliability Study) and ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) projects. These projects were very successful and showed that there is a need to promote the consistent application of the safety assessment methodologies and to explore approaches to regulatory review of safety assessments and safety cases in order to make safety related decisions. These objectives have been the basis of the IAEA follow up coordinated research project--ASAM (Application of Safety Assessment Methodologies for Near Surface Disposal Facilities), which will commence in November 2002 and continue for a period of three years

  11. Operations space diagram for ECRH and ECCD

    International Nuclear Information System (INIS)

    Bindslev, Henrik

    2004-01-01

    A Clemmov-Mullaly-Allis (CMA) type diagram, the ECW-CMA diagram, for representing the operational possibilities of electron cyclotron heating and current drive (ECRH/ECCD) systems for fusion plasmas is presented. In this diagram, with normalized density and normalized magnetic field coordinates, the parameter range in which it is possible to achieve a given task (e.g. O-mode current drive for stabilizing a neoclassical tearing mode) appears as a region. With also the Greenwald density limit shown, this diagram condenses the information on operational possibilities, facilitating the overview required at the design phase. At the operations phase it may also prove useful in setting up experimental scenarios by showing operational possibilities, avoiding the need for survey type ray-tracing at the initial planning stages. The diagram may also serve the purpose of communicating operational possibilities to non-experts. JET and ITER like plasmas are used, but the method is generic. (author)

  12. Operational safety assessment of underground test facilities for mined geologic waste disposal

    International Nuclear Information System (INIS)

    Elder, H.K.

    1993-01-01

    This paper describes the operational safety assessment for the underground facilities for the exploratory studies facility (ESF) at the Yucca Mountain Project. The systematic identification and evaluation of hazards related to the ESF is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach based on the analysis of potential accidents was used since radiological safety analysis was not required. The risk assessment summarized credible accident scenarios and the design provides mitigation of the risks to a level that the facility can be constructed and operated with an adequate level of safety. The risk assessment also provides reasonable assurance that all identifiable major accident scenarios have been reviewed and design mitigation features provided to ensure an adequate level of safety

  13. Radionuclide transport report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    This document compiles radionuclide transport calculations of a KBS-3 repository for the safety assessment SR-Site. The SR-Site assessment supports the licence application for a final repository at Forsmark, Sweden

  14. Near threshold expansion of Feynman diagrams

    International Nuclear Information System (INIS)

    Mendels, E.

    2005-01-01

    The near threshold expansion of Feynman diagrams is derived from their configuration space representation, by performing all x integrations. The general scalar Feynman diagram is considered, with an arbitrary number of external momenta, an arbitrary number of internal lines and an arbitrary number of loops, in n dimensions and all masses may be different. The expansions are considered both below and above threshold. Rules, giving real and imaginary part, are derived. Unitarity of a sunset diagram with I internal lines is checked in a direct way by showing that its imaginary part is equal to the phase space integral of I particles

  15. The probability safety assessment impact on the BR2 refurbishment

    International Nuclear Information System (INIS)

    Pouleur, Yvan

    1995-01-01

    The probabilistic safety assessment (PSA) study has proven its worth by establishing a sensitive safety screening of the reactor. It has focused engineering forces to technically improve safety systems and to measure the influence of functional modifications. In the future, the project will be developed in a living way, to reinforce the present structure along with continuous safety monitoring of the reactor and to develop engineers and operators safety skills. This paper presents the PSA impact on the BR2 (Belgian Reactor Two) refurbishment. (author)

  16. 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

  17. Safety assessment of geologic repositories for nuclear waste

    International Nuclear Information System (INIS)

    Bartlett, J.W.; Burkholder, H.C.; Winegardner, W.K.

    1977-01-01

    Consideration of geologic isolation for final disposition of radioactive wastes has led to the need for evaluation of the safety of the concept. Such evaluations require consideration of factors not encountered in conventional risk analysis: consequences at times and places far removed from the repository site; indirect, complex, and alternative pathways between the waste and the point of potential consequences; a highly limited data base; and limited opportunity for experimental verification of results. R and D programs to provide technical safety evaluations are under way. Three methods are being considered for the probabilistic aspects of the evaluations: fault tree analysis, repository simulation analysis, and system stability analysis. Nuclide transport models, currently in a relatively advanced state of development, are used to evaluate consequences of postulated loss of geologic isolation. This paper outlines the safety assessment methods, unique features of the assessment problem that affect selection of methods and reliability of results, and available results. It also discusses potential directions for future work

  18. Influence Diagram Use With Respect to Technology Planning and Investment

    Science.gov (United States)

    Levack, Daniel J. H.; DeHoff, Bryan; Rhodes, Russel E.

    2009-01-01

    Influence diagrams are relatively simple, but powerful, tools for assessing the impact of choices or resource allocations on goals or requirements. They are very general and can be used on a wide range of problems. They can be used for any problem that has defined goals, a set of factors that influence the goals or the other factors, and a set of inputs. Influence diagrams show the relationship among a set of results and the attributes that influence them and the inputs that influence the attributes. If the results are goals or requirements of a program, then the influence diagram can be used to examine how the requirements are affected by changes to technology investment. This paper uses an example to show how to construct and interpret influence diagrams, how to assign weights to the inputs and attributes, how to assign weights to the transfer functions (influences), and how to calculate the resulting influences of the inputs on the results. A study is also presented as an example of how using influence diagrams can help in technology planning and investment. The Space Propulsion Synergy Team (SPST) used this technique to examine the impact of R&D spending on the Life Cycle Cost (LCC) of a space transportation system. The question addressed was the effect on the recurring and the non-recurring portions of LCC of the proportion of R&D resources spent to impact technology objectives versus the proportion spent to impact operational dependability objectives. The goals, attributes, and the inputs were established. All of the linkages (influences) were determined. The weighting of each of the attributes and each of the linkages was determined. Finally the inputs were varied and the impacts on the LCC determined and are presented. The paper discusses how each of these was accomplished both for credibility and as an example for future studies using influence diagrams for technology planning and investment planning.

  19. Safety Assessment in the AREVA Group: Operating Experience from a Self-Assessment Tool

    International Nuclear Information System (INIS)

    Coye de Brunélis, T.; Mignot, E.; Sidaner, J.-F.

    2016-01-01

    The expression “safety culture” first appeared following analysis of the Chernobyl accident in 1986. It was first defined in INSAG-4 (International Nuclear Safety Advisory Group safety series) in 1991. Other events have occurred in nuclear facilities and during transportation since Chernobyl: Tokai Mura in 1999, Roissy Transport in 2002, Davis Besse in 2002, Thorp in 2005. These events show that the initial approach was too simplistic. Based on this observation, the definition of safety culture was supplemented by including concepts of cultural value (associated with the country and the company) and human and organizational factors, and was integrated in that form with the emergence and implementation of integrated management systems (IMS). Today, the concept of nuclear safety culture covers a wide set of factors such as safety, quality, corporate culture, defined processes and policies, organizations and related resources. Any assessment of people’s safety culture, particularly people directly involved in facility operations, is thus part of a comprehensive policy and contributes to a de facto demonstration of the priority which management assigns to safety.

  20. Environment, Safety, and Health Self-Assessment Report, Fiscal Year 2008

    Energy Technology Data Exchange (ETDEWEB)

    Chernowski, John

    2009-02-27

    Lawrence Berkeley National Laboratory's Environment, Safety, and Health (ES&H) Self-Assessment Program ensures that Integrated Safety Management (ISM) is implemented institutionally and by all divisions. The Self-Assessment Program, managed by the Office of Contract Assurance (OCA), provides for an internal evaluation of all ES&H programs and systems at LBNL. The functions of the program are to ensure that work is conducted safely, and with minimal negative impact to workers, the public, and the environment. The Self-Assessment Program is also the mechanism used to institute continuous improvements to the Laboratory's ES&H programs. The program is described in LBNL/PUB 5344, Environment, Safety, and Health Self-Assessment Program and is composed of four distinct assessments: the Division Self-Assessment, the Management of Environment, Safety, and Health (MESH) review, ES&H Technical Assurance, and the Appendix B Self-Assessment. The Division Self-Assessment uses the five core functions and seven guiding principles of ISM as the basis of evaluation. Metrics are created to measure performance in fulfilling ISM core functions and guiding principles, as well as promoting compliance with applicable regulations. The five core functions of ISM are as follows: (1) Define the Scope of Work; (2) Identify and Analyze Hazards; (3) Control the Hazards; (4) Perform the Work; and (5) Feedback and Improvement. The seven guiding principles of ISM are as follows: (1) Line Management Responsibility for ES&H; (2) Clear Roles and Responsibilities; (3) Competence Commensurate with Responsibilities; (4) Balanced Priorities; (5) Identification of ES&H Standards and Requirements; (6) Hazard Controls Tailored to the Work Performed; and (7) Operations Authorization. Performance indicators are developed by consensus with OCA, representatives from each division, and Environment, Health, and Safety (EH&S) Division program managers. Line management of each division performs the

  1. On the question of calculation methods of phase diagrams

    International Nuclear Information System (INIS)

    Vasil'ev, M.V.

    1983-01-01

    The technique of determining interaction parameters of components of binary alloys is suggested. U-Mo and Cu-Al systems are used as example with the aid of experimental state diagrams. It is shown that the search for new regularities is necessary with the aim of analytical description of state diagrams and forecast of the shape of phase equilibria curves in real systems. Optimum combinations of experimental investigations with the aim of reliable determination of supporting points and forecasting possibilities of typical equations can considerably decrease the volume of experimental work when preparing state diagrams, in cases of repeated state diagrams of more reliable state diagrams with the application of more advanced methods of investigation. The translation of state diagrams from geometric to analytical language with the use of typical equations opens up new possibilities for establishing a compact information bank for state diagrams

  2. A geometric proof of confluence by decreasing diagrams

    NARCIS (Netherlands)

    Klop, J.W.; Oostrom, V. van; Vrijer, R. de

    The criterion for confluence using decreasing diagrams is a generalization of several well-known confluence criteria in abstract rewriting, such as the strong confluence lemma. We give a new proof of the decreasing diagram theorem based on a geometric study of in finite reduction diagrams, arising

  3. Safety assessment of envisaged systems for automotive hydrogen supply and utilization

    Energy Technology Data Exchange (ETDEWEB)

    Landucci, Gabriele [Dipartimento di Ingegneria Chimica, Chimica Industriale e Scienza dei Materiali, Universita di Pisa, via Diotisalvi n.2, 56126 Pisa (Italy); Tugnoli, Alessandro; Cozzani, Valerio [Dipartimento di Ingegneria Chimica, Mineraria e delle Tecnologie Ambientali, Alma Mater Studiorum - Universita di Bologna, via Terracini n.28, 40131 Bologna (Italy)

    2010-02-15

    A novel consequence-based approach was applied to the inherent safety assessment of the envisaged hydrogen production, distribution and utilization systems, in the perspective of the widespread hydrogen utilization as a vehicle fuel. Alternative scenarios were assessed for the hydrogen system chain from large scale production to final utilization. Hydrogen transportation and delivery was included in the analysis. The inherent safety fingerprint of each system was quantified by a set of Key Performance Indicators (KPIs). Rules for KPIs aggregation were considered for the overall assessment of the system chains. The final utilization stage resulted by large the more important for the overall expected safety performance of the system. Thus, comparison was carried out with technologies proposed for the use of other low emission fuels, as LPG and natural gas. The hazards of compressed hydrogen-fueled vehicles resulted comparable, while reference innovative hydrogen technologies evidenced a potentially higher safety performance. Thus, switching to the inherently safer technologies currently under development may play an important role in the safety enhancement of hydrogen vehicles, resulting in a relevant improvement of the overall safety performance of the entire hydrogen system. (author)

  4. Compact flow diagrams for state sequences

    NARCIS (Netherlands)

    Buchin, K.A.; Buchin, M.E.; Gudmundsson, J.; Horton, M.J.; Sijben, S.

    2016-01-01

    We introduce the concept of compactly representing a large number of state sequences, e.g., sequences of activities, as a flow diagram. We argue that the flow diagram representation gives an intuitive summary that allows the user to detect patterns among large sets of state sequences. Simplified,

  5. Safety cases for the co-ordinated research project on improvement of safety assessment methodologies for near surface radioactive waste disposal facilities (ISAM)

    International Nuclear Information System (INIS)

    Kozak, M.W.; Torres-Vidal, C.; Kelly, E.; Guskov, A.; Blerk, J. van

    2002-01-01

    A Co-ordinated Research Project (CRP) has recently been completed on the Improvement of Safety Assessment Methodologies for Near-Surface Radioactive Waste Disposal Facilities (ISAM). A major aspect of the project was the use of safety cases for the practical application of safety assessment. An overview of the ISAM safety cases is given in this paper. (author)

  6. Voronoi diagram and microstructure of weldment

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jung Ho [Chungbuk National University, Cheongju (Korea, Republic of)

    2015-01-15

    Voronoi diagram, one of the well-known space decomposition algorithms has been applied to express the microstructure of a weldment for the first time due to the superficial analogy between a Voronoi cell and a metal's grain. The area of the Voronoi cells can be controlled by location and the number of the seed points. This can be correlated to the grain size in the microstructure and the number of nuclei formed. The feasibility of representing coarse and fine grain structures were tested through Voronoi diagrams and it is applied to expression of cross-sectional bead shape of a typical laser welding. As result, it successfully described coarsened grain size of heat affected zone and columnar crystals in fusion zone. Although Voronoi diagram showed potential as a microstructure prediction tool through this feasible trial but direct correlation control variable of Voronoi diagram to solidification process parameter is still remained as further works.

  7. Confidence building in safety assessment

    International Nuclear Information System (INIS)

    Osthols, E.

    1999-01-01

    Engineered disposal systems are necessary to isolate radioactive waste from humans and the environment. It is essential to have access to basic thermochemical data relevant to varying geological environments for the radioactive elements involved. The OECD/NEA Thermochemical Data Base project (TDB) aims to make widely available basic thermochemical data of the type needed for safety assessment of nuclear storage facilities. The history and the present status of the project are presented. (K.A.)

  8. Interim main report of the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Hedin, Allan [and others

    2004-08-01

    This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8

  9. Interim main report of the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Hedin, Allan

    2004-08-01

    This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10 -6 for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8. Hydrogeological

  10. Safety assessment for the underground disposal of radioactive wastes

    International Nuclear Information System (INIS)

    1981-01-01

    This document is addressed to authorities and specialists responsible for or involved in planning, performing and reviewing safety assessments of underground radioactive waste repositories. It introduces and discusses in a general manner approaches and areas to be considered in making such safety assessments; its emphasis is on repositories for long-lived radioactive wastes in deep geological formations. It is hoped that this document will contribute to providing a base for a common understanding among the authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing guidance, the document is also intended to stimulate further international discussion on this subject. It is the intention of the IAEA to develop more specific reports providing examples for the application of safety analyses for underground waste disposal

  11. Safety assessment for the underground disposal of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    This document is addressed to authorities and specialists responsible for or involved in planning, performing and reviewing safety assessments of underground radioactive waste repositories. It introduces and discusses in a general manner approaches and areas to be considered in making such safety assessments; its emphasis is on repositories for long-lived radioactive wastes in deep geological formations. It is hoped that this document will contribute to providing a base for a common understanding among the authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing guidance, the document is also intended to stimulate further international discussion on this subject. It is the intention of the IAEA to develop more specific reports providing examples for the application of safety analyses for underground waste disposal.

  12. NANoREG framework for the safety assessment of nanomaterials

    OpenAIRE

    Gottardo, Stefania; Alessandrelli, Maria; Amenta, Valeria; Atluri, Rambabu; Barberio, Grazia; Bekker, Cindy; Bergonzo, Philippe; Bleeker, Eric; Booth, Andy; Borges, Teresa; Buttol, Patrizia; Carlander, David; Castelli, Stefano; Chevillard, Sylvie

    2017-01-01

    The NANoREG framework addresses the need to ease the nanomaterials safety assessment in the REACH Regulation context. It offers forward-looking strategies: Safe-by-Design, a Nanospecific Prioritisation and Risk Assessment, and Life Cycle Assessment. It is intended for scientific experts, regulatory authorities and industry.

  13. The Eh-pH Diagram and Its Advances

    Directory of Open Access Journals (Sweden)

    Hsin-Hsiung Huang

    2016-01-01

    Full Text Available Since Pourbaix presented Eh versus pH diagrams in his “Atlas of Electrochemical Equilibria in Aqueous Solution”, diagrams have become extremely popular and are now used in almost every scientific area related to aqueous chemistry. Due to advances in personal computers, such diagrams can now show effects not only of Eh and pH, but also of variables, including ligand(s, temperature and pressure. Examples from various fields are illustrated in this paper. Examples include geochemical formation, corrosion and passivation, precipitation and adsorption for water treatment and leaching and metal recovery for hydrometallurgy. Two basic methods were developed to construct an Eh-pH diagram concerning the ligand component(s. The first method calculates and draws a line between two adjacent species based on their given activities. The second method performs equilibrium calculations over an array of points (500 × 800 or higher are preferred, each representing one Eh and one pH value for the whole system, then combines areas of each dominant species for the diagram. These two methods may produce different diagrams. The fundamental theories, illustrated results, comparison and required conditions behind these two methods are presented and discussed in this paper. The Gibbs phase rule equation for an Eh-pH diagram was derived and verified from actual plots. Besides indicating the stability area of water, an Eh-pH diagram normally shows only half of an overall reaction. However, merging two or more related diagrams together reveals more clearly the possibility of the reactions involved. For instance, leaching of Au with cyanide followed by cementing Au with Zn (Merrill-Crowe process can be illustrated by combining Au-CN and Zn-CN diagrams together. A second example of the galvanic conversion of chalcopyrite can be explained by merging S, Fe–S and Cu–Fe–S diagrams. The calculation of an Eh-pH diagram can be extended easily into another dimension, such

  14. Compact flow diagrams for state sequences

    NARCIS (Netherlands)

    Buchin, Kevin; Buchin, Maike; Gudmundsson, Joachim; Horton, Michael; Sijben, Stef

    2017-01-01

    We introduce the concept of using a flow diagram to compactly represent the segmentation of a large number of state sequences according to a set of criteria. We argue that this flow diagram representation gives an intuitive summary that allows the user to detect patterns within the segmentations. In

  15. Diagrams of natural deductions

    Energy Technology Data Exchange (ETDEWEB)

    Popov, S V

    1982-01-01

    The concept of natural deductions was investigated by the author in his analysis of the complexity of deductions in propositional computations (1975). Here some natural deduction systems are considered, and an analytical procedure proposed which results in a deduction diagram for each system. Each diagram takes the form of an orientated, charge graph, features of which can be used to establish the equivalence of classes of deductions. For each of the natural deduction systems considered, a system of equivalent transformation schemes is derived, which is complete with respect to the given definition of equivalence. 2 references.

  16. Management of safety, safety culture and self assessment

    International Nuclear Information System (INIS)

    Carnino, A.

    2000-01-01

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

  17. Ex-ante assessment of the safety effects of intelligent transport systems.

    Science.gov (United States)

    Kulmala, Risto

    2010-07-01

    There is a need to develop a comprehensive framework for the safety assessment of Intelligent Transport Systems (ITS). This framework should: (1) cover all three dimensions of road safety-exposure, crash risk and consequence, (2) cover, in addition to the engineering effect, also the effects due to behavioural adaptation and (3) be compatible with the other aspects of state of the art road safety theories. A framework based on nine ITS safety mechanisms is proposed and discussed with regard to the requirements set to the framework. In order to illustrate the application of the framework in practice, the paper presents a method based on the framework and the results from applying that method for twelve intelligent vehicle systems in Europe. The framework is also compared to two recent frameworks applied in the safety assessment of intelligent vehicle safety systems. Copyright 2010 Elsevier Ltd. All rights reserved.

  18. Re-assessment of seismic loads in conjunction with periodic safety review

    International Nuclear Information System (INIS)

    Jonczyk, Josef

    2002-01-01

    The objective of this paper is the fundamental consideration of a safeguard-aim-oriented approach for use in the re-assessment of seismic events with regard to the periodic safety review (PSR) of nuclear power plants (NPP). The re-assessment aspects of site-specific design earthquakes (DEQ), specially the procedure for seismic hazard analysis, will not, however, be considered in detail here. The proposed assessment concept clearly presents a general approach for safety assessments. The approach is based on a successive screening review of components that are considered sufficiently earthquake-resistant. In this respect, the principle of maximum practical application of the design documentation has been considered in the re-assessment process. On the other hand, the safeguard-aim-oriented evaluation will also be applied with regard to whether the requirements of the safety regulations are fulfilled with respect to the safety goals. The review in conjunction with PSR does not, however, attempt to perform this under all technical aspects. Moreover, it is possible to make extensive use of experimental knowledge and engineering judgement with regard to the structural capacity behaviour in case of a seismic event. Compared with design procedures, however, this proposed approach differs from the one applied in licensing procedures, in which such assessment freedom will not usually be exhausted. (author)

  19. Control wiring diagrams

    International Nuclear Information System (INIS)

    McCauley, T.M.; Eskinazi, M.; Henson, L.L.

    1989-01-01

    This paper discusses the changes in electrical document requirements that occur when construction is complete and a generating station starts commercial operation. The needs of operations and maintenance (O and M) personnel are analyzed and contrasted with those of construction to illustrate areas in which the construction documents (drawings, diagrams, and databases) are difficult to use for work at an operating station. The paper discusses the O and M electrical documents that the Arizona Nuclear Power Project (ANPP) believes are most beneficial for the three operating units at Palo Verde; these are control wiring diagrams and an associated document cross-reference list. The benefits offered by these new, station O and M-oriented documents are weighted against the cost of their creation and their impact on drawing maintenance

  20. Probabilistic safety assessment based expert systems in support of dynamic risk assessment

    International Nuclear Information System (INIS)

    Varde, P.V.; Sharma, U.L.; Marik, S.K.; Raina, V.K.; Tikku, A.C.

    2006-01-01

    Probabilistic Safety Assessment (PSA) studies are being performed, world over as part of integrated risk assessment for Nuclear Power Plants and in many cases PSA insight is utilized in support of decision making. Though the modern plants are built with inherent safety provisions, particularly to reduce the supervisory requirements during initial period into the accident, it is always desired to develop an efficient user friendly real-time operator advisory system for handling of plant transients/emergencies which would be of immense benefit for the enhancement of operational safety of the plant. This paper discusses an integrated approach for the development of operator support system. In this approach, PSA methodology and the insight obtained from PSA has been utilized for development of knowledge based or rule based experts system. While Artificial Neural Network (ANN) approach has been employed for transient identification, rule-base expert system shell environment was used for the development of diagnostic module in this system. Attempt has been made to demonstrate that this approach offers an efficient framework for addressing requirements related to handling of real-time/dynamic scenario. (author)

  1. New detectors for powders diagrams

    International Nuclear Information System (INIS)

    Convert, P.

    1975-01-01

    During the last few years, all the classical neutron diffractometers for powders have used one or maybe a few counters. So, it takes a long time to obtain a diagram which causes many disadvantages: 1) very long experiments: one or two days (or flux on the sample about 10 6 n/cm 2 /a); 2) necessity of big samples: many cm 3 ; 3) necessity of having the whole diagram before changing anything in the experiment: magnetic field, temperature, quality of the sample; 4) necessity of having collimators of a few times ten minutes to obtain correct statistics in the diagram. Because of these disadvantages, several attempts have been made to speed up the experimental procedure such as using more counters, the detection of neutrons on a resistive wire, etc. In Grenoble, new position-sensitive detectors have been constructed using a digital technique

  2. Hybrid approach for the assessment of PSA models by means of binary decision diagrams

    International Nuclear Information System (INIS)

    Ibanez-Llano, Cristina; Rauzy, Antoine; Melendez, Enrique; Nieto, Francisco

    2010-01-01

    Binary decision diagrams are a well-known alternative to the minimal cutsets approach to assess the reliability Boolean models. They have been applied successfully to improve the fault trees models assessment. However, its application to solve large models, and in particular the event trees coming from the PSA studies of the nuclear industry, remains to date out of reach of an exact evaluation. For many real PSA models it may be not possible to compute the BDD within reasonable amount of time and memory without considering the truncation or simplification of the model. This paper presents a new approach to estimate the exact probabilistic quantification results (probability/frequency) based on combining the calculation of the MCS and the truncation limits, with the BDD approach, in order to have a better control on the reduction of the model and to properly account for the success branches. The added value of this methodology is that it is possible to ensure a real confidence interval of the exact value and therefore an explicit knowledge of the error bound. Moreover, it can be used to measure the acceptability of the results obtained with traditional techniques. The new method was applied to a real life PSA study and the results obtained confirm the applicability of the methodology and open a new viewpoint for further developments.

  3. Hybrid approach for the assessment of PSA models by means of binary decision diagrams

    Energy Technology Data Exchange (ETDEWEB)

    Ibanez-Llano, Cristina, E-mail: cristina.ibanez@iit.upcomillas.e [Instituto de Investigacion Tecnologica (IIT), Escuela Tecnica Superior de Ingenieria ICAI, Universidad Pontificia Comillas, C/Santa Cruz de Marcenado 26, 28015 Madrid (Spain); Rauzy, Antoine, E-mail: Antoine.RAUZY@3ds.co [Dassault Systemes, 10 rue Marcel Dassault CS 40501, 78946 Velizy Villacoublay Cedex (France); Melendez, Enrique, E-mail: ema@csn.e [Consejo de Seguridad Nuclear (CSN), C/Justo Dorado 11, 28040 Madrid (Spain); Nieto, Francisco, E-mail: nieto@iit.upcomillas.e [Instituto de Investigacion Tecnologica (IIT), Escuela Tecnica Superior de Ingenieria ICAI, Universidad Pontificia Comillas, C/Santa Cruz de Marcenado 26, 28015 Madrid (Spain)

    2010-10-15

    Binary decision diagrams are a well-known alternative to the minimal cutsets approach to assess the reliability Boolean models. They have been applied successfully to improve the fault trees models assessment. However, its application to solve large models, and in particular the event trees coming from the PSA studies of the nuclear industry, remains to date out of reach of an exact evaluation. For many real PSA models it may be not possible to compute the BDD within reasonable amount of time and memory without considering the truncation or simplification of the model. This paper presents a new approach to estimate the exact probabilistic quantification results (probability/frequency) based on combining the calculation of the MCS and the truncation limits, with the BDD approach, in order to have a better control on the reduction of the model and to properly account for the success branches. The added value of this methodology is that it is possible to ensure a real confidence interval of the exact value and therefore an explicit knowledge of the error bound. Moreover, it can be used to measure the acceptability of the results obtained with traditional techniques. The new method was applied to a real life PSA study and the results obtained confirm the applicability of the methodology and open a new viewpoint for further developments.

  4. Innovative Modelling Approach of Safety Culture Assessment in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ahn, N.

    2016-01-01

    A culture is commonly defined as the shared set of norms and values that govern appropriate individual behavior. Safety culture is the subset of organizational culture that reflects the general attitude and approaches to safety and risk management. While safety is sometimes narrowly defined in terms of human death and injury, we use a more inclusive definition that also considers mission loss as a safety problem and is thus applicable to nuclear power plants and missions. The recent accident reports and investigations of the nuclear power plant mission failures (i.e., TMI, Chernobyl, and Fukushima) point to safety cultural problems in nuclear power plants. Many assessment approaches have been developed by organizations such as IAEA and INPO based on the assessment of parameters at separate levels — individuals, groups, and organizations.

  5. Self assessment of safety culture in HANARO using the code of conduct on the safety of research reactor by IAEA

    International Nuclear Information System (INIS)

    Lim, I.C.; Hwang, S.Y.; Woo, J.S.; Lee, M.; Jun, B.J.

    2003-01-01

    Full text: The safety culture in HANARO was self-assessed in accordance with the Code of Conduct on the Safety of Research Reactor drafted by IAEA. From 2002, IAEA has worked on the development of the Code of Conduct to achieve and maintain high level of nuclear safety in research reactors worldwide through the enhancement of national measures and international co-operation including, where appropriate, safety related technical cooperation. It defines the role of the state, the role of the regulatory body, the role of the operating organization and the role of the IAEA. As for the role of operating organization, the code specifies general requirements in assessment and verification of safety, financial and human resources, quality assurance, human factors, radiation protection and emergency preparedness. It also defines the role of operating organization for safety of research reactor in siting, design, operation, maintenance, modification and utilization as well. All of these items are the subjects for safety culture implementation, which means the Code could be a guideline for an operating organization to assess its safety culture. The self-assessment of safety culture in HANARO was made by using the sections of the Code describing the role of the operating organization for safety of research reactor. The major assessment items and the practices in HANARO for each items are as follow: The SAR of HANARO was reviewed by the regulatory body before the construction and the fuel loading of HANARO. Major design modifications and new installation of utilization facility needs the approval from regulatory body and safety assessment is a requirement for the approval. The Tech. Spec. for HANARO Operation specifies the analysis, surveillance, testing and inspection for HANARO operation. The reactor operation is mainly supported by the government and partly by nuclear R and D fund. The education and training of operation staff are one of major tasks of operating organization

  6. Assessment of the safety of foods derived from genetically modified (GM) crops.

    Science.gov (United States)

    König, A; Cockburn, A; Crevel, R W R; Debruyne, E; Grafstroem, R; Hammerling, U; Kimber, I; Knudsen, I; Kuiper, H A; Peijnenburg, A A C M; Penninks, A H; Poulsen, M; Schauzu, M; Wal, J M

    2004-07-01

    This paper provides guidance on how to assess the safety of foods derived from genetically modified crops (GM crops); it summarises conclusions and recommendations of Working Group 1 of the ENTRANSFOOD project. The paper provides an approach for adapting the test strategy to the characteristics of the modified crop and the introduced trait, and assessing potential unintended effects from the genetic modification. The proposed approach to safety assessment starts with the comparison of the new GM crop with a traditional counterpart that is generally accepted as safe based on a history of human food use (the concept of substantial equivalence). This case-focused approach ensures that foods derived from GM crops that have passed this extensive test-regime are as safe and nutritious as currently consumed plant-derived foods. The approach is suitable for current and future GM crops with more complex modifications. First, the paper reviews test methods developed for the risk assessment of chemicals, including food additives and pesticides, discussing which of these methods are suitable for the assessment of recombinant proteins and whole foods. Second, the paper presents a systematic approach to combine test methods for the safety assessment of foods derived from a specific GM crop. Third, the paper provides an overview on developments in this area that may prove of use in the safety assessment of GM crops, and recommendations for research priorities. It is concluded that the combination of existing test methods provides a sound test-regime to assess the safety of GM crops. Advances in our understanding of molecular biology, biochemistry, and nutrition may in future allow further improvement of test methods that will over time render the safety assessment of foods even more effective and informative. Copryright 2004 Elsevier Ltd.

  7. Global Water Cycle Diagrams Minimize Human Influence and Over-represent Water Security

    Science.gov (United States)

    Abbott, B. W.; Bishop, K.; Zarnetske, J. P.; Minaudo, C.; Chapin, F. S., III; Plont, S.; Marçais, J.; Ellison, D.; Roy Chowdhury, S.; Kolbe, T.; Ursache, O.; Hampton, T. B.; GU, S.; Chapin, M.; Krause, S.; Henderson, K. D.; Hannah, D. M.; Pinay, G.

    2017-12-01

    The diagram of the global water cycle is the central icon of hydrology, and for many people, the point of entry to thinking about key scientific concepts such as conservation of mass, teleconnections, and human dependence on ecological systems. Because humans now dominate critical components of the hydrosphere, improving our understanding of the global water cycle has graduated from an academic exercise to an urgent priority. To assess how the water cycle is conceptualized by researchers and the general public, we analyzed 455 water cycle diagrams from textbooks, scientific articles, and online image searches performed in different languages. Only 15% of diagrams integrated human activity into the water cycle and 77% showed no sign of humans whatsoever, although representation of humans varied substantially by region (lowest in China, N. America, and Australia; highest in Western Europe). The abundance and accessibility of freshwater resources were overrepresented, with 98% of diagrams omitting water pollution and climate change, and over 90% of diagrams making no distinction for saline groundwater and lakes. Oceanic aspects of the water cycle (i.e. ocean size, circulation, and precipitation) and related teleconnections were nearly always underrepresented. These patterns held across disciplinary boundaries and through time. We explore the historical and contemporary reasons for some of these biases and present a revised version of the global water cycle based on research from natural and social sciences. We conclude that current depictions of the global water cycle convey a false sense of water security and that reintegrating humans into water cycle diagrams is an important first step towards understanding and sustaining the hydrosocial cycle.

  8. Assessment of elementary school safety restraint programs.

    Science.gov (United States)

    1985-06-01

    The purpose of this research was to identify elementary school (K-6) safety belt : education programs in use in the United States, to review their development, and : to make administrative and impact assessments of their use in selected States. : Six...

  9. Safety-licensing assessment of NASAP reactor concepts and fuel cycle facilities

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Prohammer, F.G.; van Erp, J.B.; Seefeldt, W.B.

    1978-06-01

    Assessments are presented of the safety/licensability of reactor concepts based on information supplied by the Nonproliferation Alternative Systems Assessment Program (NASAP) characterization contractors in their updated responses to the data package for NASAP Rolling Report II. The assessment of the LMFBR includes information from a characterization contractor on alternate fuel cycles but does not include information provided by a characterization contractor on plant-related safety issues. The information provided by the characterization contractors was supplemented by assessments provided by the U. S. Nuclear Regulatory Commission

  10. Causal Diagrams for Empirical Research

    OpenAIRE

    Pearl, Judea

    1994-01-01

    The primary aim of this paper is to show how graphical models can be used as a mathematical language for integrating statistical and subject-matter information. In particular, the paper develops a principled, nonparametric framework for causal inference, in which diagrams are queried to determine if the assumptions available are sufficient for identifiying causal effects from non-experimental data. If so the diagrams can be queried to produce mathematical expressions for causal effects in ter...

  11. Graphical matching rules for cardinality based service feature diagrams

    Directory of Open Access Journals (Sweden)

    Faiza Kanwal

    2017-03-01

    Full Text Available To provide efficient services to end-users, variability and commonality among the features of the product line is a challenge for industrialist and researchers. Feature modeling provides great services to deal with variability and commonality among the features of product line. Cardinality based service feature diagrams changed the basic framework of service feature diagrams by putting constraints to them, which make service specifications more flexible, but apart from their variation in selection third party services may have to be customizable. Although to control variability, cardinality based service feature diagrams provide high level visual notations. For specifying variability, the use of cardinality based service feature diagrams raises the problem of matching a required feature diagram against the set of provided diagrams.

  12. Ultraviolet safety assessments of insect light traps.

    Science.gov (United States)

    Sliney, David H; Gilbert, David W; Lyon, Terry

    2016-01-01

    Near-ultraviolet (UV-A: 315-400 nm), "black-light," electric lamps were invented in 1935 and ultraviolet insect light traps (ILTs) were introduced for use in agriculture around that time. Today ILTs are used indoors in several industries and in food-service as well as in outdoor settings. With recent interest in photobiological lamp safety, safety standards are being developed to test for potentially hazardous ultraviolet emissions. A variety of UV "Black-light" ILTs were measured at a range of distances to assess potential exposures. Realistic time-weighted human exposures are shown to be well below current guidelines for human exposure to ultraviolet radiation. These UV-A exposures would be far less than the typical UV-A exposure in the outdoor environment. Proposals are made for realistic ultraviolet safety standards for ILT products.

  13. Using Photographs and Diagrams to Test Young Children's Mass Thinking

    Science.gov (United States)

    Cheeseman, Jill; McDonough, Andrea

    2013-01-01

    This paper reports the results of a pencil-and-paper test developed to assess young children's understanding of mass measurement. The innovative element of the test was its use of photographs. We found many children of the 295 6-8 year-old children tested could "read" the photographs and diagrams and recognise the images as…

  14. A study on the methodology of probabilistic safety assessment for KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kwon, Young Min; Lee, Yong Bum; Jeong, Hae Yong; Yang, Joon Eon; Ha, Kyu Suk; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    Existing Probabilistic Safety Assessment(PSA) is a method for Light Water Reactor or Pressurized Heavy Water Reactor. Because KALIMER is different from these reactor, the new methodology of PSA need to be developed. In this paper, the PSA of Power Reactor Inherently Safety Module(PRISM) is analyzed, and Initiating Event such as Experiential Assessment, Logical Assessment and Failure Mode Effect Analysis(FMEA) is reviewed. Also, Pipe Damage Frequency Method is suggested for KALIMER. And the Reliability Physical method of Passive System, which is a chief safety system of KALIMER, is reviewed and its applicability is investigated. Finally, for the Preliminary PSA of KALIMER, Intermediate Heat Transfer System is analyzed. 23 refs., 10 figs., 13 tabs. (Author)

  15. Climate and climate-related issues for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    2010-12-01

    The purpose of this report is to document current scientific knowledge on climate and climate-related conditions, relevant to the long-term safety of a KBS-3 repository, to a level required for an adequate treatment in the safety assessment SR-Site. The report also presents a number of dedicated studies on climate and selected climate-related processes of relevance for the assessment of long term repository safety. Based on this information, the report presents a number of possible future climate developments for Forsmark, the site selected for building a repository for spent nuclear fuel in Sweden (Figure 1-1). The presented climate developments are used as basis for the selection and analysis of SR-Site safety assessment scenarios in the SR-Site main report /SKB 2011/. The present report is based on research conducted and published by SKB as well as on research reported in the general scientific literature

  16. Climate and climate-related issues for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    The purpose of this report is to document current scientific knowledge on climate and climate-related conditions, relevant to the long-term safety of a KBS-3 repository, to a level required for an adequate treatment in the safety assessment SR-Site. The report also presents a number of dedicated studies on climate and selected climate-related processes of relevance for the assessment of long term repository safety. Based on this information, the report presents a number of possible future climate developments for Forsmark, the site selected for building a repository for spent nuclear fuel in Sweden (Figure 1-1). The presented climate developments are used as basis for the selection and analysis of SR-Site safety assessment scenarios in the SR-Site main report /SKB 2011/. The present report is based on research conducted and published by SKB as well as on research reported in the general scientific literature

  17. Probabilistic safety assessment for food irradiation facility

    International Nuclear Information System (INIS)

    Solanki, R.B.; Prasad, M.; Sonawane, A.U.; Gupta, S.K.

    2012-01-01

    Highlights: ► Different considerations are required in PSA for Non-Reactor Nuclear Facilities. ► We carried out PSA for food irradiation facility as a part of safety evaluation. ► The results indicate that the fatal exposure risk is below the ‘acceptable risk’. ► Adequate operator training and observing good safety culture would reduce the risk. - Abstract: Probabilistic safety assessment (PSA) is widely used for safety evaluation of Nuclear Power Plants (NPPs) worldwide. The approaches and methodologies are matured and general consensus exists on using these approaches in PSA applications. However, PSA applications for safety evaluation for non-reactor facilities are limited. Due to differences in the processes in nuclear reactor facilities and non-reactor facilities, the considerations are different in application of PSA to these facilities. The food irradiation facilities utilize gamma irradiation sources, X-ray machines and electron accelerators for the purpose of radiation processing of variety of food items. This is categorized as Non-Reactor Nuclear Facility. In this paper, the application of PSA to safety evaluation of food irradiation facility is presented considering the ‘fatality due to radiation overexposure’ as a risk measure. The results indicate that the frequency of the fatal exposure is below the numerical acceptance guidance for the risk to the individual. Further, it is found that the overall risk to the over exposure can be reduced by providing the adequate operator training and observing good safety culture.

  18. Fusion safety data base

    International Nuclear Information System (INIS)

    Laats, E.T.; Hardy, H.A.

    1983-01-01

    The purpose of this Fusion Safety Data Base Program is to provide a repository of data for the design and development of safe commercial fusion reactors. The program is sponsored by the United States Department of Energy (DOE), Office of Fusion Energy. The function of the program is to collect, examine, permanently store, and make available the safety data to the entire US magnetic-fusion energy community. The sources of data will include domestic and foreign fusion reactor safety-related research programs. Any participant in the DOE Program may use the Data Base Program from his terminal through user friendly dialog and can view the contents in the form of text, tables, graphs, or system diagrams

  19. Mort User's Manual: For use with the Management Oversight and Risk Tree analytical logic diagram

    Energy Technology Data Exchange (ETDEWEB)

    Knox, N.W.; Eicher, R.W.

    1992-02-01

    This report contains the User's Manual for MORT (Management Oversight and Risk Tree), a logic diagram in the form of a work sheet'' that illustrates a long series of interrelated questions. MORT is a comprehensive analytical procedure that provides a disciplined method for determining the causes and contributing factors of major accidents. Alternatively, it serves as a tool to evaluate the quality of an existing system. While similar in many respects to fault tree analysis, MORT is more generalized and presents over 1500 specific elements of an ideal universal'' management program for optimizing environment, safety and health, and other programs. This User's Manual is intended to be used with the MORT diagram dated February 1992.

  20. On the Impact of Layout Quality to Understanding UML Diagrams: Diagram Type and Expertise

    DEFF Research Database (Denmark)

    Störrle, Harald

    2012-01-01

    Practical experience suggests that the use and understanding of UML diagrams is greatly affected by the quality of their layout. In previous work, we have presented evidence supporting this intuition. This contrasts with earlier experiments that yielded weak or inconclusive evidence only. In the ......Practical experience suggests that the use and understanding of UML diagrams is greatly affected by the quality of their layout. In previous work, we have presented evidence supporting this intuition. This contrasts with earlier experiments that yielded weak or inconclusive evidence only...

  1. Y-12 Plant Remedial Action technology logic diagram. Volume I: Technology evaluation

    International Nuclear Information System (INIS)

    1994-09-01

    The Y-12 Plant Remedial Action Program addresses remediation of the contaminated groundwater, surface water and soil in the following areas located on the Oak Ridge Reservation: Chestnut Ridge, Bear Creek Valley, the Upper and Lower East Fork Popular Creek Watersheds, CAPCA 1, which includes several areas in which remediation has been completed, and CAPCA 2, which includes dense nonaqueous phase liquid wells and a storage facility. There are many facilities within these areas that are contaminated by uranium, mercury, organics, and other materials. This Technology Logic Diagram identifies possible remediation technologies that can be applied to the soil, water, and contaminants for characterization, treatment, and waste management technology options are supplemented by identification of possible robotics or automation technologies. These would facilitate the cleanup effort by improving safety, of remediation, improving the final remediation product, or decreasing the remediation cost. The Technology Logic Diagram was prepared by a diverse group of more than 35 scientists and engineers from across the Oak Ridge Reservation. Most are specialists in the areas of their contributions. 22 refs., 25 tabs

  2. Complementary safety assessment assessment of nuclear facilities - La Hague plant - AREVA

    International Nuclear Information System (INIS)

    2011-01-01

    This complementary safety assessment analyses the robustness of La Hague plant to extreme situations such as those that led to the Fukushima accident. Robustness is the ability for the plant to withstand events beyond which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Moreover, safety is not only a matter of design or engineered systems but also a matter of organizing: task organization (including subcontracting) as well as the setting of emergency plans or the inventory of nuclear materials are taken into consideration in this assessment. This report is divided into 10 main chapters: 1) the feedback experience of the Fukushima accident; 2) description of the site; 3) featuring the activities and installations; 4) accidental sequences 5) protection from the earthquake; 6) protection from the flood; 7) protection from other extreme natural disasters; 8) the loss of electrical power and of the heat sink; 9) the management of severe accidents; and 10) subcontracting policy. This study shows a globally good robustness of the plant for the considered risks and, in the case of a severe accident, specified remedial actions can be brought into play by the staff to secure the installations. (A.C.)

  3. Cyber Security Risk Assessment for the KNICS Safety Systems

    International Nuclear Information System (INIS)

    Lee, C. K.; Park, G. Y.; Lee, Y. J.; Choi, J. G.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.

    2008-01-01

    In the Korea Nuclear I and C Systems Development (KNICS) project the platforms for plant protection systems are developed, which function as a reactor shutdown, actuation of engineered safety features and a control of the related equipment. Those are fully digitalized through the use of safety-grade programmable logic controllers (PLCs) and communication networks. In 2006 the Regulatory Guide 1.152 (Rev. 02) was published by the U.S. NRC and it describes the application of a cyber security to the safety systems in the Nuclear Power Plant (NPP). Therefore it is required that the new requirements are incorporated into the developed platforms to apply to NPP, and a cyber security risk assessment is performed. The results of the assessment were input for establishing the cyber security policies and planning the work breakdown to incorporate them

  4. Complementary safety assessment assessment of nuclear facilities - Tricastin facility - AREVA

    International Nuclear Information System (INIS)

    2011-01-01

    This complementary safety assessment analyses the robustness of the Areva part of the Tricastin nuclear site to extreme situations such as those that led to the Fukushima accident. This study includes the following facilities: Areva NC Pierrelatte, EURODIF production, Comurhex Pierrelatte, Georges Besse II plant and Socatri. Robustness is the ability for the plant to withstand events beyond which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accidental sequence. Moreover, safety is not only a matter of design or engineered systems but also a matter of organizing: task organization (including subcontracting) as well as the setting of emergency plans or the inventory of nuclear materials are taken into consideration in this assessment. This report is divided into 10 main chapters: 1) the feedback experience of the Fukushima accident; 2) description of the site and its surroundings; 3) featuring of the site's activities and installations; 4) accidental sequences; 5) protection from earthquakes; 6) protection from floods; 7) protection from other extreme natural disasters; 8) the loss of electrical power and of the heat sink; 9) the management of severe accidents; and 10) subcontracting policy. This analysis has identified 5 main measures to be taken to limit the risks linked to natural disasters: -) continuing the program for replacing the current conversion plant and the enrichment plant; -) renewing the storage of hydrofluoric acid at the de-fluorination workshop; -) assessing the seismic behaviour of some parts of the de-fluorination workshop and of the fluorine fabrication workshop; -) improving the availability of warning and information means in case of emergency; and -) improving the means to mitigate accidental gaseous releases. (A.C.)

  5. Safety and immunotoxicity assessment of immunomodulatory monoclonal antibodies

    Science.gov (United States)

    Morton, Laura Dill; Spindeldreher, Sebastian; Kiessling, Andrea; Allenspach, Roy; Hey, Adam; Muller, Patrick Y; Frings, Werner; Sims, Jennifer

    2010-01-01

    Most therapeutic monoclonal antibodies (mAbs) licensed for human use or in clinical development are indicated for treatment of patients with cancer and inflammatory/autoimmune disease and as such, are designed to directly interact with the immune system. A major hurdle for the development and early clinical investigation of many of these immunomodulatory mAbs is their inherent risk for adverse immune-mediated drug reactions in humans such as infusion reactions, cytokine storms, immunosuppression and autoimmunity. A thorough understanding of the immunopharmacology of a mAb in humans and animals is required to both anticipate the clinical risk of adverse immunotoxicological events and to select a safe starting dose for first-in-human (FIH) clinical studies. This review summarizes the most common adverse immunotoxicological events occurring in humans with immunomodulatory mAbs and outlines non-clinical strategies to define their immunopharmacology and assess their immunotoxic potential, as well as reduce the risk of immunotoxicity through rational mAb design. Tests to assess the relative risk of mAb candidates for cytokine release syndrome, innate immune system (dendritic cell) activation and immunogenicity in humans are also described. The importance of selecting a relevant and sensitive toxicity species for human safety assessment in which the immunopharmacology of the mAb is similar to that expected in humans is highlighted, as is the importance of understanding the limitations of the species selected for human safety assessment and supplementation of in vivo safety assessment with appropriate in vitro human assays. A tiered approach to assess effects on immune status, immune function and risk of infection and cancer, governed by the mechanism of action and structural features of the mAb, is described. Finally, the use of immunopharmacology and immunotoxicity data in determining a minimum anticipated biologic effect Level (MABEL) and in the selection of safe human

  6. Beyond Feynman Diagrams (1/3)

    CERN Multimedia

    CERN. Geneva

    2013-01-01

    For decades the central theoretical tool for computing scattering amplitudes has been the Feynman diagram. However, Feynman diagrams are just too slow, even on fast computers, to be able to go beyond the leading order in QCD, for complicated events with many jets of hadrons in the final state. Such events are produced copiously at the LHC, and constitute formidable backgrounds to many searches for new physics. Over the past few years, alternative methods that go beyond ...

  7. Wind Diagrams in Medieval Iceland

    DEFF Research Database (Denmark)

    Kedwards, Dale

    2014-01-01

    This article presents a study of the sole wind diagram that survives from medieval Iceland, preserved in the encyclopaedic miscellany in Copenhagen's Arnamagnæan Institute with the shelf mark AM 732b 4to (c. 1300-25). It examines the wind diagram and its accompanying text, an excerpt on the winds...... from Isidore of Seville's Etymologies. It also examines the perimeter of winds on two medieval Icelandic world maps, and the visual traditions from which they draw....

  8. Bayesian Networks and Influence Diagrams

    DEFF Research Database (Denmark)

    Kjærulff, Uffe Bro; Madsen, Anders Læsø

     Probabilistic networks, also known as Bayesian networks and influence diagrams, have become one of the most promising technologies in the area of applied artificial intelligence, offering intuitive, efficient, and reliable methods for diagnosis, prediction, decision making, classification......, troubleshooting, and data mining under uncertainty. Bayesian Networks and Influence Diagrams: A Guide to Construction and Analysis provides a comprehensive guide for practitioners who wish to understand, construct, and analyze intelligent systems for decision support based on probabilistic networks. Intended...

  9. Probabilistic assessment methods as a tool for developing nations to make safety decisions

    International Nuclear Information System (INIS)

    Gumley, P.; Inamdar, S.V.

    1985-01-01

    This paper advocates the use of probabilistic safety assessment methods in making safety decisions. It discusses the question of adequate safety - what it means to a country buying a nuclear power plant, and how probabilistic safety assessment studies of the reference plant can be used for ensuring this adequate safety. It is proposed that adequate safety means ensuring that the plant would behave, in accident conditions, in a manner similar to the way it is expected to behave were it in the country of origin. For this one needs to know how the plant responds under somewhat altered conditions. These altered conditions can arise from such factors as varying reliability of electrical grids, different manufacturing technology, local systems design and operator capability. In the design of nuclear power plants, the traditional approach to safety has led to the belief that availability and effectiveness of safety systems alone are all that is required to ensure plant safety. This belief can result in design oversights leading to potential problems arising from the power production systems and the service systems. Participation by the buying country in the design of such systems, and understanding the safety implications thereof, can be facilitated by probabilistic safety assessment methods. This philosophy is illustrated in this paper by examples. (author)

  10. Proof test diagrams for Zerodur glass-ceramic

    Science.gov (United States)

    Tucker, D. S.

    1991-01-01

    Proof test diagrams for Zerodur glass-ceramics are calculated from available fracture mechanics data. It is shown that the environment has a large effect on minimum time-to-failure as predicted by proof test diagrams.

  11. Assessing Risk-Based Performance Indicators in Safety-Critical Systems for Nuclear Power Plants

    OpenAIRE

    TONT Gabriela

    2011-01-01

    The paper proposes framework for a multidisciplinary nuclear risk and safety assessment by modeling uncertainty and combining diverse evidence provided in such a way that it could be used to represent an entire argument about a system's dependability. The identified safety issues are being treated by means of probabilistic safety assessment (PSA). The behavior simulation of power plant in thepresence of risk factors is analyzed from the vulnerability, risk and functional safety viewpoints, hi...

  12. Assessment of patient safety culture in private and public hospitals in Peru.

    Science.gov (United States)

    Arrieta, Alejandro; Suárez, Gabriela; Hakim, Galed

    2018-04-01

    To assess the patient safety culture in Peruvian hospitals from the perspective of healthcare professionals, and to test for differences between the private and public healthcare sectors. Patient safety is defined as the avoidance and prevention of patient injuries or adverse events resulting from the processes of healthcare delivery. A non-random cross-sectional study conducted online. An online survey was administered from July to August 2016, in Peru. This study reports results from Lima and Callao, which are the capital and the port region of Peru. A total of 1679 healthcare professionals completed the survey. Participants were physicians, medical residents and nurses working in healthcare facilities from the private sector and public sector. Assessment of the degree of patient safety and 12 dimensions of patient safety culture in hospital units as perceived by healthcare professionals. Only 18% of healthcare professionals assess the degree of patient safety in their unit of work as excellent or very good. Significant differences are observed between the patient safety grades in the private sector (37%) compared to the public sub-sectors (13-15%). Moreover, in all patient safety culture dimensions, healthcare professionals from the private sector give more favorable responses for patient safety, than those from the public sub-systems. The most significant difference in support comes from patient safety administrators through communication and information about errors. Overall, the degree of patient safety in Peru is low, with significant gaps that exist between the private and the public sectors.

  13. HANFORD SAFETY ANALYSIS and RISK ASSESSMENT HANDBOOK (SARAH)

    International Nuclear Information System (INIS)

    EVANS, C.B.

    2004-01-01

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S and M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard

  14. A safety assessment of the SEAFP fuel cycle systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Kalyanam, K.; Ciattaglia, S.; Pace, L. di

    1995-01-01

    CFFTP and ENEA participated in a joint safety assessment of the fuel cycle design developed for the SEAFP fusion power reactor study (SEAFP: Safety and Environmental Assessment of Fusion Power). The assessment considered both conventional (deflagation/detonation) and radioactive hazards associated with the handling of significant quantities of hydrogen isotopes (H, D and T). Accordingly, the assessment focused on systems or equipment where either the flow rate, or inventory, of hydrogen isotopes was large. A systematic and thorough assessment of initiating events that can lead to an accidental release of tritium into the environment was the first step of the analysis process. This review demonstrated that, in all cases, there are at least two lines of defence available for mitigating the consequences of such accidents -i.e., secondary confinement (glove box, second pipe, caisson, etc.) and the building confinement, backed-up by an air detritiation capability. Therefore, large releases of tritium to the environment will occur only at very low frequencies. (orig.)

  15. Method of operator safety assessment for underground mobile mining equipment

    Science.gov (United States)

    Działak, Paulina; Karliński, Jacek; Rusiński, Eugeniusz

    2018-01-01

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

  16. Method of operator safety assessment for underground mobile mining equipment

    Directory of Open Access Journals (Sweden)

    Działak Paulina

    2018-01-01

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

  17. Establishment of safety goal and its quantification based on risk assessment

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Muramatsu, Ken

    2017-01-01

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

  18. Regulatory status on the safety assessment of a HLW repository in other countries

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Hwang, Yong Soo

    2008-12-01

    To construct a HLW repository, it is essential to meet the requirements on the regulation for a deep geological disposal. Even if the construction of a HLW repository is determined positively, technical standards which assert the performance of a repository will be needed. Among various technical standards, safety assessment based on the repository evolution in the future will play an important role in the licensing process. The foreign countries' technical standards on the safety assessment of a HLW repository may be an indicator to carry out the R and D activities on geological disposal effectively. In this report, assessment period, limit of radiation dose and uncertainty related to the safety assessment are investigated and analyzed in detail. Especially, the technical reviews of USA regulation bodies seems to be reasonable in the point of the intrinsic attribute of safety assessment

  19. Non-technical issues in safety assessments for nuclear disposal facilities

    International Nuclear Information System (INIS)

    Kallenbach-Herbert, Beate; Brohmann, Bettina

    2010-09-01

    The paper highlights that a comprehensive approach to safety affords the consideration of technology, organisation, personnel and social environment. In several safety relevant contexts of nuclear waste disposal these fields are closely interrelated. The approach for the consideration of socio-scientific aspects which is sketched in this paper supports the systematic treatment of safety relevant non-technical issues in the safety case or in safety assessments for a disposal project. Furthermore it may foster the dialogue among specialists from the technical, the natural- and the socio-scientific field on questions of disposal safety. In this way it may contribute to a better understanding among the affected scientific disciplines in nuclear waste disposal.

  20. Safety assessment of Olkiluoto NPP units 1 and 2. Decision of the Radiation and Nuclear Safety Authority regarding the periodic safety review of the Olkiluoto NPP

    International Nuclear Information System (INIS)

    2010-02-01

    In this safety assessment the Radiation and Nuclear Safety Authority (STUK) has evaluated the safety of the Olkiluoto Nuclear Power Plant units 1 and 2 in connection with the periodic safety review. This safety assessment provides a summary of the reviews, inspections and continuous oversight carried out by STUK. The issues addressed in the assessment and the related evaluation criteria are set forth in the nuclear energy and radiation safety legislation and the regulations issued thereunder. The provisions of the Nuclear Energy Act concerning the safe use of nuclear energy, security and emergency preparedness arrangements, and waste management are specified in more detail in the Government Decrees and Regulatory Guides issued by STUK. Based on the assessment, STUK consideres that the Olkiluoto Nuclear Power Plant units 1 and 2 meet the set safety requirements for operational nuclear power plants, the emergency preparedness arrangements are sufficient and the necessary control to prevent the proliferation of nuclear weapons has been appropriately arranged. The physical protection of the Olkiluoto nuclear power plant is not yet completely in compliance with the requirements of Government Decree 734/2008, which came into force in December 2008. Further requirements concerning this issue based also on the principle of continuous improvement were included in the decision relating to the periodic safety review. The safety of the Olkiluoto nuclear power plant was assessed in compliance with the Government Decree on the Safety of Nuclear Power Plants (733/2008), which came into force in 2008. The decree notes that existing nuclear power plants need not meet all the requirements set out for new plants. Most of the design bases pertaining to the Olkiluoto 1 and 2 nuclear power plant units were set in the 1970s. Substantial modernisations have been carried out at the Olkiluoto 1 and 2 nuclear power plant units since their commissioning to improve safety. This is in line with