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

International Nuclear Information System (INIS)

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

2017-01-01

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

Hybrid probabilistic and possibilistic safety assessment. Methodology and application

International Nuclear Information System (INIS)

Kato, Kazuyuki; Amano, Osamu; Ueda, Hiroyoshi; Ikeda, Takao; Yoshida, Hideji; Takase, Hiroyasu

2002-01-01

This paper presents a unified methodology to handle variability and ignorance by using probabilistic and possibilistic techniques respectively. The methodology has been applied to the safety assessment of geological disposal of high-level radioactive waste. Uncertainties associated with scenarios, models and parameters were defined in terms of fuzzy membership functions derived through a series of interviews to the experts, while variability was formulated by means of probability density functions (pdfs) based on available data sets. The exercise demonstrated the applicability of the new methodology and, in particular, its advantage in quantifying uncertainties based on expert opinion and in providing information on the dependence of assessment results on the level of conservatism. In addition, it was shown that sensitivity analysis can identify key parameters contributing to uncertainties associated with results of the overall assessment. The information mentioned above can be utilized to support decision-making and to guide the process of disposal system development and optimization of protection against potential exposure. (author)

Safety assessment methodology for waste repositories in deep geological formations

International Nuclear Information System (INIS)

Chapuis, A.M.; Lewi, J.; Pradel, J.; Queniart, D.; Raimbault, P.; Assouline, M.

1986-06-01

The long term safety of a nuclear waste repository relies on the evaluation of the doses which could be transferred to man in the future. This implies a detailed knowledge of the medium where the waste will be confined, the identification of the basic phenomena which govern the migration of the radionuclides and the investigation of all possible scenarios that may affect the integrity of the barriers between the waste and the biosphere. Inside the Institute of protection and nuclear safety of the French Atomic Energy Commission (CEA/IPSN), the Department of the Safety Analysis (DAS) is currently developing a methodology for assessing the safety of future geological waste repositories, and is in charge of the modelling development, while the Department of Technical Protection (DPT) is in charge of the geological experimental studies. Both aspects of this program are presented. The methodology for risk assessment stresses the needs for coordination between data acquisition and model development which should result in the obtention of an efficient tool for safety evaluation. Progress needs to be made in source and geosphere modelling. Much more sophisticated models could be used than the ones which is described; however sensitivity analysis will determine the level of sophistication which is necessary to implement. Participation to international validation programs are also very important for gaining confidence in the approaches which have been chosen

Methodology for assessment of safety risk due to potential accidents in US gaseous diffusion plants

International Nuclear Information System (INIS)

Turner, J.H.; O'Kain, D.U.

1991-01-01

Gaseous diffusion plants that operate in the United States represent a unique combination of nuclear and chemical hazards. Assessing and controlling the health, safety, and environmental risks that can result from natural phenomena events, process upset conditions, and operator errors require a unique methodology. Such a methodology has been developed for the diffusion plants and is being utilized to assess and control the risk of operating the plants. A summary of the methodology developed to assess the unique safety risks at the US gaseous diffusion plants is presented in this paper

A safety assessment methodology applied to CNS/ATM-based air traffic control system

Energy Technology Data Exchange (ETDEWEB)

Vismari, Lucio Flavio, E-mail: lucio.vismari@usp.b [Safety Analysis Group (GAS), School of Engineering at University of Sao Paulo (Poli-USP), Av. Prof. Luciano Gualberto, Trav.3, n.158, Predio da Engenharia de Eletricidade, Sala C2-32, CEP 05508-900, Sao Paulo (Brazil); Batista Camargo Junior, Joao, E-mail: joaocamargo@usp.b [Safety Analysis Group (GAS), School of Engineering at University of Sao Paulo (Poli-USP), Av. Prof. Luciano Gualberto, Trav.3, n.158, Predio da Engenharia de Eletricidade, Sala C2-32, CEP 05508-900, Sao Paulo (Brazil)

2011-07-15

In the last decades, the air traffic system has been changing to adapt itself to new social demands, mainly the safe growth of worldwide traffic capacity. Those changes are ruled by the Communication, Navigation, Surveillance/Air Traffic Management (CNS/ATM) paradigm , based on digital communication technologies (mainly satellites) as a way of improving communication, surveillance, navigation and air traffic management services. However, CNS/ATM poses new challenges and needs, mainly related to the safety assessment process. In face of these new challenges, and considering the main characteristics of the CNS/ATM, a methodology is proposed at this work by combining 'absolute' and 'relative' safety assessment methods adopted by the International Civil Aviation Organization (ICAO) in ICAO Doc.9689 , using Fluid Stochastic Petri Nets (FSPN) as the modeling formalism, and compares the safety metrics estimated from the simulation of both the proposed (in analysis) and the legacy system models. To demonstrate its usefulness, the proposed methodology was applied to the 'Automatic Dependent Surveillance-Broadcasting' (ADS-B) based air traffic control system. As conclusions, the proposed methodology assured to assess CNS/ATM system safety properties, in which FSPN formalism provides important modeling capabilities, and discrete event simulation allowing the estimation of the desired safety metric.

A safety assessment methodology applied to CNS/ATM-based air traffic control system

International Nuclear Information System (INIS)

Vismari, Lucio Flavio; Batista Camargo Junior, Joao

2011-01-01

In the last decades, the air traffic system has been changing to adapt itself to new social demands, mainly the safe growth of worldwide traffic capacity. Those changes are ruled by the Communication, Navigation, Surveillance/Air Traffic Management (CNS/ATM) paradigm , based on digital communication technologies (mainly satellites) as a way of improving communication, surveillance, navigation and air traffic management services. However, CNS/ATM poses new challenges and needs, mainly related to the safety assessment process. In face of these new challenges, and considering the main characteristics of the CNS/ATM, a methodology is proposed at this work by combining 'absolute' and 'relative' safety assessment methods adopted by the International Civil Aviation Organization (ICAO) in ICAO Doc.9689 , using Fluid Stochastic Petri Nets (FSPN) as the modeling formalism, and compares the safety metrics estimated from the simulation of both the proposed (in analysis) and the legacy system models. To demonstrate its usefulness, the proposed methodology was applied to the 'Automatic Dependent Surveillance-Broadcasting' (ADS-B) based air traffic control system. As conclusions, the proposed methodology assured to assess CNS/ATM system safety properties, in which FSPN formalism provides important modeling capabilities, and discrete event simulation allowing the estimation of the desired safety metric.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

ASAM - The international programme on application of safety assessment methodologies for near surface radioactive waste disposal facilities

International Nuclear Information System (INIS)

Batandjieva, B.

2002-01-01

The IAEA has launched a new Co-ordinated Research Project (CRP) on Application of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ASAM). The CRP will focus on the practical application of the safety assessment methodology, developed under the ISAM programme, for different purposes, such as developing design concepts, licensing, upgrading existing repositories, reassessment of operating disposal facilities. The overall aim of the programme is to assist safety assessors, regulators and other specialists involved in the development and review of safety assessment for near surface disposal facilities in order to achieve transparent, traceable and defendable evaluation of safety of these facilities. (author)

Safety Assessment for Decommissioning of Nuclear Facilities - From Methodology to the Use of Results in Decision Making

International Nuclear Information System (INIS)

Batandjieva, B.; Ferch, R.; Joubert, A.; Kaulard, J.; Manson, P.; Percival, K.; Thierfeldt, St.

2008-01-01

The safety assessment of operational facilities in the nuclear industry is well understood and methodologies have been developed and refined over several decades. Similarly safety assessment methodologies for near surface disposal facilities have been harmonized internationally during the last few years. There is however relatively less widespread and documented experience of safety assessment for decommissioning among Member States of the International Atomic Energy Agency (IAEA) and consequently there is less commonalty of approaches internationally. The importance of safety during decommissioning was further emphasized at the first review meeting of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, and the Berlin Conference 'Safe Decommissioning for Nuclear Activities' (14-18 October 2002). As a consequence during its June 2004 meeting the IAEA Board of Governors approved an Action Plan on Decommissioning of nuclear Facilities that requested the Secretariat 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 the IAEA launched the International Project Evaluation and Demonstration of Safety during Decommissioning of Nuclear Facilities (DeSa) in November 2004 with the following objectives: - To develop a harmonized approach to safety assessment and define the elements of safety assessment for decommissioning; - To investigate the practical applicability of the methodology and performance of safety assessments for the decommissioning of various types of facilities through a selected number of test cases; - To investigate approaches for review of safety assessments for decommissioning activities and the development of a regulatory

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)

A survey of dynamic methodologies for probabilistic safety assessment of nuclear power plants

International Nuclear Information System (INIS)

Aldemir, Tunc

2013-01-01

Highlights: ► Dynamic methodologies for probabilistic safety assessment (PSA) are surveyed. ► These methodologies overcome the limitations of the traditional approach to PSA. ► They are suitable for PSA using a best estimate plus uncertainty approach. ► They are highly computation intensive and produce very large number of scenarios. ► Use of scenario clustering can assist the analysis of the results. -- Abstract: Dynamic methodologies for probabilistic safety assessment (PSA) are defined as those which use a time-dependent phenomenological model of system evolution along with its stochastic behavior to account for possible dependencies between failure events. Over the past 30 years, numerous concerns have been raised in the literature regarding the capability of the traditional static modeling approaches such as the event-tree/fault-tree methodology to adequately account for the impact of process/hardware/software/firmware/human interactions on the stochastic system behavior. A survey of the types of dynamic PSA methodologies proposed to date is presented, as well as a brief summary of an example application for the PSA modeling of a digital feedwater control system of an operating pressurized water reactor. The use of dynamic methodologies for PSA modeling of passive components and phenomenological uncertainties are also discussed.

ISAM news. International programme on implementation of safety assessment methodologies for near surface disposal facilities for radioactive waste (ISAM 1997-1999)

International Nuclear Information System (INIS)

Torres, Carlos

1996-01-01

The scope of the programme will be the scientific and technical aspects related to the long term safety assessment of near disposal facilities. The primary focus of ISAM will be on the methodological aspects of safety assessment with emphasis on the practical application of these methodologies. Furthermore, practical application is necessary for for a thorough understanding of safety assessment methodologies. The programme will address important methodological issues associated with long term safety assessment of near surface disposal systems. At least three important areas will be covered: (1) scenario generation and justification; (2) modelling, data and tools; and (3) analysis of results and confidence building

Safety assessment methodologies for near surface disposal facilities. Results of a co-ordinated research project (ISAM). Volume 1: Review and enhancement of safety assessment approaches and tools. Volume 2: Test cases

International Nuclear Information System (INIS)

2004-07-01

For several decades, countries have made use of near surface facilities for the disposal of low and intermediate level radioactive waste. In line with the internationally agreed principles of radioactive waste management, the safety of these facilities needs to be ensured during all stages of their lifetimes, including the post-closure period. By the mid 1990s, formal methodologies for evaluating the long term safety of such facilities had been developed, but intercomparison of these methodologies had revealed a number of discrepancies between them. Consequently, in 1997, the International Atomic Energy Agency launched a Co-ordinated Research Project (CRP) on Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities (ISAM). The particular objectives of the CRP were to provide a critical evaluation of the approaches and tools used in post-closure safety assessment for proposed and existing near-surface radioactive waste disposal facilities, enhance the approaches and tools used and build confidence in the approaches and tools used. The CRP ran until 2000 and resulted in the development of a harmonized assessment methodology (the ISAM project methodology), which was applied to a number of test cases. Over seventy participants from twenty-two Member States played an active role in the project and it attracted interest from around seven hundred persons involved with safety assessment in seventy-two Member States. The results of the CRP have contributed to the Action Plan on the Safety of Radioactive Waste Management which was approved by the Board of Governors and endorsed by the General Conference in September 2001. Specifically, they contribute to Action 5, which requests the IAEA Secretariat to 'develop a structured and systematic programme to ensure adequate application of the Agency's waste safety standards', by elaborating on the Safety Requirements on 'Near Surface Disposal of Radioactive Waste' (Safety Standards Series No. WS-R-1) and

Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems

Directory of Open Access Journals (Sweden)

Franco Pierro

2009-01-01

Full Text Available The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems. The REPAS can be used in the design of the passive safety systems to assess their goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment studies and it can be seen as a support to Probabilistic Safety Analysis studies. With regard to this, some examples in the application of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the analyses and to model the selected systems. Probability distributions have been assigned to the uncertain input parameters through engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks' formula has been taken into account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets.

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

Application of the integrated safety assessment methodology to the protection of electric systems

International Nuclear Information System (INIS)

Hortal, Javier; Izquierdo, Jose M.

1996-01-01

The generalization of classical techniques for risk assessment incorporating dynamic effects is the main objective of the Integrated Safety Assessment Methodology, as practical implementation of Protection Theory. Transient stability, contingency analysis and protection setpoint verification in electric power systems are particularly appropriate domains of application, since the coupling of reliability and dynamic analysis in the protection assessment process is being increasingly demanded. Suitable techniques for dynamic simulation of sequences of switching events in power systems are derived from the use of quasi-linear equation solution algorithms. The application of the methodology, step by step, is illustrated in a simple but representative example

A study on safety assessment methodology for a vitrification plant

Energy Technology Data Exchange (ETDEWEB)

Seo, Y. C.; Lee, G. S.; Choi, Y. C.; Kim, G. H. [Yonsei Univ., Seoul (Korea, Republic of)

2002-03-15

In this study, the technical and regulatory status of radioactive waste vitrification technologies in foreign and domestic plants is investigated and analyzed, and then significant factors are suggested which must be contained in the final technical guideline or standard for the safety assessment of vitrification plants. Also, the methods to estimate the stability of vitrified waste forms are suggested with property analysis of them. The contents and scope of the study are summarized as follows : survey of the status on radioactive waste vitrification technologies in foreign and domestic plants, survey of the characterization methodology for radioactive waste form, analysis of stability for vitrified waste forms, survey and analysis of technical standards and regulations concerned with them in foreign and domestic plants, suggestion of significant factors for the safety assessment of vitrification plants, submission of regulated technical standard on radioactive waste vitrification plats.

Analysis of cold leg LOCA with failed HPSI by means of integrated safety assessment methodology

International Nuclear Information System (INIS)

Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayorga, J.

2014-01-01

Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties

The IAEA research project on improvement of safety assessment methodologies for near surface disposal facilities

International Nuclear Information System (INIS)

Torres-Vidal, C.; Graham, D.; Batandjieva, B.

2002-01-01

The International Atomic Energy Agency (IAEA) Research Coordinated Project on Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities (ISAM) was launched in November 1997 and it has been underway for three years. The ISAM project was developed to provide a critical evaluation of the approaches and tools used in long-term safety assessment of near surface repositories. It resulted in the development of a harmonised approach and illustrated its application by way of three test cases - vault, borehole and Radon (a particular range of repository designs developed within the former Soviet Union) type repositories. As a consequence, the ISAM project had over 70 active participants and attracted considerable interest involving around 700 experts from 72 Member States. The methodology developed, the test cases, the main lessons learnt and the conclusions have been documented and will be published in the form of an IAEA TECDOC. This paper presents the work of the IAEA on improvement of safety assessment methodologies for near surface waste disposal facilities and the application of these methodologies for different purposes in the individual stages of the repository development. The paper introduces the main objectives, activities and outcome of the ISAM project and summarizes the work performed by the six working groups within the ISAM programme, i.e. Scenario Generation and Justification, Modelling, Confidence Building, Vault, Radon Type Facility and Borehole test cases. (author)

Development of a methodology for assessing the safety of embedded software systems

Science.gov (United States)

Garrett, C. J.; Guarro, S. B.; Apostolakis, G. E.

1993-01-01

A Dynamic Flowgraph Methodology (DFM) based on an integrated approach to modeling and analyzing the behavior of software-driven embedded systems for assessing and verifying reliability and safety is discussed. DFM is based on an extension of the Logic Flowgraph Methodology to incorporate state transition models. System models which express the logic of the system in terms of causal relationships between physical variables and temporal characteristics of software modules are analyzed to determine how a certain state can be reached. This is done by developing timed fault trees which take the form of logical combinations of static trees relating the system parameters at different point in time. The resulting information concerning the hardware and software states can be used to eliminate unsafe execution paths and identify testing criteria for safety critical software functions.

Uncertainty and sensitivity analysis methodology in a level-I PSA (Probabilistic Safety Assessment)

International Nuclear Information System (INIS)

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

1997-01-01

This work presents a methodology for sensitivity and uncertainty analysis, applicable to a probabilistic safety assessment level I. The work contents are: correct association of distributions to parameters, importance and qualification of expert opinions, generations of samples according to sample sizes, and study of the relationships among system variables and system response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well different graphical visualization for the control of the study. (author) [es

Development of a new methodology for quantifying nuclear safety culture

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2017-01-01

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

Development of a new methodology for quantifying nuclear safety culture

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-15

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

Safety, mobility and comfort assessment methodologies of intelligent transport systems for vulnerable road users

NARCIS (Netherlands)

Malone, K.; Silla, A.; Johanssen, C.; Bell, D.

2017-01-01

Introduction: This paper describes the modification and development of methodologies to assess the impacts of Intelligent Transport Systems (ITS) applications for Vulnerable Road users (VRUs) in the domains of safety, mobility and comfort. This effort was carried out in the context of the VRUITS

A study on methodologies for assessing safety critical network's risk impact on Nuclear Power Plant

International Nuclear Information System (INIS)

Lim, T. J.; Lee, H. J.; Park, S. K.; Seo, S. J.

2006-08-01

The objectives of this project is to investigate and study existing reliability analysis techniques for communication networks in order to develop reliability analysis models for Nuclear Power Plant's safety-critical networks. It is necessary to make a comprehensive survey of current methodologies for communication network reliability. Major outputs of the first year study are design characteristics of safety-critical communication networks, efficient algorithms for quantifying reliability of communication networks, and preliminary models for assessing reliability of safety-critical communication networks

RADON-type disposal facility safety case 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)

Guskov, A.; Batanjieva, B.; Kozak, M.W.; Torres-Vidal, C.

2002-01-01

The ISAM safety assessment methodology was applied to RADON-type facilities. The assessments conducted through the ISAM project were among the first conducted for these kinds of facilities. These assessments are anticipated to lead to significantly improved levels of safety in countries with such facilities. Experience gained though this RADON-type Safety Case was already used in Russia while developing national regulatory documents. (author)

An overview of performance assessment methodology

International Nuclear Information System (INIS)

Hongnian Jow

2010-01-01

The definition of performance assessment (PA) within the context of a geologic repository program is a post-closure safety assessment; a system analysis of hazards associated with the facility and the ability of the site and the design of the facility to provide for the safety functions. For the last few decades, PA methodology bas been developed and applied to different waste disposal programs around the world. PA has been used in the safety analyses for waste disposal repositories for low-level waste, intermediate level waste, and high-level waste including spent nuclear fuels. This paper provides an overview of the performance assessment methodology and gives examples of its applications for the Yucca Mountain Project. (authors)

Steps towards the international regulatory acceptance of non-animal methodology in safety assessment.

Science.gov (United States)

Sewell, Fiona; Doe, John; Gellatly, Nichola; Ragan, Ian; Burden, Natalie

2017-10-01

The current animal-based paradigm for safety assessment must change. In September 2016, the UK National Centre for Replacement, Refinement and Reduction of Animals in Research (NC3Rs) brought together scientists from regulatory authorities, academia and industry to review progress in bringing new methodology into regulatory use, and to identify ways to expedite progress. Progress has been slow. Science is advancing to make this possible but changes are necessary. The new paradigm should allow new methodology to be adopted once it is developed rather than being based on a fixed set of studies. Regulatory authorities can help by developing Performance-Based Standards. The most pressing need is in repeat dose toxicology, although setting standards will be more complex than in areas such as sensitization. Performance standards should be aimed directly at human safety, not at reproducing the results of animal studies. Regulatory authorities can also aid progress towards the acceptance of non-animal based methodology by promoting "safe-haven" trials where traditional and new methodology data can be submitted in parallel to build up experience in the new methods. Industry can play its part in the acceptance of new methodology, by contributing to the setting of performance standards and by actively contributing to "safe-haven" trials. Copyright © 2017 The Authors. Published by Elsevier Inc. All rights reserved.

Development of a methodology for the safety assessment of near surface disposal facilities for radioactive waste

International Nuclear Information System (INIS)

Simon, I.; Cancio, D.; Alonso, L.F.; Agueero, A.; Lopez de la Higuera, J.; Gil, E.; Garcia, E.

2000-01-01

The Project on the Environmental Radiological Impact in CIEMAT is developing, for the Spanish regulatory body Consejo de Seguridad Nuclear (CSN), a methodology for the Safety Assessment of near surface disposal facilities. This method has been developed incorporating some elements developed through the participation in the IAEA's ISAM Programme (Improving Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities). The first step of the approach is the consideration of the assessment context, including the purpose of the assessment, the end-Points, philosophy, disposal system, source term and temporal scales as well as the hypothesis about the critical group. Once the context has been established, and considering the peculiarities of the system, an specific list of features, events and processes (FEPs) is produced. These will be incorporated into the assessment scenarios. The set of scenarios will be represented in the conceptual and mathematical models. By the use of mathematical codes, calculations are performed to obtain results (i.e. in terms of doses) to be analysed and compared against the criteria. The methodology is being tested by the application to an hypothetical engineered disposal system based on an exercise within the ISAM Programme, and will finally be applied to the Spanish case. (author)

Progress in the development of methodology for fusion safety systems studies

International Nuclear Information System (INIS)

Ho, S.K.; Cambi, G.; Ciattaglia, S.; Fujii-e, Y.; Seki, Y.

1994-01-01

The development of fusion safety systems-study methodology, including the aspects of schematic classification of overall fusion safety system, qualitative assessment of fusion system for identification of critical accident scenarios, quantitative analysis of accident consequences and risk for safety design evaluation, and system-level analysis of accident consequences and risk for design optimization, by a consortium of international efforts is presented. The potential application of this methodology into reactor design studies will facilitate the systematic assessment of safety performance of reactor designs and enhance the impacts of safety considerations on the selection of design configurations

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)

Methodology for the Assessment of Confidence in Safety Margin for Small Break Loss of Coolant Accident Sequences

Energy Technology Data Exchange (ETDEWEB)

Nagrale, D. B.; Prasad, M.; Rao, R. S.; Gaikwad, A.J., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used concurrently to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis is increasingly being used for deterministic calculation in NPPs. The PSA methodology aims to be as realistic as possible while integrating information about accident phenomena, plant design, operating practices, component reliability and human behaviour. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with the concept of calculating the peak clad temperature with 95 percent confidence and 95 percent probability (PCT{sub 95/95}) in small break loss of coolant accident (SBLOCA) and methodologies for assessing safety margin. Five input parameters mainly, nominal power level, decay power, fuel clad gap conductivity, fuel thermal conductivity and discharge coefficient, were selected. A Uniform probability density function was assigned to the uncertain parameters and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for 5 parameters were randomly mixed by LHS to obtain 25 input sets. A non-core damage accident sequence was selected from the SBLOCA event tree of a typical VVER study to estimate the PCTs and safety margin. A Kolmogorov– Smirnov goodness-of-fit test was carried out for PCTs. The smallest value of safety margin would indicate the robustness of the system with 95% confidence and 95% probability. Regression analysis was also carried out using 1000 sample size for the estimating PCTs. Mean, variance and finally safety margin were analysed. (author)

Review of SKB's interim report of SR-Can: SKI's and SSI's evaluation of SKB's up-dated methodology for safety assessment

International Nuclear Information System (INIS)

Dverstorp, Bjoern; Moberg, Leif; Wiebert, Anders; Xu Shulan; Stroemberg, Bo; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Sundstroem, Benny; Toverud, Oeivind

2005-07-01

This report presents the findings of a review of the Swedish Nuclear Fuel and Waste Management Co.'s (SKB) interim report of the safety assessment SR-Can (SKB TR 04-11), conducted by the Swedish Radiation Protection Authority (SSI) and the Swedish Nuclear Power Inspectorate (SKI). SKB's interim report describes and exemplifies the safety assessment methodology that SKB plans to use in the oncoming licence applications for an encapsulation plant and a final repository for spent nuclear fuel. The authorities' review takes into account the findings of an international peer review of SKB's interim report. The authorities conclude that SKB has improved its safety assessment methodology in several aspects compared to earlier safety reports. Among other things the authorities commend SKB for giving a comprehensive account of relevant regulations and guidance, and for the systematic approach to identification and documentation of features, events and processes that need to be considered in the safety assessment. However, the authorities also conclude that important parts of SKB's method need to be further developed before they are mature enough to be used as a basis for a license application. The authorities' overall assessment is summarised in chapter 8 of this report

A reliability assessment methodology for the VHTR passive safety system

International Nuclear Information System (INIS)

Lee, Hyungsuk; Jae, Moosung

2014-01-01

The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

Analytical methodology for safety validation of computer controlled subsystems. Volume 1 : state-of-the-art and assessment of safety verification/validation methodologies

Science.gov (United States)

1995-09-01

This report describes the development of a methodology designed to assure that a sufficiently high level of safety is achieved and maintained in computer-based systems which perform safety critical functions in high-speed rail or magnetic levitation ...

Preliminary safety analysis methodology for the SMART

Energy Technology Data Exchange (ETDEWEB)

Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

Biosphere modeling for safety assessment to high-level radioactive waste geological disposal. Application of reference biosphere methodology to safety assesment of geological disposal

International Nuclear Information System (INIS)

Baba, Tomoko; Ishihara, Yoshinao; Ishiguro, Katsuhiko; Suzuki, Yuji; Naito, Morimasa

2000-01-01

In the safety assessment of a high-level radioactive waste disposal system, it is required to estimate future radiological impacts on human beings. Consideration of living habits and the human environment in the future involves a large degree of uncertainty. To avoid endless speculation aimed at reducing such uncertainty, an approach is applied for identifying and justifying a 'reference biosphere' for use in safety assessment in Japan. considering a wide range of Japanese geological environments, saline specific reference biospheres' were developed using an approach consistent with the BIOMOVS II reference biosphere methodology. (author)

The International Atomic Energy Agency (IAEA) research program to improve safety assessment methodologies for near-surface radioactive waste disposal facilities (ISAM)

International Nuclear Information System (INIS)

Torres-Vidal, C.; Kozak, M.W.

2000-01-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Program in November 1997 on Improvement of Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities (ISAM). The purpose of this paper is to describe the program and its goals, and to describe achievements of the program to date. The main objectives of the ISAM program are outlined. The primary focus of ISAM is on the practical application of safety assessment methodologies. Three kinds of practical situations are being addressed in the program: safety assessments for large vaults typical of those in Western Europe and North America, smaller vaults for medium and industrial wastes typical in eastern Europe and the former Soviet Union, and a proposed borehole technology for disposal of spent sources in low-technology conditions. (author)

Safety Assessment for Decommissioning

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-06-15

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

Safety class methodology

International Nuclear Information System (INIS)

Donner, E.B.; Low, J.M.; Lux, C.R.

1992-01-01

DOE Order 6430.1A, General Design Criteria (GDC), requires that DOE facilities be evaluated with respect to ''safety class items.'' Although the GDC defines safety class items, it does not provide a methodology for selecting safety class items. The methodology described in this paper was developed to assure that Safety Class Items at the Savannah River Site (SRS) are selected in a consistent and technically defensible manner. Safety class items are those in the highest of four categories determined to be of special importance to nuclear safety and, merit appropriately higher-quality design, fabrication, and industrial test standards and codes. The identification of safety class items is approached using a cascading strategy that begins at the 'safety function' level (i.e., a cooling function, ventilation function, etc.) and proceeds down to the system, component, or structure level. Thus, the items that are required to support a safety function are SCls. The basic steps in this procedure apply to the determination of SCls for both new project activities, and for operating facilities. The GDC lists six characteristics of SCls to be considered as a starting point for safety item classification. They are as follows: 1. Those items whose failure would produce exposure consequences that would exceed the guidelines in Section 1300-1.4, ''Guidance on Limiting Exposure of the Public,'' at the site boundary or nearest point of public access 2. Those items required to maintain operating parameters within the safety limits specified in the Operational Safety Requirements during normal operations and anticipated operational occurrences. 3. Those items required for nuclear criticality safety. 4. Those items required to monitor the release of radioactive material to the environment during and after a Design Basis Accident. Those items required to achieve, and maintain the facility in a safe shutdown condition 6. Those items that control Safety Class Item listed above

International Expert Review of Sr-Can: Safety Assessment Methodology - External review contribution in support of SSI's and SKI's review of SR-Can

International Nuclear Information System (INIS)

Sagar, Budhi; Egan, Michael; Roehlig, Klaus-Juergen; Chapman, Neil; Wilmot, Roger

2008-03-01

In 2006, SKB published a safety assessment (SR-Can) as part of its work to support a licence application for the construction of a final repository for spent nuclear fuel. The purposes of the SR-Can project were stated in the main project report to be: 1. To make a first assessment of the safety of potential KBS-3 repositories at Forsmark and Laxemar to dispose of canisters as specified in the application for the encapsulation plant. 2. To provide feedback to design development, to SKB's research and development (R and D) programme, to further site investigations and to future safety assessments. 3. To foster a dialogue with the authorities that oversee SKB's activities, i.e. the Swedish Nuclear Power Inspectorate, SKI, and the Swedish Radiation Protection Authority, SSI, regarding interpretation of applicable regulations, as a preparation for the SR-Site project. To help inform their review of SKB's proposed approach to development of the longterm safety case, the authorities appointed three international expert review teams to carry out a review of SKB's SR-Can safety assessment report. Comments from one of these teams - the Safety Assessment Methodology (SAM) review team - are presented in this document. The SAM review team's scope of work included an examination of SKB's documentation of the assessment ('Long-term safety for KBS-3 Repositories at Forsmark and Laxemar - a first evaluation' and several supporting reports) and hearings with SKB staff and contractors, held in March 2007. As directed by SKI and SSI, the SAM review team focused on methodological aspects and sought to determine whether SKB's proposed safety assessment methodology is likely to be suitable for use in the future SR-Site and to assess its consistency with the Swedish regulatory framework. No specific evaluation of long-term safety or site acceptability was undertaken by any of the review teams. SKI and SSI's Terms of Reference for the SAM review team requested that consideration be given

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

International Nuclear Information System (INIS)

2014-09-01

Strong reliance on inherent and passive design features has become a hallmark of many advanced reactor designs, including several evolutionary designs and nearly all advanced small and medium sized reactor (SMR) designs. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones — not only to enhance the operational safety of the reactors but also to eliminate the possibility of serious accidents. Accordingly, the assessment of the reliability of passive safety systems is a crucial issue to be resolved before their extensive use in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are unknown a priori. The functions of passive systems are based on basic physical laws and thermodynamic principals, and they may not experience the same kind of failures as active systems. Hence, consistent efforts are required to qualify the reliability of passive systems. To support the development of advanced nuclear reactor designs with passive systems, investigations into their reliability using various methodologies are being conducted in several Member States with advanced reactor development programmes. These efforts include reliability methods for passive systems by the French Atomic Energy and Alternative Energies Commission, reliability evaluation of passive safety system by the University of Pisa, Italy, and assessment of passive system reliability by the Bhabha Atomic Research Centre, India. These different approaches seem to demonstrate a consensus on some aspects. However, the developers of the approaches have been unable to agree on the definition of reliability in a passive system. Based on these developments and in order to foster collaboration, the IAEA initiated the Coordinated Research Project (CRP) on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors in 2008. The

Development of assessment methodology for plant configuration control

Energy Technology Data Exchange (ETDEWEB)

Cheong, Chang Hyeon; Yu, Yeong Woo; Cho, Jae Seon; Kim, Ju Yeol; Kim, Yun Ik; Yang, Hui Chang; Park, Gang Min; Hur, Byeong Gil [Seoul National Univ., Seoul (Korea, Republic of)

1999-03-15

The purpose of this study is the development of effective and overall assessment methodology which reflects the characteristics of plants for the surveillance, maintenance, repair and operation of Nuclear Power Plant. The development of this methodology can contribute to enhance safety. In the first year of this study, recent researches are surveyed and concept definition, procedures, current PSA methodologies, implementation of various models are evaluated. Through this survey, systematic assessment methodology is suggested.

A Comparison of the Safety Analysis Process and the Generation IV Proliferation Resistance/Physical Protection Assessment Methodology

International Nuclear Information System (INIS)

T. A. Bjornard; M. D. Zentner

2006-01-01

The Generation IV International Forum (GIF) is a vehicle for the cooperative international development of future nuclear energy systems. The Generation IV program has established primary objectives in the areas of sustainability, economics, safety and reliability, and Proliferation Resistance and Physical Protection (PR and PP). In order to help meet the latter objective a program was launched in December 2002 to develop a rigorous means to assess nuclear energy systems with respect to PR and PP. The study of Physical Protection of a facility is a relatively well established methodology, but an approach to evaluate the Proliferation Resistance of a nuclear fuel cycle is not. This paper will examine the Proliferation Resistance (PR) evaluation methodology being developed by the PR group, which is largely a new approach and compare it to generally accepted nuclear facility safety evaluation methodologies. Safety evaluation methods have been the subjects of decades of development and use. Further, safety design and analysis is fairly broadly understood, as well as being the subject of federally mandated procedures and requirements. It is therefore extremely instructive to compare and contrast the proposed new PR evaluation methodology process with that used in safety analysis. By so doing, instructive and useful conclusions can be derived from the comparison that will help to strengthen the PR methodological approach as it is developed further. From the comparison made in this paper it is evident that there are very strong parallels between the two processes. Most importantly, it is clear that the proliferation resistance aspects of nuclear energy systems are best considered beginning at the very outset of the design process. Only in this way can the designer identify and cost effectively incorporate intrinsic features that might be difficult to implement at some later stage. Also, just like safety, the process to implement proliferation resistance should be a dynamic

Blanket safety by GEMSAFE methodology

International Nuclear Information System (INIS)

Sawada, Tetsuo; Saito, Masaki

2001-01-01

General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

Organization and methodology approach for the safety assessment of the present situation and the future works on Chernobyl-4 and the site

International Nuclear Information System (INIS)

Bachner, D.; Benoist, E.; Duco, J.; Jahns, A.

1995-01-01

This work deals with the organization and methodology approach for the safety assessment of the present situation and the future works on Chernobyl 4 and the site. It presents the results of a common preliminary discussion in order to formulate advices on the basic management of the Chernobyl safety assessment process. (O.L.)

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

A study on a reliability assessment methodology for the VHTR safety systems

International Nuclear Information System (INIS)

Lee, Hyung Sok

2012-02-01

The passive safety system of a 300MWt VHTR (Very High Temperature Reactor)which has attracted worldwide attention recently is actively considered for designing the improvement in the safety of the next generation nuclear power plant. The passive system functionality does not rely on an external source of the electrical support system,but on an intelligent use of the natural phenomena, such as convection, conduction, radiation, and gravity. It is not easy to evaluate quantitatively the reliability of the passive safety for the risk analysis considering the existing active system failure since the classical reliability assessment method could not be applicable. Therefore a new reliability methodology needs to be developed and applied for evaluating the reliability of the conceptual designed VHTR in this study. The preliminary evaluation and conceptualization are performed using the concept of the load and capacity theory related to the reliability physics model. The method of response surface method (RSM) is also utilized for evaluating the maximum temperature of nuclear fuel in this study. The significant variables and their correlation are considered for utilizing the GAMMA+ code. The proposed method might contribute to designing the new passive system of the VHTR

Evaluation and assessment of nuclear power plant seismic methodology

International Nuclear Information System (INIS)

Bernreuter, D.; Tokarz, F.; Wight, L.; Smith, P.; Wells, J.; Barlow, R.

1977-01-01

The major emphasis of this study is to develop a methodology that can be used to assess the current methods used for assuring the seismic safety of nuclear power plants. The proposed methodology makes use of system-analysis techniques and Monte Carlo schemes. Also, in this study, we evaluate previous assessments of the current seismic-design methodology

Evaluation and assessment of nuclear power plant seismic methodology

Energy Technology Data Exchange (ETDEWEB)

Bernreuter, D.; Tokarz, F.; Wight, L.; Smith, P.; Wells, J.; Barlow, R.

1977-03-01

The major emphasis of this study is to develop a methodology that can be used to assess the current methods used for assuring the seismic safety of nuclear power plants. The proposed methodology makes use of system-analysis techniques and Monte Carlo schemes. Also, in this study, we evaluate previous assessments of the current seismic-design methodology.

Safety case methodology for decommissioning of research reactors. Assessment of the long term impact of a flooding scenario

International Nuclear Information System (INIS)

Vladescu, G.; Banciu, O.

1999-01-01

The paper contains the assessment methodology of a Safety Case fuel decommissioning of research reactors, taking into account the international approach principles. The paper also includes the assessment of a flooding scenario for a decommissioned research reactor (stage 1 of decommissioning). The scenario presents the flooding of reactor basement, radionuclide migration through environment and long term radiological impact for public. (authors)

Hydrogen safety risk assessment methodology applied to a fluidized bed membrane reactor for autothermal reforming of natural gas

NARCIS (Netherlands)

Psara, N.; Van Sint Annaland, M.; Gallucci, F.

2015-01-01

The scope of this paper is the development and implementation of a safety risk assessment methodology to highlight hazards potentially prevailing during autothermal reforming of natural gas for hydrogen production in a membrane reactor, as well as to reveal potential accidents related to hydrogen

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

A Methodology for Evaluating Quantitative Nuclear Safety Culture Impact

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

A Methodology for Evaluating Quantitative Nuclear Safety Culture Impact

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2015-01-01

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

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.

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.

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.

International Expert Review of Sr-Can: Safety Assessment Methodology - External review contribution in support of SSI's and SKI's review of SR-Can

Energy Technology Data Exchange (ETDEWEB)

Sagar, Budhi (Center for Nuclear Waste Regulatory Analyses, Southwest Research Inst., San Antonio, TX (US)); Egan, Michael (Quintessa Limited, Henley-on-Thames (GB)); Roehlig, Klaus-Juergen (Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (DE)); Chapman, Neil (Independent Consultant (XX)); Wilmot, Roger (Galson Sciences Limited, Oakham (GB))

2008-03-15

In 2006, SKB published a safety assessment (SR-Can) as part of its work to support a licence application for the construction of a final repository for spent nuclear fuel. The purposes of the SR-Can project were stated in the main project report to be: 1. To make a first assessment of the safety of potential KBS-3 repositories at Forsmark and Laxemar to dispose of canisters as specified in the application for the encapsulation plant. 2. To provide feedback to design development, to SKB's research and development (R and D) programme, to further site investigations and to future safety assessments. 3. To foster a dialogue with the authorities that oversee SKB's activities, i.e. the Swedish Nuclear Power Inspectorate, SKI, and the Swedish Radiation Protection Authority, SSI, regarding interpretation of applicable regulations, as a preparation for the SR-Site project. To help inform their review of SKB's proposed approach to development of the longterm safety case, the authorities appointed three international expert review teams to carry out a review of SKB's SR-Can safety assessment report. Comments from one of these teams - the Safety Assessment Methodology (SAM) review team - are presented in this document. The SAM review team's scope of work included an examination of SKB's documentation of the assessment ('Long-term safety for KBS-3 Repositories at Forsmark and Laxemar - a first evaluation' and several supporting reports) and hearings with SKB staff and contractors, held in March 2007. As directed by SKI and SSI, the SAM review team focused on methodological aspects and sought to determine whether SKB's proposed safety assessment methodology is likely to be suitable for use in the future SR-Site and to assess its consistency with the Swedish regulatory framework. No specific evaluation of long-term safety or site acceptability was undertaken by any of the review teams. SKI and SSI's Terms of Reference for the SAM

Evaluation of replacement tritium facility (RTF) compliance with DOE safety goals using probabilistic consequence assessment methodology

International Nuclear Information System (INIS)

O'Kula, K.R.; East, J.M.; Moore, M.L.

1993-01-01

The Savannah River Site (SRS), operated by the Westinghouse Savannah River Company (WSRC) for the US Department of Energy (DOE), is a major center for the processing of nuclear materials for national defense, deep-space exploration, and medical treatment applications in the United States. As an integral part of the DOE's effort to modernize facilities, implement improved handling and processing technology, and reduce operational risk to the general public and onsite workers, transition of tritium processing at SRS from the Consolidated Tritium Facility to the Replacement Tritium Facility (RTF) began in 1993. To ensure that operation of new DOE facilities such as RTF present minimum involuntary and voluntary risks to the neighboring public and workers, indices of risk have been established to serve as target levels or safety goals of performance for assessing nuclear safety. These goals are discussed from a historical perspective in the initial part of this paper. Secondly, methodologies to quantify risk indices are briefly described. Lastly, accident, abnormal event, and normal operation source terms from RTF are evaluated for consequence assessment purposes relative to the safety targets

CATHARE code development and assessment methodologies

International Nuclear Information System (INIS)

Micaelli, J.C.; Barre, F.; Bestion, D.

1995-01-01

The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

Safety analysis methodology for OPR 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun

2005-01-01

Full text: Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

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

Safety analysis and evaluation methodology for fusion systems

International Nuclear Information System (INIS)

Fujii-e, Y.; Kozawa, Y.; Namba, C.

1987-03-01

Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)

Safety at Work : Research Methodology

NARCIS (Netherlands)

Beurden, van K. (Karin); Boer, de J. (Johannes); Brinks, G. (Ger); Goering-Zaburnenko, T. (Tatiana); Houten, van Y. (Ynze); Teeuw, W. (Wouter)

2012-01-01

In this document, we provide the methodological background for the Safety atWork project. This document combines several project deliverables as defined inthe overall project plan: validation techniques and methods (D5.1.1), performanceindicators for safety at work (D5.1.2), personal protection

A study on assessment methodology of surveillance test interval and allowed outage time

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Huh, Chang Wook; Kim, Do Hyoung; Kim, Ju Youl; Kim, Yoon Ik; Yang, Hui Chang; Park, Kang Min [Seoul National Univ., Seoul (Korea, Republic of)

1998-03-15

The objectives of this study is the development of methodology by which assesses the optimization of Surveillance Test Internal(STI) and Allowed Outage Time(AOT) using PSA method that can supplement the current deterministic methods and the improvement of Korean nuclear power plant safety. In this study, the survey about the assessment methodologies, modelings and results performed by domestic and international researches are performed as the basic step before developing the assessment methodology of this study. The assessment methodology that supplement the revealed problems in many other studies is presented and the application of new methodology into the example system assures the feasibility of this method. The sensitivity analyses about the failure factors of the components are performed in the bases of the and AOT is quantified. And the reliability assessment methodology about the diesel generator is reviewed and applied to the PSA code. The qualitative assessment for the STI/AOR of RPS/ESFAS assured safety the most important system in the nuclear power plant are performed.

Latest developments on safety analysis methodologies at the Juzbado plant

International Nuclear Information System (INIS)

Zurron-Cifuentes, Oscar; Ortiz-Trujillo, Diego; Blanco-Fernandez, Luis A.

2010-01-01

Over the last few years the Juzbado Plant has developed and implemented several analysis methodologies to cope with specific issues regarding safety management. This paper describes the three most outstanding of them, so as to say, the Integrated Safety Analysis (ISA) project, the adaptation of the MARSSIM methodology for characterization surveys of radioactive contamination spots, and the programme for the Systematic Review of the Operational Conditions of the Safety Systems (SROCSS). Several reasons motivated the decision to implement such methodologies, such as Regulator requirements, operational experience and of course, the strong commitment of ENUSA to maintain the highest standards of nuclear industry on all the safety relevant activities. In this context, since 2004 ENUSA is undertaking the ISA project, which consists on a systematic examination of plant's processes, equipment, structures and personnel activities to ensure that all relevant hazards that could result in unacceptable consequences have been adequately evaluated and the appropriate protective measures have been identified. On the other hand and within the framework of a current programme to ensure the absence of radioactive contamination spots on unintended areas, the MARSSIM methodology is being applied as a tool to conduct the radiation surveys and investigation of potentially contaminated areas. Finally, the SROCSS programme was initiated earlier this year 2009 to assess the actual operating conditions of all the systems with safety relevance, aiming to identify either potential non-conformities or areas for improvement in order to ensure their high performance after years of operation. The following paragraphs describe the key points related to these three methodologies as well as an outline of the results obtained so far. (authors)

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.

Methodology for identifying boundaries of systems important to safety in CANDU nuclear power plants

International Nuclear Information System (INIS)

Therrien, S.; Komljenovic, D.; Therrien, P.; Ruest, C.; Prevost, P.; Vaillancourt, R.

2007-01-01

This paper presents a methodology developed to identify the boundaries of the systems important to safety (SIS) at the Gentilly-2 Nuclear Power Plant (NPP), Hydro-Quebec. The SIS boundaries identification considers nuclear safety only. Components that are not identified as important to safety are systematically identified as related to safety. A global assessment process such as WANO/INPO AP-913 'Equipment Reliability Process' will be needed to implement adequate changes in the management rules of those components. The paper depicts results in applying the methodology to the Shutdown Systems 1 and 2 (SDS 1, 2), and to the Emergency Core Cooling System (ECCS). This validation process enabled fine tuning the methodology, performing a better estimate of the effort required to evaluate a system, and identifying components important to safety of these systems. (author)

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.

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

Nuclear reactor conceptual design: methodology for cost-effective internalisation of nuclear safety

International Nuclear Information System (INIS)

Gimenez, M.; Grinblat, P.; Schlamp, M.

2002-01-01

A novel and promising methodology to perform nuclear reactor design is presented in this work. It achieves to balance efficiently safety and economics at the conceptual engineering stage. The key to this integral approach is to take into account safety aspects in a design optimisation process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behaviour during accidents and from its probabilistic safety assessment -safety performance indicators-, are synthesised on Safety Design Maps. These maps allow one to compare these indicators with limit values, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimisation process, by means of additional rules to the neutronic, thermal-hydraulic and mechanical calculations. This methodology turns out to be promising to balance and optimise reactor and safety system design in an early engineering stage, in order to internalise cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels. Furthermore, through this methodology, a simplified design can be obtained, compared to the resultant complexity when these concepts are introduced in a later engineering stage. (author)

Application of Bow-tie methodology to improve patient safety.

Science.gov (United States)

Abdi, Zhaleh; Ravaghi, Hamid; Abbasi, Mohsen; Delgoshaei, Bahram; Esfandiari, Somayeh

2016-05-09

Purpose - The purpose of this paper is to apply Bow-tie methodology, a proactive risk assessment technique based on systemic approach, for prospective analysis of the risks threatening patient safety in intensive care unit (ICU). Design/methodology/approach - Bow-tie methodology was used to manage clinical risks threatening patient safety by a multidisciplinary team in the ICU. The Bow-tie analysis was conducted on incidents related to high-alert medications, ventilator associated pneumonia, catheter-related blood stream infection, urinary tract infection, and unwanted extubation. Findings - In total, 48 potential adverse events were analysed. The causal factors were identified and classified into relevant categories. The number and effectiveness of existing preventive and protective barriers were examined for each potential adverse event. The adverse events were evaluated according to the risk criteria and a set of interventions were proposed with the aim of improving the existing barriers or implementing new barriers. A number of recommendations were implemented in the ICU, while considering their feasibility. Originality/value - The application of Bow-tie methodology led to practical recommendations to eliminate or control the hazards identified. It also contributed to better understanding of hazard prevention and protection required for safe operations in clinical settings.

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)

Application of a safety assessment methodology to a hypothetical surface disposal at Serpong site, Indonesia

International Nuclear Information System (INIS)

Lubis, E.; Mallants, D.; Volckaert, G.; Marivoet, J.; Neerdael, B.

2000-01-01

A preliminary and generic safety assessment of a candidate shallow land burial (SLB) repository at Serpong site, Indonesia, has been performed. The step-by-step safety assessment methodology included an analysis of features, events, and processes (FEPs), and mathematical modelling of radionuclide migration in the near field, geosphere and biosphere. On the basis of an extensive FEP catalogue the most relevant scenarios to be considered in the consequence analysis were selected. Both the normal evolution scenario (NES) and the alternative scenarios were identified. On the basis of these scenarios a conceptual model that included all the important physical-chemical processes was built for the near field and geosphere. A two-dimensional numerical model was then used to solve the governing flow and transport equations for appropriate initial and boundary conditions. The calculations were performed using a repository-specific value for the total disposed activity in combination with hypothetical values for radionuclide composition based on a typical radionuclide content of low level waste in Belgium. Site-specific data on hydrogeological properties were used for the geosphere calculations. Typical results of the consequence analysis in terms of radionuclide fluxes to the geosphere and radionuclide concentrations in the groundwater are discussed. (author)

Application of the Biosphere Assessment Methodology to the ENRESA, 1997 Performance and Safety Assessment

International Nuclear Information System (INIS)

Pinedo, P.; Simon, I.; Aguero, A.

1998-01-01

For several years CIEMAT has been developing for ENRESA knowledge and tools to support the modelling of the migration and accumulation of radionuclides within the biosphere once those radionuclides are released or reach one or more parts of the biosphere (atmosphere, water bodies or soils). The model development also includes evaluation of radiological impacts arising from the resulting distribution of radionuclides in the biosphere. In 1996, a Methodology to analyse the biosphere in this context proposed to ENRESA. The level of development of the different aspects proposed within the Methodology was quite heterogeneous and, while aspects of radionuclide transport modelling were already well developed in theoretical and practical terms, other aspects like the procedure for conceptual model development and the description of biosphere system representatives of the long term needed further developments. At present, the International Atomic Energy Agency (IAEA) Programme on Biosphere Modelling and Assessment (BIOMASS) in collaboration with several national organizations, ENRESA and CIEMAT among them, is working to complete and augment the Reference Biosphere Methodology and to produce some practical descriptions of Reference Systems. The overall purpose of this document is to apply the Methodology, taking account of on-going developments in biosphere modelling, to the last performance assessment (PA) exercise made by ENRESA (ENRESA, 1997), using from it the general and particular information about the assessment context, radionuclide information, geosphere and geobiosphere interface data. There are three particular objectives to this work: (a) to determine the practicability of the Methodology in an application to a realistic assessment situation, (b) To compare and contrast previous biosphere modelling in HLW PA and, (c) to test software development related with data management and modelling. (Author) 42 refs

Assessment methodology applicable to safe decommissioning of Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Baniu, O.; Vladescu, G.; Vidican, D.; Penescu, M.

2002-01-01

The paper contains the results of research activity performed by CITON specialists regarding the assessment methodology intended to be applied to safe decommissioning of the research reactors, developed taking into account specific conditions of the Romanian VVR-S Research Reactor. The Romanian VVR-S Research Reactor is an old reactor (1957) and its Decommissioning Plan is under study. The main topics of paper are as follows: Safety approach of nuclear facilities decommissioning. Applicable safety principles; Main steps of the proposed assessment methodology; Generic content of Decommissioning Plan. Main decommissioning activities. Discussion about the proposed Decommissioning Plan for Romanian Research Reactor; Safety risks which may occur during decommissioning activities. Normal decommissioning operations. Fault conditions. Internal and external hazards; Typical development of a scenario. Features, Events and Processes List. Exposure pathways. Calculation methodology. (author)

Development of risk-informed assessment (RIA) design methodology

International Nuclear Information System (INIS)

Ji, S. K.; Park, S. J.; Park, B. R.; Kim, M. R.; Choi, C. J.

2001-01-01

It has been assessed that the capital cost for future nuclear power plants needs to be reduced on the order of 35% to 40% for Advanced Light Water Reactors such as KNGR and System 80+. Such reduction in the capital cost will require a fundamental re-evaluation of the industry standards and regulatory basis under which nuclear plants are designed and licensed. The objective of this study is to develop the risk-informed assessment (RIA) design methodology for future nuclear power plants. In order to meet this objective, the design simplification method is developed and RIA design methodology exercised for conceptual system. For the methodology verification, simplified conceptual ECCS and feedwater system are developed, then LOCA sensitivity analyses and agressive secondary cooldown analyses for these systems are performed. In addition, the probability safety assessment (PSA) model for LOCA is developed and the validation of RIA design methodology is demonstrated

Assessment of ALWR passive safety system reliability. Phase 1: Methodology development and component failure quantification

International Nuclear Information System (INIS)

Hake, T.M.; Heger, A.S.

1995-04-01

Many advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive systems to perform safety functions, rather than active systems as in current reactor designs. These passive systems depend to a great extent on physical processes such as natural circulation for their driving force, and not on active components, such as pumps. An NRC-sponsored study was begun at Sandia National Laboratories to develop and implement a methodology for evaluating ALWR passive system reliability in the context of probabilistic risk assessment (PRA). This report documents the first of three phases of this study, including methodology development, system-level qualitative analysis, and sequence-level component failure quantification. The methodology developed addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. Traditional PRA methods, such as fault and event tree modeling, are applied to the component failure aspect. Thermal-hydraulic calculations are incorporated into a formal expert judgment process to address uncertainties in selected natural processes and success criteria. The first phase of the program has emphasized the component failure element of passive system reliability, rather than the natural process uncertainties. Although cursory evaluation of the natural processes has been performed as part of Phase 1, detailed assessment of these processes will take place during Phases 2 and 3 of the program

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

Status of IAEA CRPI31018 “Development of Methodologies for the Assessment of Passive Safety System Performance in Advanced Reactors”

International Nuclear Information System (INIS)

Subki, Hadid M.

2011-01-01

Purpose of research coordination meeting: • To review progress and milestones on all research activities; • To discuss the preliminary experimental data obtained from the Natural Circulation Loop Facility L2 in Italy constructed for the assessment of different methodologies for the evaluation of the reliability of passive safety system; • To discuss lessons-to be-learned from the Fukushima Daiichi Accident in Japan and its implications to near future R&D needs on thermal-hydraulics and reactor safety; • To develop an outline of integrated annual technical report and future collaboration plan

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)

Geosphere process report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Skagius, Kristina

2010-11-01

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

Geosphere process report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

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

2010-11-15

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

Methodology of safety assessment and sensitivity analysis for geologic disposal of high-level radioactive waste

International Nuclear Information System (INIS)

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

1995-01-01

A deterministic safety assessment methodology has been developed to evaluate long-term radiological consequences associated with geologic disposal of high-level radioactive waste, and to demonstrate a generic feasibility of geologic disposal. An exposure scenario considered here is based on a normal evolution scenario which excludes events attributable to probabilistic alterations in the environment. A computer code system GSRW thus developed is based on a non site-specific model, and consists of a set of sub-modules for calculating the release of radionuclides from engineered barriers, the transport of radionuclides in and through the geosphere, the behavior of radionuclides in the biosphere, and radiation exposures of the public. In order to identify the important parameters of the assessment models, an automated procedure for sensitivity analysis based on the Differential Algebra method has been developed to apply to the GSRW. (author)

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

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)

Development of performance assessment methodology for establishment of quantitative acceptance criteria of near-surface radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Kim, C. R.; Lee, E. Y.; Park, J. W.; Chang, G. M.; Park, H. Y.; Yeom, Y. S. [Korea Hydro and Nuclear Power Co., Ltd., Seoul (Korea, Republic of)

2002-03-15

The contents and the scope of this study are as follows : review of state-of-the-art on the establishment of waste acceptance criteria in foreign near-surface radioactive waste disposal facilities, investigation of radiological assessment methodologies and scenarios, investigation of existing models and computer codes used in performance/safety assessment, development of a performance assessment methodology(draft) to derive quantitatively radionuclide acceptance criteria of domestic near-surface disposal facility, preliminary performance/safety assessment in accordance with the developed methodology.

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)

Review of SKB's interim report of SR-Can: SKI's and SSI's evaluation of SKB's up-dated methodology for safety assessment

Energy Technology Data Exchange (ETDEWEB)

Dverstorp, Bjoern; Moberg, Leif; Wiebert, Anders; Xu Shulan [Swedish Radiation Protection Authority, Stockholm (Sweden); Stroemberg, Bo; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Sundstroem, Benny; Toverud, Oeivind [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

2005-07-01

This report presents the findings of a review of the Swedish Nuclear Fuel and Waste Management Co.'s (SKB) interim report of the safety assessment SR-Can (SKB TR 04-11), conducted by the Swedish Radiation Protection Authority (SSI) and the Swedish Nuclear Power Inspectorate (SKI). SKB's interim report describes and exemplifies the safety assessment methodology that SKB plans to use in the oncoming licence applications for an encapsulation plant and a final repository for spent nuclear fuel. The authorities' review takes into account the findings of an international peer review of SKB's interim report. The authorities conclude that SKB has improved its safety assessment methodology in several aspects compared to earlier safety reports. Among other things the authorities commend SKB for giving a comprehensive account of relevant regulations and guidance, and for the systematic approach to identification and documentation of features, events and processes that need to be considered in the safety assessment. However, the authorities also conclude that important parts of SKB's method need to be further developed before they are mature enough to be used as a basis for a license application. The authorities' overall assessment is summarised in chapter 8 of this report.

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.

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

Methodology for assessing the safety of Hydrogen Systems: HyRAM 1.1 technical reference manual

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina; Hecht, Ethan; Reynolds, John Thomas; Blaylock, Myra L.; Erin E. Carrier

2017-03-01

The HyRAM software toolkit provides a basis for conducting quantitative risk assessment and consequence modeling for hydrogen infrastructure and transportation systems. HyRAM is designed to facilitate the use of state-of-the-art science and engineering models to conduct robust, repeatable assessments of hydrogen safety, hazards, and risk. HyRAM is envisioned as a unifying platform combining validated, analytical models of hydrogen behavior, a stan- dardized, transparent QRA approach, and engineering models and generic data for hydrogen installations. HyRAM is being developed at Sandia National Laboratories for the U. S. De- partment of Energy to increase access to technical data about hydrogen safety and to enable the use of that data to support development and revision of national and international codes and standards. This document provides a description of the methodology and models contained in the HyRAM version 1.1. HyRAM 1.1 includes generic probabilities for hydrogen equipment fail- ures, probabilistic models for the impact of heat flux on humans and structures, and computa- tionally and experimentally validated analytical and first order models of hydrogen release and flame physics. HyRAM 1.1 integrates deterministic and probabilistic models for quantifying accident scenarios, predicting physical effects, and characterizing hydrogen hazards (thermal effects from jet fires, overpressure effects from deflagrations), and assessing impact on people and structures. HyRAM is a prototype software in active development and thus the models and data may change. This report will be updated at appropriate developmental intervals.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

A study on assessment methodology of surveillance test interval and allowed outage time

International Nuclear Information System (INIS)

Che, Moo Seong; Cheong, Chang Hyeon; Lee, Byeong Cheol

1996-07-01

The objectives of this study is the development of methodology by which assessing the optimizes Surveillance Test Interval(STI) and Allowed Outage Time(AOT) using PSA method that can supplement the current deterministic methods and the improvement of Korea nuclear power plants safety. In the first year of this study, the survey about the assessment methodologies, modeling and results performed by domestic and international researches is performed as the basic step before developing the assessment methodology of this study. The assessment methodology that supplement the revealed problems in many other studies is presented and the application of new methodology into the example system assures the feasibility of this method

A study on assessment methodology of surveillance test interval and allowed outage time

Energy Technology Data Exchange (ETDEWEB)

Che, Moo Seong; Cheong, Chang Hyeon; Lee, Byeong Cheol [Seoul Nationl Univ., Seoul (Korea, Republic of)] (and others)

1996-07-15

The objectives of this study is the development of methodology by which assessing the optimizes Surveillance Test Interval(STI) and Allowed Outage Time(AOT) using PSA method that can supplement the current deterministic methods and the improvement of Korea nuclear power plants safety. In the first year of this study, the survey about the assessment methodologies, modeling and results performed by domestic and international researches is performed as the basic step before developing the assessment methodology of this study. The assessment methodology that supplement the revealed problems in many other studies is presented and the application of new methodology into the example system assures the feasibility of this method.

UPC Scaling-up methodology for Deterministic Safety Assessment and Support to Plant Operation

Energy Technology Data Exchange (ETDEWEB)

Martínez-Quiroga, V.; Reventós, F.; Batet, Il.

2015-07-01

Best Estimate codes along with necessary nodalizations are widely used tools in nuclear engineering for both Deterministic Safety Assessment (DSA) and Support to Plant Operation and Control. In this framework, the application of quality assurance procedures in both codes and nodalizations becomes an essential step prior any significant study. Along these lines the present paper introduces the UPC SCUP, a systematic methodology based on the extrapolation of the Integral Test Facilities (ITF) post-test simulations by means of scaling analyses. In that sense, SCUP fulfills a gap in current nodalization qualification procedures, the related with the validation of NPP nodalizations for Design Basis Accidents conditions. Three are the pillars that support SCUP: judicial selection of the experimental transients, full confidence in the quality of the ITF simulations, and simplicity in justifying discrepancies that appear between ITF and NPP counterpart transients. The techniques that are presented include the socalled Kv scaled calculations as well as the use of two new approaches, ”Hybrid nodalizations” and ”Scaled-up nodalizations”. These last two methods have revealed themselves to be very helpful in producing the required qualification and in promoting further improvements in nodalization. The study of both LSTF and PKL counterpart tests have allowed to qualify the methodology by the comparison with experimental data. Post-test simulations at different sizes allowed to define which phenomena could be well reproduced by system codes and which not, in this way also establishing the basis for the extrapolation to an NPP scaled calculation. Furthermore, the application of the UPC SCUP methodology demonstrated that selected phenomena can be scaled-up and explained between counterpart simulations by carefully considering the differences in scale and design. (Author)

Safety analysis methodology with assessment of the impact of the prediction errors of relevant parameters

International Nuclear Information System (INIS)

Galia, A.V.

2011-01-01

The best estimate plus uncertainty approach (BEAU) requires the use of extensive resources and therefore it is usually applied for cases in which the available safety margin obtained with a conservative methodology can be questioned. Outside the BEAU methodology, there is not a clear approach on how to deal with the issue of considering the uncertainties resulting from prediction errors in the safety analyses performed for licensing submissions. However, the regulatory document RD-310 mentions that the analysis method shall account for uncertainties in the analysis data and models. A possible approach is presented, that is simple and reasonable, representing just the author's views, to take into account the impact of prediction errors and other uncertainties when performing safety analysis in line with regulatory requirements. The approach proposes taking into account the prediction error of relevant parameters. Relevant parameters would be those plant parameters that are surveyed and are used to initiate the action of a mitigating system or those that are representative of the most challenging phenomena for the integrity of a fission barrier. Examples of the application of the methodology are presented involving a comparison between the results with the new approach and a best estimate calculation during the blowdown phase for two small breaks in a generic CANDU 6 station. The calculations are performed with the CATHENA computer code. (author)

Safety of High Speed Ground Transportation Systems : Analytical Methodology for Safety Validation of Computer Controlled Subsystems : Volume 2. Development of a Safety Validation Methodology

Science.gov (United States)

1995-01-01

This report describes the development of a methodology designed to assure that a sufficiently high level of safety is achieved and maintained in computer-based systems which perform safety cortical functions in high-speed rail or magnetic levitation ...

Obtention to the methodology for evaluation to the confirmation of the hazardous wastes safety isolation

International Nuclear Information System (INIS)

Peralta, J.L.; Gil, R.; Castillo, R.; Leyva, D.

2003-01-01

Taking into account, the practical experience of the safety assessment in the radioactive wastes management, the International Atomic Energy Agency (IAEA) recommendations in this topics, the norms and national and international legislation about noxious substances to the environment and their restriction limits, the best international practices and approaches of isolation hazardous wastes sites, a Methodology is developed (Cuba particular conditions) to obtaining and/or confirmation of the hazardous wastes safety isolation, as a tool able to carry out the assessment of facilities to build and all installation and/or place where hazardous wastes isolated from the environment. The Methodology, embraces the evaluation of technical, economic and social topics, allowing to develop an integral safety assessment which allows to estimate the environment possible impact for hazardous waste isolation (radioactive and non radioactive); Just are shown in this paper the selection approaches for the obtaining and/or evaluation of the best site, the steps description to continue for the definition of the main scenarios and the models to take into account in the valuation of the possible liberation and pathway to the environment of the non radioactive pollutants. The main contribution of this Methodology resides in the creation of a scientific-technique necessary guide for the evident demand of carrying out the most organized, effective and hazardous wastes safety management

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)

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

Legal basis for risk analysis methodology while ensuring food safety in the Eurasian Economic union and the Republic of Belarus

Directory of Open Access Journals (Sweden)

E.V. Fedorenko

2015-09-01

Full Text Available Health risk analysis methodology is an internationally recognized tool for ensuring food safety. Three main elements of risk analysis are risk assessment, risk management and risk communication to inform the interested parties on the risk, are legislated and implemented in the Eurasian Economic Union and the Republic of Belarus. There is a corresponding organizational and functional framework for the application of risk analysis methodology as in the justification of production safety indicators and the implementation of public health surveillance. Common methodological approaches and criteria for evaluating public health risk are determined, which are used in the development and application of food safety requirements. Risk assessment can be used in justifying the indicators of safety (contaminants, food additives, and evaluating the effectiveness of programs on enrichment of food with micronutrients.

A study on assessment methodology of surveillance test interval and Allowed Outage Time

Energy Technology Data Exchange (ETDEWEB)

Che, Moo Seong; Cheong, Chang Hyeon; Ryu, Yeong Woo; Cho, Jae Seon; Heo, Chang Wook; Kim, Do Hyeong; Kim, Joo Yeol; Kim, Yun Ik; Yang, Hei Chang [Seoul National Univ., Seoul (Korea, Republic of)

1997-07-15

Objectives of this study is the development of methodology by which assesses the optimization of Surveillance Test Interval(STI) and Allowed Outage Time(AOT) using PSA method that can supplement the current deterministic methods and the improvement of Korean nuclear power plants safety. In the first year of this study, the survey about the assessment methodologies, modeling and results performed by domestic and international researches are performed as the basic step before developing the assessment methodology of this study. The assessment methodology that supplement the revealed problems in many other studies is presented and the application of new methodology into the example system assures the feasibility of this method. In the second year of this study, the sensitivity analyses about the failure factors of the components are performed in the bases of the assessment methodologies of the first study, the interaction modeling of the STI and AOT is quantified. And the reliability assessment methodology about the diesel generator is reviewed and applied to the PSA code.

A study on assessment methodology of surveillance test interval and Allowed Outage Time

International Nuclear Information System (INIS)

Che, Moo Seong; Cheong, Chang Hyeon; Ryu, Yeong Woo; Cho, Jae Seon; Heo, Chang Wook; Kim, Do Hyeong; Kim, Joo Yeol; Kim, Yun Ik; Yang, Hei Chang

1997-07-01

Objectives of this study is the development of methodology by which assesses the optimization of Surveillance Test Interval(STI) and Allowed Outage Time(AOT) using PSA method that can supplement the current deterministic methods and the improvement of Korean nuclear power plants safety. In the first year of this study, the survey about the assessment methodologies, modeling and results performed by domestic and international researches are performed as the basic step before developing the assessment methodology of this study. The assessment methodology that supplement the revealed problems in many other studies is presented and the application of new methodology into the example system assures the feasibility of this method. In the second year of this study, the sensitivity analyses about the failure factors of the components are performed in the bases of the assessment methodologies of the first study, the interaction modeling of the STI and AOT is quantified. And the reliability assessment methodology about the diesel generator is reviewed and applied to the PSA code

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)

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/

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/

Combining soft system methodology and pareto analysis in safety management performance assessment : an aviation case

NARCIS (Netherlands)

Karanikas, Nektarios

2016-01-01

Although reengineering is strategically advantageous for organisations in order to keep functional and sustainable, safety must remain a priority and respective efforts need to be maintained. This paper suggests the combination of soft system methodology (SSM) and Pareto analysis on the scope of

Regulatory review and confidence building in post-closure safety assessments and safety cases for near surface disposal facilities-IAEA ASAM coordinated research programme

International Nuclear Information System (INIS)

Gonzales, A.; Simeonov, G.; Bennett, D.G.; Nys, V.; Ben Belfadhel, M.

2005-01-01

Some years ago, the IAEA successfully concluded a Coordinated Research Program (CRP) called Islam, which focussed on the development of an Improved Safety Assessment Methodology for near-surface radioactive waste disposal facilities. In November 2002, and as an extension of ISAM, the IAEA launched a new CRP called ASAM, designed to test the Application of the Safety Assessment Methodology by considering a range of near-surface disposal facilities. The ASAM work programme is being implemented by three application working groups and two cross-cutting working groups. The application working groups are testing the applicability of the ISAM methodology by assessing an existing disposal facility in Hungary, a copper mine in South Africa, and a hypothetical facility containing heterogenous wastes, such as disused sealed sources. The first cross-cutting working group is addressing a number of technical issues that are common to all near-surface disposal facilities, while the second group, the Regulatory Review Working Group (RRWG) is developing guidance on how to gain confidence in safety assessments and safety cases, and on how to conduct regulatory reviews of safety assessments. This paper provides a brief overview of the work being conducted by the Regulatory Review Working Group. (author)

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

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

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

Safety analysis methodologies for radioactive waste repositories in shallow ground

International Nuclear Information System (INIS)

1984-01-01

The report is part of the IAEA Safety Series and is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of shallow ground radioactive waste repositories. It discusses approaches that are applicable for safety analysis of a shallow ground repository. The methodologies, analysis techniques and models described are pertinent to the task of predicting the long-term performance of a shallow ground disposal system. They may be used during the processes of selection, confirmation and licensing of new sites and disposal systems or to evaluate the long-term consequences in the post-sealing phase of existing operating or inactive sites. The analysis may point out need for remedial action, or provide information to be used in deciding on the duration of surveillance. Safety analysis both general in nature and specific to a certain repository, site or design concept, are discussed, with emphasis on deterministic and probabilistic studies

Seismic reliability assessment methodology for CANDU concrete containment structures

International Nuclear Information System (INIS)

Stephens, M.J.; Nessim, M.A.; Hong, H.P.

1995-05-01

A study was undertaken to develop a reliability-based methodology for the assessment of existing CANDU concrete containment structures with respect to seismic loading. The focus of the study was on defining appropriate specified values and partial safety factors for earthquake loading and resistance parameters. Key issues addressed in the work were the identification of an approach to select design earthquake spectra that satisfy consistent safety levels, and the use of structure-specific data in the evaluation of structural resistance. (author). 23 refs., 9 tabs., 15 figs

MAPLE research reactor safety uncertainty assessment methodology

International Nuclear Information System (INIS)

Sills, H.E.; Duffey, R.B.; Andres, T.H.

1999-01-01

The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

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.)

Contextual assessment of organisational culture - methodological development in two case studies

International Nuclear Information System (INIS)

Reiman, T.; Oedewald, P.

2002-01-01

Despite the acknowledged significance of organisational culture in the nuclear field, previous cultural studies have concentrated on purely safety related matters, or been only descriptive in nature. New kinds of methods, taking into account the overall objectives of the organisation, were needed to assess culture and develop its working practices appropriately. VTT developed the Contextual Assessment of Organisational Culture (CAOC) methodology during the FINNUS programme. The methodology utilises two concepts, organisational culture and core task. The core task can be defined as the core demands and content of work that the organisation has to accomplish in order to be effective. The core task concept is used in assessing the central dimensions of the organisation's culture. Organisational culture is defined as a solution the company has generated in order to fulfil the perceived demands of its core task. The CAOC-methodology was applied in two case studies, in the Radiation and Nuclear Safety Authority of Finland and in the maintenance unit of Loviisa NPP. The aim of the studies was not only to assess the given culture, but also to give the personnel new concepts and new tools for reflecting on their organisation, their jobs and on appropriate working practices. The CAOC-methodology contributes to the design and redesign of work in complex sociotechnical systems. It strives to enhance organisations' capability to assess their current working practices and the meanings attached to them and compare these to the actual demands of their basic mission and so change unadaptive practices. (orig.)

Methodology for the application of probabilistic safety assessment techniques (PSA) to the cobalt-therapy units in Cuba

International Nuclear Information System (INIS)

Vilaragut Llanes, J.J.; Ferro Fernandez, R.; Troncoso Fleitas, M.; Lozano Lima, B.; Fuente Puch, A. de la; Perez Reyes, Y.; Dumenigo Gonzalez, C.

2001-01-01

The applications of PSA techniques in the nuclear power plants during the last two decades and the positive results obtained for decision making in relation with safety, as a complement to deterministic methods, have increased their use in the rest of the nuclear applications. At present a large set of documents from international institutions can be found summarizing the investigations carried out in this field and promoting their use in radioactive facilities. Although still without a mandatory character, the new regulations on radiological safety also promote the complete or partial application of the PSA techniques in the safety assessment of the radiological practices. Also the IAEA, through various programs in which Cuba has been inserted, is taking a group of actions so that the nuclear community will encourage the application of the probabilistic risk methods for the evaluations and decision making with respect to safety. However, the fact that in no radioactive installation has a complete PSA study been carried out, makes that certain methodological aspects require to be improved and modified for the application of these techniques. This work presents the main elements for the use of PSA in the evaluation of the safety of cobalt-therapy units in Cuba. Also presented, as part of the results of the first stage of the Study, are the Guidelines that are being applied in a Research Contract with the Agency by the authors themselves, who belong to the CNSN, together with other specialists from the Cuban Ministry of Public Health. (author) [es

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Quantitative risk assessment of digitalized safety systems

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Safety assessment and geosphere transport methodology for the geologic isolation of nuclear waste materials

International Nuclear Information System (INIS)

Burkholder, H.C.; Stottlemyre, J.A.; Raymond, J.R.

1977-01-01

As part of the National Waste Terminal Storage Program in the United States, the Waste Isolation Safety Assessment Program (WISAP) is underway to develop and demonstrate the methods and obtain the data necessary to assess the safety of geologic isolation repositories and to communicate the assessment results to the public. This paper reviews past analysis efforts, discusses the WISAP technical approach to the problem, and points out areas where work is needed

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)

Flammability Assessment Methodology Program Phase I: Final Report

Energy Technology Data Exchange (ETDEWEB)

C. A. Loehr; S. M. Djordjevic; K. J. Liekhus; M. J. Connolly

1997-09-01

The Flammability Assessment Methodology Program (FAMP) was established to investigate the flammability of gas mixtures found in transuranic (TRU) waste containers. The FAMP results provide a basis for increasing the permissible concentrations of flammable volatile organic compounds (VOCs) in TRU waste containers. The FAMP results will be used to modify the ''Safety Analysis Report for the TRUPACT-II Shipping Package'' (TRUPACT-II SARP) upon acceptance of the methodology by the Nuclear Regulatory Commission. Implementation of the methodology would substantially increase the number of drums that can be shipped to the Waste Isolation Pilot Plant (WIPP) without repackaging or treatment. Central to the program was experimental testing and modeling to predict the gas mixture lower explosive limit (MLEL) of gases observed in TRU waste containers. The experimental data supported selection of an MLEL model that was used in constructing screening limits for flammable VOC and flammable gas concentrations. The MLEL values predicted by the model for individual drums will be utilized to assess flammability for drums that do not meet the screening criteria. Finally, the predicted MLEL values will be used to derive acceptable gas generation rates, decay heat limits, and aspiration time requirements for drums that do not pass the screening limits. The results of the program demonstrate that an increased number of waste containers can be shipped to WIPP within the flammability safety envelope established in the TRUPACT-II SARP.

Selection methodology for LWR safety programs and proposals. Volume 2. Methodology application

International Nuclear Information System (INIS)

Ritzman, R.L.; Husseiny, A.A.

1980-08-01

The results of work done to update and apply a methodology for selecting (prioritizing) LWR safety technology R and D programs are described. The methodology is based on multiattribute utility (MAU) theory. Application of the methodology to rank-order a group of specific R and D programs included development of a complete set of attribute utility functions, specification of individual attribute scaling constants, and refinement and use of an interactive computer program (MAUP) to process decision-maker inputs and generate overall (multiattribute) program utility values. The output results from several decision-makers are examined for consistency and conclusions and recommendations regarding general use of the methodology are presented. 3 figures, 18 tables

Risk-assessment methodology for fast breeder reactors

International Nuclear Information System (INIS)

Ott, K.O.

1976-04-01

The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed

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

The methodological seminar “Psychological Safety in Transport”

Directory of Open Access Journals (Sweden)

Sviridenko I.N.

2018-03-01

Full Text Available This paper provides a short brief overview of the methodological seminar “Psychological Safety in Transport” organized in Yekaterinburg on the 17th November 2017. This seminar consisted of the plenary session and four workshops focused on analyzing most important issues of Human Factor of Road Safety.

Interim main report of the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan (ed.) [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

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

Current status and future expectation concerning probabilistic risk assessment of NPPs. 1. Features and issues of probabilistic risk assessment methodology

International Nuclear Information System (INIS)

Yamashita, Masahiro

2012-01-01

Probabilistic risk assessment (PRA) of Nuclear Power Plants (NPPs) could play an important role in assuring safety of NPPs. However PRA had not always effectively used, which was indicated in Japanese government's report on Fukushima Daiichi NPP accident. At the Risk Technical Committee (RTC) of Standards Committee of Atomic Energy Society of Japan, preparation of standards (implementing criteria) focusing on PRA methodology and investigation on basic philosophy for use of PRA had been in progress. Based on activities of RTC, a serial in three articles including this described current status and future expectation concerning probabilistic risk assessment of NPPs. This article introduced features and issues of PRA methodology related to the use of PRA. Features of PRA methodology could be shown as (1) systematic and comprehensive understanding of risk, (2) support of grading approach, (3) identification of effective safety upgrade measures and (4) quantitative understanding of effects of uncertainty. Issues of PRA methodology were (1) extension of PRA application area, (2) upgrade of PRA methodology, (3) quality assurance of PRA, (4) treatment of uncertainty and (5) quantitative evaluation criteria. (T. Tanaka)

Methodology and results of the seismic probabilistic safety assessment of Krsko nuclear power plant

International Nuclear Information System (INIS)

Vermaut, M.K.; Monette, P.; Campbell, R.D.

1995-01-01

A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Krsko plant. The methodology adopted is the seismic PSA (Probabilistic Safety Assessment). The Krsko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of the site hazard, calculation of plant structures response including soil-structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Also relay chatter analysis and soil stability studies were performed. The seismic PSA described here is limited to the analysis of CDF (level I PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Krsko NPP but are not further described in this paper. The results of the seismic PSA study indicate that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable to that of most US and Western Europe NPPs. (author)

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.)

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)

Radiological risk assessment for the public under the loss of medium and large sources using bayesian methodology

International Nuclear Information System (INIS)

Kim, Joo Yeon; Jang, Han Ki; Lee, Jai Ki

2005-01-01

Bayesian methodology is appropriated for use in PRA because subjective knowledges as well as objective data are applied to assessment. In this study, radiological risk based on Bayesian methodology is assessed for the loss of source in field radiography. The exposure scenario for the lost source presented in U.S. NRC is reconstructed by considering the domestic situation and Bayes theorem is applied to updating of failure probabilities of safety functions. In case of updating of failure probabilities, it shows that 5% Bayes credible intervals using Jeffreys prior distribution are lower than ones using vague prior distribution. It is noted that Jeffreys prior distribution is appropriated in risk assessment for systems having very low failure probabilities. And, it shows that the mean of the expected annual dose for the public based on Bayesian methodology is higher than the dose based on classical methodology because the means of the updated probabilities are higher than classical probabilities. The database for radiological risk assessment are sparse in domestic. It summarizes that Bayesian methodology can be applied as an useful alternative for risk assessment and the study on risk assessment will be contributed to risk-informed regulation in the field of radiation safety

Reliability assessment for safety critical systems by statistical random testing

International Nuclear Information System (INIS)

Mills, S.E.

1995-11-01

In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs

Reliability assessment for safety critical systems by statistical random testing

Energy Technology Data Exchange (ETDEWEB)

Mills, S E [Carleton Univ., Ottawa, ON (Canada). Statistical Consulting Centre

1995-11-01

In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs.

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)

Accidental safety analysis methodology development in decommission of the nuclear facility

Energy Technology Data Exchange (ETDEWEB)

Park, G. H.; Hwang, J. H.; Jae, M. S.; Seong, J. H.; Shin, S. H.; Cheong, S. J.; Pae, J. H.; Ang, G. R.; Lee, J. U. [Seoul National Univ., Seoul (Korea, Republic of)

2002-03-15

Decontamination and Decommissioning (D and D) of a nuclear reactor cost about 20% of construction expense and production of nuclear wastes during decommissioning makes environmental issues. Decommissioning of a nuclear reactor in Korea is in a just beginning stage, lacking clear standards and regulations for decommissioning. This work accident safety analysis in decommissioning of the nuclear facility can be a solid ground for the standards and regulations. For source term analysis for Kori-1 reactor vessel, MCNP/ORIGEN calculation methodology was applied. The activity of each important nuclide in the vessel was estimated at a time after 2008, the year Kori-1 plant is supposed to be decommissioned. And a methodology for risk analysis assessment in decommissioning was developed.

Third (3rd) Research Coordination Meeting of the CRP on Development of Methodologies for the Assessment of Passive Safety System Performance in Advanced Reactors. Presentations

International Nuclear Information System (INIS)

2011-01-01

Purpose of the meeting: • To review progress and milestones on all research activities; • To discuss the preliminary experimental data obtained from the Natural Circulation Loop Facility L2 in Italy constructed for the assessment of different methodologies for the evaluation of the reliability of passive safety system; • To discuss lessons-to be-learned from the Fukushima Daiichi Accident in Japan and its implications to near future R&D needs on thermal-hydraulics and reactor safety; • To develop an outline of integrated annual technical report and future collaboration plan

Formal safety assessment based on relative risks model in ship navigation

Energy Technology Data Exchange (ETDEWEB)

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

2007-03-15

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

Formal safety assessment based on relative risks model in ship navigation

International Nuclear Information System (INIS)

Hu Shenping; Fang Quangen; Xia Haibo; Xi Yongtao

2007-01-01

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

Inclusion of Premeditated Threats in the Safety Methodology for NPPs

International Nuclear Information System (INIS)

Levanon, I.

2014-01-01

During the last decade the global effort to prevent terrorism or to mitigate its harm, if prevention fails, has increased. The nuclear power community was involved in this effort trying to prevent terrorist attacks on NPPs (Nuclear Power Plants). A natural extension of terror restraining is the prevention of any premeditated damage to the plant, including acts of state. The pre-feasibility study of an Israeli NPP, conducted by the Ministry of National Infrastructures, has identified the risk of hostile damage to the NPP as a major obstacle to the establishment of nuclear power in Israel, second only to the refusal of nuclear exporting nations to sell an NPP to Israelv. The General Director of the Ministry and the Head of the IAEC (Israeli Atomic Energy Commission) have approved continuation of the pre-feasibility study. This synopsis presents a study, regarding premeditated threats to NPPs, commissioned by the Ministry of National Infrastructures as part of the continuation. It focuses on the safety aspect of premeditated threats originating outside the plant, although a significant part of the analysis can be extended to other subjects such as theft or diversion of strategic materials. The study deals only with methodology and does not encompass specific threats or protection measures. Conclusions and recommendations and marked by bold italics Arial font. The theory of nuclear safety regarding non-premeditated safety events (equipment failures, human errors, natural events, etc.) is well developed. The study refers to these events and the theory attached to them as c lassical , distinguishing them from premeditated events. The study defines two postulates, related to premeditated threats: Correspondence – We should adopt the classical methodology whenever possible. Regulation – The safety of an NPP from premeditated threats requires examination, approval and inspection by a regulator. Key issues of the methodology with substantial differences from the

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.

Tracer methodology: an appropriate tool for assessing compliance with accreditation standards?

Science.gov (United States)

Bouchard, Chantal; Jean, Olivier

2017-10-01

Tracer methodology has been used by Accreditation Canada since 2008 to collect evidence on the quality and safety of care and services, and to assess compliance with accreditation standards. Given the importance of this methodology in the accreditation program, the objective of this study is to assess the quality of the methodology and identify its strengths and weaknesses. A mixed quantitative and qualitative approach was adopted to evaluate consistency, appropriateness, effectiveness and stakeholder synergy in applying the methodology. An online questionnaire was sent to 468 Accreditation Canada surveyors. According to surveyors' perceptions, tracer methodology is an effective tool for collecting useful, credible and reliable information to assess compliance with Qmentum program standards and priority processes. The results show good coherence between methodology components (appropriateness of the priority processes evaluated, activities to evaluate a tracer, etc.). The main weaknesses are the time constraints faced by surveyors and management's lack of cooperation during the evaluation of tracers. The inadequate amount of time allowed for the methodology to be applied properly raises questions about the quality of the information obtained. This study paves the way for a future, more in-depth exploration of the identified weaknesses to help the accreditation organization make more targeted improvements to the methodology. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-01

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

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

International Nuclear Information System (INIS)

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

2001-03-01

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

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

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

Model summary report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Vahlund, Fredrik

2006-10-15

This document is the model summary report for the safety assessment SR-Can. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SR-Can, a number of different computer codes are used. In order to better understand how these codes are related Assessment Model Flowcharts, AMFs, have been produced within the project. From these, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A large number of different computer codes are used in the assessment of which some are commercial while others are developed especially for the current assessment project. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: It must be demonstrated that the code is suitable for its purpose; It must be demonstrated that the code has been properly used; and, It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met.

Model summary report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Vahlund, Fredrik

2006-10-01

This document is the model summary report for the safety assessment SR-Can. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SR-Can, a number of different computer codes are used. In order to better understand how these codes are related Assessment Model Flowcharts, AMFs, have been produced within the project. From these, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A large number of different computer codes are used in the assessment of which some are commercial while others are developed especially for the current assessment project. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: It must be demonstrated that the code is suitable for its purpose; It must be demonstrated that the code has been properly used; and, It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

An approach to handling timescales in post-closure safety assessment

International Nuclear Information System (INIS)

Bailey, L.; Littleboy, A.

2002-01-01

Previous Nirex post-closure assessments of deep geological disposal have been based on the use of probabilistic safety analysis covering many millions of years. However, Nirex has also published an assessment methodology in which the assessment timescale is divided into a number of discrete periods of time (time frames). Nirex is currently at the stage of planning the next update to its generic post-closure performance assessment and is considering the merits of using an assessment methodology based on time frames, in order to improve links with operational assessments and the provision of advice on the packaging of wastes, and to encourage stakeholder dialogue. This paper has been prepared as part of Nirex's aim, wherever possible, to 'preview', or seek input from others on, its ideas for new work to generate discussion and feedback. It describes an evolution of Nirex's published assessment methodology and outlines how it could be applied in an updated post-closure performance assessment of the Nirex generic phased disposal concept. (authors)

SKI's and SSI's joint review of SKB's safety assessment report, SR 97. Summary

International Nuclear Information System (INIS)

2001-01-01

The Swedish Nuclear Fuel and Waste Management Co (SKB) has a programme for the siting of a repository for spent nuclear fuel in Swedish bedrock. In 1996, the Swedish Government decided that SKB must perform an assessment of the repository's long-term safety before undertaking the next step of the programme which entails drilling in a minimum of two municipalities (site investigations). SKB has presented such a safety assessment in SR 97 Post-closure Safety (henceforth referred to as SR 97). SR 97 is one of the documents in the comprehensive reporting that SKB must provide when it proposes sites for investigation. The Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Institute (SSI) have evaluated SR 97 in terms of its purposes which are to demonstrate a methodology for safety assessment, to show that Swedish bedrock can provide a safe repository using SKB's method, to provide a basis for specifying the factors that are important for site selection and to derive preliminary requirements on the function of the engineered barriers. The authorities have reached the following conclusions: SR 97 does not indicate any conditions that would mean that geological final disposal in accordance with SKB's method would have significant deficiencies in relation to the safety and radiation protection requirements of the authorities. SR 97 contains the elements required for a comprehensive assessment of safety and radiation protection. SKB's safety assessment methodology has improved within several important areas, such as the documentation of processes and properties that can affect repository performance and the development of models for safety assessment calculations. The methodology used in SR 97 has some deficiencies, for example, the specification of future events to be described in the safety assessment. SR 97 has not, to an adequate extent, dealt with unfavourable conditions that can affect the future safety of a repository. SKB states that the

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

Characterization and improvement of the nuclear safety culture through self-assessment

International Nuclear Information System (INIS)

Levin, H.A.; McGehee, R.B.; Cottle, W.T.

1996-01-01

Organizational culture has a powerful influence on overall corporate performance. The ability to sustain superior results in ensuring the public's health and safety is predicated on an organization's deeply embedded values and behavioral norms and how these affect the ability to change and seek continuous improvement. The nuclear industry is developing increased recognition of the relationship of culture to nuclear safety performance as a critical element of corporate strategy. This paper describes a self-assessment methodology designed to characterize and improve the nuclear safety culture, including processes for addressing employee concerns. This methodology has been successfully applied on more than 30 occasions in the last several years, resulting in measurable improvements in safety performance and quality and employee motivation, productivity, and morale. Benefits and lessons learned are also presented

Safety Assessment for Inertial Fusion Energy Power Plants: Methodology and Application to the Analysis of the HYLIFE-II and SOMBRERO Conceptual Designs

Science.gov (United States)

Reyes, S.; Latkowski, J. F.; Sanz, J.; Gomez del Rio, J.

2001-06-01

Although the safety and environmental (S & E) characteristics of fusion energy have long been emphasized, these benefits are not automatically achieved. To maximize the potential S & E attractiveness of the inertial fusion energy (IFE), analyses must be performed early in the designs so that lessons can be learned and intelligent decisions made. In this work we have introduced for the first time heat transfer and thermal-hydraulics calculations as part of a state-of-the-art set of codes and libraries in order to establish an updated methodology for IFE safety analysis. We have focused our efforts primarily on two IFE power plant conceptual designs: HYLIFE-II and SOMBRERO. To some degree, these designs represent the extremes in IFE power plant designs. Also, a preliminary safety assessment has been performed for a generic target fabrication facility producing various types of targets and using various production techniques. Although this study cannot address all issues and hazards posed by an IFE power plant, it advances our understanding of radiological safety of such facilities. This will enable better comparisons between IFE designs and competing technologies from the safety point of view.

Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

International Nuclear Information System (INIS)

Pritrsky, J.

2008-01-01

Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

International Nuclear Information System (INIS)

Pritrsky, J.

2008-02-01

Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

Soft systems methodology as a systemic approach to nuclear safety management

International Nuclear Information System (INIS)

Vieira Neto, Antonio S.; Guilhen, Sabine N.; Rubin, Gerson A.; Caldeira Filho, Jose S.; Camargo, Iara M.C.

2017-01-01

Safety approach currently adopted by nuclear installations is built almost exclusively upon analytical methodologies based, mainly, on the belief that the properties of a system, such as its safety, are given by its constituent parts. This approach, however, does not properly address the complex dynamic interactions between technical, human and organizational factors occurring within and outside the organization. After the accident at Fukushima Daiichi nuclear power plant in March 2011, experts of the International Atomic Energy Agency (IAEA) recommended a systemic approach as a complementary perspective to nuclear safety. The aim of this paper is to present an overview of the systems thinking approach and its potential use for structuring socio technical problems involved in the safety of nuclear installations, highlighting the methodologies related to the soft systems thinking, in particular the Soft Systems Methodology (SSM). The implementation of a systemic approach may thus result in a more holistic picture of the system by the complex dynamic interactions between technical, human and organizational factors. (author)

Soft systems methodology as a systemic approach to nuclear safety management

Energy Technology Data Exchange (ETDEWEB)

Vieira Neto, Antonio S.; Guilhen, Sabine N.; Rubin, Gerson A.; Caldeira Filho, Jose S.; Camargo, Iara M.C., E-mail: asvneto@ipen.br, E-mail: snguilhen@ipen.br, E-mail: garubin@ipen.br, E-mail: jscaldeira@ipen.br, E-mail: icamargo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

Safety approach currently adopted by nuclear installations is built almost exclusively upon analytical methodologies based, mainly, on the belief that the properties of a system, such as its safety, are given by its constituent parts. This approach, however, does not properly address the complex dynamic interactions between technical, human and organizational factors occurring within and outside the organization. After the accident at Fukushima Daiichi nuclear power plant in March 2011, experts of the International Atomic Energy Agency (IAEA) recommended a systemic approach as a complementary perspective to nuclear safety. The aim of this paper is to present an overview of the systems thinking approach and its potential use for structuring socio technical problems involved in the safety of nuclear installations, highlighting the methodologies related to the soft systems thinking, in particular the Soft Systems Methodology (SSM). The implementation of a systemic approach may thus result in a more holistic picture of the system by the complex dynamic interactions between technical, human and organizational factors. (author)

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

Performance assessment methodology (PAM) for low level radioactive waste (LLRW) disposal facilities

International Nuclear Information System (INIS)

Selander, W.N.

1992-01-01

An overview is given for Performance Assessment Methodology (PAM) for Low Level Radioactive Waste (LLRW) disposal technologies, as required for licensing and safety studies. This is a multi-disciplinary activity, emphasizing applied mathematics, mass transfer, geohydrology and radiotoxicity effects on humans. (author). 2 refs

Review of Software Reliability Assessment Methodologies for Digital I and C Software of Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Cho, Jae Hyun; Lee, Seung Jun; Jung, Won Dea [KAERI, Daejeon (Korea, Republic of)

2014-08-15

Digital instrumentation and control (I and C) systems are increasingly being applied to current nuclear power plants (NPPs) due to its advantages; zero drift, advanced data calculation capacity, and design flexibility. Accordingly, safety issues of software that is main part of the digital I and C system have been raised. As with hardware components, the software failure in NPPs could lead to a large disaster, therefore failure rate test and reliability assessment of software should be properly performed, and after that adopted in NPPs. However, the reliability assessment of the software is quite different with that of hardware, owing to the nature difference between software and hardware. The one of the most different thing is that the software failures arising from design faults as 'error crystal', whereas the hardware failures are caused by deficiencies in design, production, and maintenance. For this reason, software reliability assessment has been focused on the optimal release time considering the economy. However, the safety goal and public acceptance of the NPPs is so distinctive with other industries that the software in NPPs is dependent on reliability quantitative value rather than economy. The safety goal of NPPs compared to other industries is exceptionally high, so conventional methodologies on software reliability assessment already used in other industries could not adjust to safety goal of NPPs. Thus, the new reliability assessment methodology of the software of digital I and C on NPPs need to be developed. In this paper, existing software reliability assessment methodologies are reviewed to obtain the pros and cons of them, and then to assess the usefulness of each method to software of NPPs.

Review of Software Reliability Assessment Methodologies for Digital I and C Software of Nuclear Power Plants

International Nuclear Information System (INIS)

Cho, Jae Hyun; Lee, Seung Jun; Jung, Won Dea

2014-01-01

Digital instrumentation and control (I and C) systems are increasingly being applied to current nuclear power plants (NPPs) due to its advantages; zero drift, advanced data calculation capacity, and design flexibility. Accordingly, safety issues of software that is main part of the digital I and C system have been raised. As with hardware components, the software failure in NPPs could lead to a large disaster, therefore failure rate test and reliability assessment of software should be properly performed, and after that adopted in NPPs. However, the reliability assessment of the software is quite different with that of hardware, owing to the nature difference between software and hardware. The one of the most different thing is that the software failures arising from design faults as 'error crystal', whereas the hardware failures are caused by deficiencies in design, production, and maintenance. For this reason, software reliability assessment has been focused on the optimal release time considering the economy. However, the safety goal and public acceptance of the NPPs is so distinctive with other industries that the software in NPPs is dependent on reliability quantitative value rather than economy. The safety goal of NPPs compared to other industries is exceptionally high, so conventional methodologies on software reliability assessment already used in other industries could not adjust to safety goal of NPPs. Thus, the new reliability assessment methodology of the software of digital I and C on NPPs need to be developed. In this paper, existing software reliability assessment methodologies are reviewed to obtain the pros and cons of them, and then to assess the usefulness of each method to software of NPPs

Performance Testing Methodology for Safety-Critical Programmable Logic Controller

International Nuclear Information System (INIS)

Kim, Chang Ho; Oh, Do Young; Kim, Ji Hyeon; Kim, Sung Ho; Sohn, Se Do

2009-01-01

The Programmable Logic Controller (PLC) for use in Nuclear Power Plant safety-related applications is being developed and tested first time in Korea. This safety-related PLC is being developed with requirements of regulatory guideline and industry standards for safety system. To test that the quality of the developed PLC is sufficient to be used in safety critical system, document review and various product testings were performed over the development documents for S/W, H/W, and V/V. This paper provides the performance testing methodology and its effectiveness for PLC platform conducted by KOPEC

Risk assessment methodology for Hanford high-level waste tanks

International Nuclear Information System (INIS)

Bott, T.F.; Mac Farlane, D.R.; Stack, D.W.; Kindinger, J.

1992-01-01

A methodology is presented for applying Probabilistic Safety Assessment techniques to quantification of the health risks posed by the high-level waste (HLW) underground tanks at the Department of Energy's Hanford reservation. This methodology includes hazard screening development of a list of potential accident initiators, systems fault trees development and quantification, definition of source terms for various release categories, and estimation of health consequences from the releases. Both airborne and liquid pathway releases to the environment, arising from aerosol and spill/leak releases from the tanks, are included in the release categories. The proposed methodology is intended to be applied to a representative subset of the total of 177 tanks, thereby providing a baseline risk profile for the HLW tank farm that can be used for setting clean-up/remediation priorities. Some preliminary results are presented for Tank 101-SY

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

International Nuclear Information System (INIS)

Metcalf, P.

2003-01-01

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

A global view on ARAMIS, a risk assessment methodology for industries in the framework of the SEVESO II directive

International Nuclear Information System (INIS)

Salvi, Olivier; Debray, Bruno

2006-01-01

The ARAMIS methodology was developed in an European project co-funded in the fifth Framework Programme of the European Commission with the objective to answer the specific requirements of the SEVESO II directive. It offers an alternative to purely deterministic and probabilistic approaches to risk assessment of process plants. It also answers the needs of the various stakeholders interested by the results of the risk assessment for land use or emergency planning, enforcement or, more generally, public decision-making. The methodology is divided into the following major steps: identification of major accident hazards (MIMAH), identification of the safety barriers and assessment of their performances, evaluation of safety management efficiency to barrier reliability, identification of reference accident scenarios (MIRAS), assessment and mapping of the risk severity of reference scenarios and of the vulnerability of the plant surroundings. The methodology was tested during five case studies, which provided useful information about the applicability of the method and, by identifying the most sensitive parts of it opened way to new research activity for an improved industrial safety

A cost-effective methodology to internalize nuclear safety in nuclear reactor conceptual design

International Nuclear Information System (INIS)

Gimenez, M.; Grinblat, P.; Schlamp, M.

2003-01-01

A new methodology to perform nuclear reactor design, balancing safety and economics at the conceptual engineering stage, is presented in this work. The goal of this integral methodology is to take into account safety aspects in an optimization design process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behavior during accidents (safety performance indicators), are synthesized on Design Maps. These maps allow one to compare the safety indicator with limits, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimization process, by means of additional rules to the neutronic, thermal-hydraulic, and mechanical calculations. An application of the methodology, implemented in Integrated Reactor Evaluation Program 3 (IREP3) code, to optimize safety systems of CAREM prototype is presented. It consists in balancing the designs of the Emergency Injection System (EIS), the Residual Heat Removal System (RHRS), the primary circuit water inventory and the containment height, to cope with loss of coolant and loss of heat sink (LOHS) accidental sequences, taking into account cost and reactor performance. This methodology turns out to be promising to internalize cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels

An Evaluation Methodology Development and Application Process for Severe Accident Safety Issue Resolution

Directory of Open Access Journals (Sweden)

Robert P. Martin

2012-01-01

Full Text Available A general evaluation methodology development and application process (EMDAP paradigm is described for the resolution of severe accident safety issues. For the broader objective of complete and comprehensive design validation, severe accident safety issues are resolved by demonstrating comprehensive severe-accident-related engineering through applicable testing programs, process studies demonstrating certain deterministic elements, probabilistic risk assessment, and severe accident management guidelines. The basic framework described in this paper extends the top-down, bottom-up strategy described in the U.S Nuclear Regulatory Commission Regulatory Guide 1.203 to severe accident evaluations addressing U.S. NRC expectation for plant design certification applications.

A comparison of integrated safety analysis and probabilistic risk assessment

International Nuclear Information System (INIS)

Damon, Dennis R.; Mattern, Kevin S.

2013-01-01

The U.S. Nuclear Regulatory Commission conducted a comparison of two standard tools for risk informing the regulatory process, namely, the Probabilistic Risk Assessment (PRA) and the Integrated Safety Analysis (ISA). PRA is a calculation of risk metrics, such as Large Early Release Frequency (LERF), and has been used to assess the safety of all commercial power reactors. ISA is an analysis required for fuel cycle facilities (FCFs) licensed to possess potentially critical quantities of special nuclear material. A PRA is usually more detailed and uses more refined models and data than an ISA, in order to obtain reasonable quantitative estimates of risk. PRA is considered fully quantitative, while most ISAs are typically only partially quantitative. The extension of PRA methodology to augment or supplant ISAs in FCFs has long been considered. However, fuel cycle facilities have a wide variety of possible accident consequences, rather than a few surrogates like LERF or core damage as used for reactors. It has been noted that a fuel cycle PRA could be used to better focus attention on the most risk-significant structures, systems, components, and operator actions. ISA and PRA both identify accident sequences; however, their treatment is quite different. ISA's identify accidents that lead to high or intermediate consequences, as defined in 10 Code of Federal Regulations (CFR) 70, and develop a set of Items Relied on For Safety (IROFS) to assure adherence to performance criteria. PRAs identify potential accident scenarios and estimate their frequency and consequences to obtain risk metrics. It is acceptable for ISAs to provide bounding evaluations of accident consequences and likelihoods in order to establish acceptable safety; but PRA applications usually require a reasonable quantitative estimate, and often obtain metrics of uncertainty. This paper provides the background, features, and methodology associated with the PRA and ISA. The differences between the

The awareness of employees in safety culture through the improved nuclear safety culture evaluation method

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Ga; Sung, Chan Ho; Jung, Yeon Sub [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2012-10-15

After the Chernobyl nuclear accident in 1986, nuclear safety culture terminology was at first introduced emphasizing the importance of employees' attitude and organizational safety. The concept of safety culture was spread by INSAG 4 published in 1991. From that time, IAEA had provided the service of ASCOT for the safety culture assessment. However, many people still are thinking that safety culture is abstract and is not clear. It is why the systematic and reliable assessment methodology was not developed. Assessing safety culture is to identify what is the basic assumption for any organization to accept unconsciously. Therefore, it is very difficult to reach a meaningful conclusion by a superficial investigation alone. KHNP had been doing the safety culture assessment which was based on ASCOT methodology every 2 years. And this result had contributed to improving safety culture. But this result could not represent the level of organization's safety culture due to the limitation of method. So, KHNP has improved the safety culture method by benchmarking the over sea assessment techniques in 2011. The effectiveness of this improved methodology was validated through a pilot assessment. In this paper, the level of employees' safety culture awareness was analyzed by the improved method and reviewed what is necessary for the completeness and objectivity of the nuclear safety culture assessment methodology.

The awareness of employees in safety culture through the improved nuclear safety culture evaluation method

International Nuclear Information System (INIS)

Kim, Young Ga; Sung, Chan Ho; Jung, Yeon Sub

2012-01-01

After the Chernobyl nuclear accident in 1986, nuclear safety culture terminology was at first introduced emphasizing the importance of employees' attitude and organizational safety. The concept of safety culture was spread by INSAG 4 published in 1991. From that time, IAEA had provided the service of ASCOT for the safety culture assessment. However, many people still are thinking that safety culture is abstract and is not clear. It is why the systematic and reliable assessment methodology was not developed. Assessing safety culture is to identify what is the basic assumption for any organization to accept unconsciously. Therefore, it is very difficult to reach a meaningful conclusion by a superficial investigation alone. KHNP had been doing the safety culture assessment which was based on ASCOT methodology every 2 years. And this result had contributed to improving safety culture. But this result could not represent the level of organization's safety culture due to the limitation of method. So, KHNP has improved the safety culture method by benchmarking the over sea assessment techniques in 2011. The effectiveness of this improved methodology was validated through a pilot assessment. In this paper, the level of employees' safety culture awareness was analyzed by the improved method and reviewed what is necessary for the completeness and objectivity of the nuclear safety culture assessment methodology

Model summary report for the safety assessment SR-Site

International Nuclear Information System (INIS)

Vahlund, Fredrik; Zetterstroem Evins, Lena; Lindgren, Maria

2010-12-01

This document is the model summary report for the safety assessment SR-Site. In the report, the quality assurance (QA) measures conducted for assessment codes are presented together with the chosen QA methodology. In the safety assessment project SR-Site, a large number of numerical models are used to analyse the system and to show compliance. In order to better understand how the different models interact and how information are transferred between the different models Assessment Model Flowcharts, AMFs, are used. From these, different modelling tasks can be identify and the computer codes used. As a large number of computer codes are used in the assessment the complexity of these differs to a large extent, some of the codes are commercial while others are developed especially for the assessment at hand. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined for all codes: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. - It must be described how data are transferred between the different computational tasks. Although the requirements are identical for all codes in the assessment, the measures used to show that the requirements are fulfilled will be different for different types of codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented together with a discussion on how the requirements are met

Model summary report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

Vahlund, Fredrik; Zetterstroem Evins, Lena (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)); Lindgren, Maria (Kemakta Konsult AB, Stockholm (Sweden))

2010-12-15

This document is the model summary report for the safety assessment SR-Site. In the report, the quality assurance (QA) measures conducted for assessment codes are presented together with the chosen QA methodology. In the safety assessment project SR-Site, a large number of numerical models are used to analyse the system and to show compliance. In order to better understand how the different models interact and how information are transferred between the different models Assessment Model Flowcharts, AMFs, are used. From these, different modelling tasks can be identify and the computer codes used. As a large number of computer codes are used in the assessment the complexity of these differs to a large extent, some of the codes are commercial while others are developed especially for the assessment at hand. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined for all codes: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. - It must be described how data are transferred between the different computational tasks. Although the requirements are identical for all codes in the assessment, the measures used to show that the requirements are fulfilled will be different for different types of codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented together with a discussion on how the requirements are met

Improvement of operational safety: The self-assessment at the Russian NPPs

International Nuclear Information System (INIS)

Kolotov, Aleksander

2002-01-01

The operating organization has scheduled for the forthcoming year to elaborate the NPP safety self-assessment standards and to settle precise criterion for its performance. Toward this end it was decided to form a Working Team including of NPP and VNIIAES representatives to elaborate major tasks on the self-assessment improvement, development and implementation of new documentation and training the personnel in new methodology. Actions developed by R osenergoatom , one of its items is the participation of 'Rosenergoatom' and VNIIAES representatives in the workshop, contains the sequence of Russian NPP safety self-assessment improvement at the first stage

Comparison of plant-specific probabilistic safety assessments and lessons learned

International Nuclear Information System (INIS)

Balfanz, H.P.; Berg, H.P.; Steininger, U.

2001-01-01

Probabilistic safety assessments (PSA) have been performed for all German nuclear power plants in operation. These assessments are mainly based on the recent German PSA guide and an earlier draft, respectively. However, comparison of these PSA show differences in the results which are discussed in this paper. Lessons learned from this comparison and further development of the PSA methodology are described. (orig.) [de

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

Knowledge Management Methodologies for Improving Safety Culture

International Nuclear Information System (INIS)

Rusconi, C.

2016-01-01

Epistemic uncertainties could affect operator’s capability to prevent rare but potentially catastrophic accident sequences. Safety analysis methodologies are powerful but fragile tools if basic assumptions are not sound and exhaustive. In particular, expert judgments and technical data could be invalidated by organizational context change (e.g., maintenance planning, supply systems etc.) or by unexpected events. In 1986 accidents like Chernobyl, the explosion of Shuttle Challenger and, two years before, the toxic release at Bhopal chemical plant represented the point of no return with respect to the previous vision of safety and highlighted the undelayable need to change paradigm and face safety issues in complex systems not only from a technical point of view but to adopt a systemic vision able to include and integrate human and organizational aspects.

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

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

Using Risk Assessment Methodologies to Meet Management Objectives

Science.gov (United States)

DeMott, D. L.

2015-01-01

Corporate and program objectives focus on desired performance and results. ?Management decisions that affect how to meet these objectives now involve a complex mix of: technology, safety issues, operations, process considerations, employee considerations, regulatory requirements, financial concerns and legal issues. ?Risk Assessments are a tool for decision makers to understand potential consequences and be in a position to reduce, mitigate or eliminate costly mistakes or catastrophic failures. Using a risk assessment methodology is only a starting point. ?A risk assessment program provides management with important input in the decision making process. ?A pro-active organization looks to the future to avoid problems, a reactive organization can be blindsided by risks that could have been avoided. ?You get out what you put in, how useful your program is will be up to the individual organization.

Methodology of environmental risk assessment management

Directory of Open Access Journals (Sweden)

Saša T. Bakrač

2012-04-01

Full Text Available Successful protection of environment is mostly based on high-quality assessment of potential and present risks. Environmental risk management is a complex process which includes: identification, assessment and control of risk, namely taking measures in order to minimize the risk to an acceptable level. Environmental risk management methodology: In addition to these phases in the management of environmental risk, appropriate measures that affect the reduction of risk occurrence should be implemented: - normative and legal regulations (laws and regulations, - appropriate organizational structures in society, and - establishing quality monitoring of environment. The emphasis is placed on the application of assessment methodologies (three-model concept, as the most important aspect of successful management of environmental risk. Risk assessment methodology - European concept: The first concept of ecological risk assessment methodology is based on the so-called European model-concept. In order to better understand this ecological risk assessment methodology, two concepts - hazard and risk - are introduced. The European concept of environmental risk assessment has the following phases in its implementation: identification of hazard (danger, identification of consequences (if there is hazard, estimate of the scale of consequences, estimate of consequence probability and risk assessment (also called risk characterization. The European concept is often used to assess risk in the environment as a model for addressing the distribution of stressors along the source - path - receptor line. Risk assessment methodology - Canadian concept: The second concept of the methodology of environmental risk assessment is based on the so-called Canadian model-concept. The assessment of ecological risk includes risk arising from natural events (floods, extreme weather conditions, etc., technological processes and products, agents (chemical, biological, radiological, etc

Handling of future human actions in the safety assessment SR-Can

International Nuclear Information System (INIS)

Moren, Lena

2006-10-01

This report documents the future human actions (FHA) considered in the long-term safety analysis of a KBS-3 repository. The report is one of the supporting documents to the safety assessment SR-Can. The purpose of this report is to provide an account of: General considerations concerning FHA; The methodology applied in SR-Can to assess FHA; The aspects of FHA that need to be considered in the evaluation of their impact on a deep geological repository; and The selection of representative scenarios for illustrative consequence analysis

Handling of future human actions in the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Moren, Lena

2006-10-15

This report documents the future human actions (FHA) considered in the long-term safety analysis of a KBS-3 repository. The report is one of the supporting documents to the safety assessment SR-Can. The purpose of this report is to provide an account of: General considerations concerning FHA; The methodology applied in SR-Can to assess FHA; The aspects of FHA that need to be considered in the evaluation of their impact on a deep geological repository; and The selection of representative scenarios for illustrative consequence analysis.

AEGIS methodology and a perspective from AEGIS methodology demonstrations

International Nuclear Information System (INIS)

Dove, F.H.

1981-03-01

Objectives of AEGIS (Assessment of Effectiveness of Geologic Isolation Systems) are to develop the capabilities needed to assess the post-closure safety of waste isolation in geologic formation; demonstrate these capabilities on reference sites; apply the assessment methodology to assist the NWTS program in site selection, waste package and repository design; and perform repository site analyses for the licensing needs of NWTS. This paper summarizes the AEGIS methodology, the experience gained from methodology demonstrations, and provides an overview in the following areas: estimation of the response of a repository to perturbing geologic and hydrologic events; estimation of the transport of radionuclides from a repository to man; and assessment of uncertainties

A study on the methodology of integrated safety assessment on low and intermediate level waste (LILW) managed in temporary storage facility at NPP

International Nuclear Information System (INIS)

Ahn, Min Ho

2010-02-01

Since 1978, the KHNP has been operating 20 NPPs (16 PWRs and 4 CANDUs) and generating about 67,000 drums (200 L) of LILW (as of December 31, 2005), which have been stored in the temporary storage facility (TSF) at each NPP due to the absence of a repository for the disposal of LILW. Therefore, the period of temporary storage of LILW is so long compared to other countries. Furthermore, the details with respect to the safety analyse on the TSF have not been considered in PSAR and FSAR. Especially, the risk assessment on the TSF has scarcely been conducted as opposed to many researches on the disposal of LILW. Since 2003, however, the IAEA has been recognized on the importance of predispoal management of LILW. And then, the regulatory frame of U.S. NRC was being shifted to risk-based regulation from the deterministic approach. Therefore, most of radioactive wastes including the LILW will be managed in terms of the risk-based graded approach to future regulation system called RIR (risk informed regulation). If the radioactive wastes do not quantitatively deal with the risk-based regulation, the radiological risk on some of radioactive wastes might be overestimated or underestimated regardless of the degree of the risk. According to a consequence of these situations, the numbers of the researches on the predisposal management of LILW have been required for the preparation on new regulatory frame. In this study, the main objective of this study is to establish the methodology of integrated safety assessment on LILW managed in the TSF at NPP, and to develop the integrated safety assessment code for routine operating condition and for for accident analysis on LILW managed in the TSF. In order to establish the methodology of integrated safety assessment on LILW managed in the TSF at NPP, three main parameters were considered: risk-based accident scenarios, radionuclide inventory, and atmospheric dispersion factor (χ/Q). Arbitrary accidents related to LILW management in the

Taipower's reload safety evaluation methodology for pressurized water reactors

International Nuclear Information System (INIS)

Huang, Ping-Hue; Yang, Y.S.

1996-01-01

For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

INPRO Methodology for Sustainability Assessment of Nuclear Energy Systems: Environmental Impact of Stressors. INPRO Manual

International Nuclear Information System (INIS)

2016-01-01

This publication provides guidance on assessing of sustainability of a nuclear energy system (NES) in the area of environmental impact of stressors. The INPRO methodology is a comprehensive tool for the assessment of sustainability of an NES. Basic principles, user requirements and criteria have been defined in different areas of INPRO methodology. These include economics, infrastructure, waste management, proliferation resistance, environmental impact of stressors, environmental impact from depletion of resources, and safety of nuclear reactors and fuel cycle facilities. The ultimate goal of the application of the INPRO methodology is to check whether the assessed NES fulfils all the criteria, and hence the user requirements and basic principles, and therefore presents a system for a Member State that is sustainable in the long term

A Methodological Framework for Software Safety in Safety Critical Computer Systems

OpenAIRE

P. V. Srinivas Acharyulu; P. Seetharamaiah

2012-01-01

Software safety must deal with the principles of safety management, safety engineering and software engineering for developing safety-critical computer systems, with the target of making the system safe, risk-free and fail-safe in addition to provide a clarified differentaition for assessing and evaluating the risk, with the principles of software risk management. Problem statement: Prevailing software quality models, standards were not subsisting in adequately addressing the software safety ...

Review of probabilistic safety assessments: insights and recommendations regarding further developments

International Nuclear Information System (INIS)

Spitzer, C.

1996-01-01

Probabilistic Safety Assessments (PSAs) performed by utilities in the framework of Periodic Safety Reviews for German nuclear power plants are reviewed by TUeV Suedwest. Insights gained and recommendations concerning the necessity and focus of further developments and applications according to practical requests for the performance and assessment of PSAs within regulatory procedures are presented in this paper. Further on, recommendations are made in order to ensure the validity of the results of PSAs necessary in order to achieve the goals thereof. Beside some general points of view the emphasis of the paper is on methodological aspects with respect to evaluation methods and assessment of common cause failures as well as human reliability assessment

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

Safety methodology and risk targets

International Nuclear Information System (INIS)

Kazimi, M.S.

1983-01-01

In assessing the potential safety concerns of fusion, the experience from other energy sources lead to a variety of safety assessment approaches. The available approaches are: (1) The maximum possible accident approach; (2) The maximum credible accident approach; (3) The probabilistic total risk assessment. In the first approach, the mechanistic development of the events leading to the safety concern is ignored. Instead, the total radioactivity of the plant is assumed accessible to the public. Such an approach is obviously conservative and unrealistic. In the second approach a selection is made among the most severe of the possible accidents, and the progression of the accident is modeled as mechanistically as possible. In this case, the passive and active accident mitigation capabilities of the plant are taken into consideration. The result is expected to be that none or only a fraction of the total radioactivity can be released to the public. The adverse effect of this approach is to concentrate attention on a particular accident class, and perhaps not allow for other classes, a judgement that may later become undesirable. The probabilistic risk assessment requires the safety analysts to consider all classes of accidents and estimate both the probabilities of their occurrences and their consequences. Thus, the plant design in fact is subjected to a thorough investigation and the impact of alterations in design can be reflected in the total risk estimate. The disadvantage of this approach lies in the absence of well defined acceptable risk criteria as well as the large effect of public perception factors on the accepted risk. This paper will review the impact of application of these approaches in determination of the level of protection needed against activation product release to the atmosphere. (author)

Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

2010-03-01

The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the probabilistic safety assessment (PSA) domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

The development of a methodology to assess population doses from multiple sources and exposure pathways of radioactivity

International Nuclear Information System (INIS)

Hancox, J.; Stansby, S.; Thorne, M.

2002-01-01

The Environment Agency (EA) has new duties in accordance with the Basic Safety Standards Directive under which it is required to ensure that doses to individuals received from exposure to anthropogenic sources of radioactivity are within defined limits. In order to assess compliance with these requirements, the EA needs to assess the doses to members of the most highly exposed population groups ('critical' groups) from all relevant potential sources of anthropogenic radioactivity and all relevant potential exposure pathways to such radioactivity. The EA has identified a need to develop a methodology for the retrospective assessment of effective doses from multiple sources of radioactive materials and exposure pathways associated with those sources. Under contract to the EA, AEA Technology has undertaken the development of a suitable methodology as part of EA R and D Project P3-070. The methodology developed under this research project has been designed to support the EA in meeting its obligations under the Euratom Basic Safety Standards Directive and is consistent with UK and international approaches to radiation dosimetry and radiological protection. The development and trial application of the methodology is described in this report

Work-related stress risk assessment in Italy: a methodological proposal adapted to regulatory guidelines.

Science.gov (United States)

Persechino, Benedetta; Valenti, Antonio; Ronchetti, Matteo; Rondinone, Bruna Maria; Di Tecco, Cristina; Vitali, Sara; Iavicoli, Sergio

2013-06-01

Work-related stress is one of the major causes of occupational ill health. In line with the regulatory framework on occupational health and safety (OSH), adequate models for assessing and managing risk need to be identified so as to minimize the impact of this stress not only on workers' health, but also on productivity. After close analysis of the Italian and European reference regulatory framework and work-related stress assessment and management models used in some European countries, we adopted the UK Health and Safety Executive's (HSE) Management Standards (MS) approach, adapting it to the Italian context in order to provide a suitable methodological proposal for Italy. We have developed a work-related stress risk assessment strategy, meeting regulatory requirements, now available on a specific web platform that includes software, tutorials, and other tools to assist companies in their assessments. This methodological proposal is new on the Italian work-related stress risk assessment scene. Besides providing an evaluation approach using scientifically validated instruments, it ensures the active participation of occupational health professionals in each company. The assessment tools provided enable companies not only to comply with the law, but also to contribute to a database for monitoring and assessment and give access to a reserved area for data analysis and comparisons.

Model summary report for the safety assessment SFR 1 SAR-08

Energy Technology Data Exchange (ETDEWEB)

2008-03-15

This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

Model summary report for the safety assessment SFR 1 SAR-08

International Nuclear Information System (INIS)

2008-03-01

This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

Methodology briefing students in the safety on physical education in the gym

Directory of Open Access Journals (Sweden)

N. N. Mukhamediarov

2014-01-01

Full Text Available Purpose : to determine the effective components of the methodology of coaching students on safety on physical education in the gym. Material : in the experiment involved 90 students aged 17-18 years. Results : the method of forming responsible attitude to the basics of safety during physical training in the gym. Developed special funds: lectures, seminars, analysis of articles, diagnostic interviews, questionnaires, analysis of log data of attendance, log injuries. The criteria of formation responsible attitude to physical training. The results of the implementation methodology. Conclusions : proposed method recommend to instruct students on safety. Use the means of forming a responsible attitude to safety during physical training in the gym that will help reduce injuries and improve quality of the physical training at the university.

Problems of probabilistic safety assessment after Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

Sugiyama, Naoki

2011-01-01

Probabilistic safety assessment (PSA) methodology to assure nuclear safety is had great expectations of lessons learned from Fukushima Daiichi nuclear power plant (NPP) accident and on the other hand this accident made actualized technical problems of PSA. Effectiveness of current PSA methodology for risk assessment was confirmed by comparing the accident development with accident scenario of PSA and equipment failure rate. From a viewpoint of nuclear safety objective and defense in depth approach of IAEA, technical problems of PSA were (1) extension of PSA for spent fuel pool and waste disposal system as well as level 3PSA for broader environmental contamination and (2) overlapping of accident scenario of plural unit site, balance of high quality plant management and preceding negation, treatment of uncertainty of external events, severe accident measure and human reliability analysis and reflection of disaster prevention capability to level 3PSA. In order to upgrade PSA technology, six proposals were described for nuclear safety and defense in depth, comprehensive evaluation scope and catch-up of latest technology, necessity of strategic preparation of PSA standard, human resources fostering and risk communication. (T. Tanaka)

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.

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

Methodology for reliability based condition assessment

International Nuclear Information System (INIS)

Mori, Y.; Ellingwood, B.

1993-08-01

Structures in nuclear power plants may be exposed to aggressive environmental effects that cause their strength to decrease over an extended period of service. A major concern in evaluating the continued service for such structures is to ensure that in their current condition they are able to withstand future extreme load events during the intended service life with a level of reliability sufficient for public safety. This report describes a methodology to facilitate quantitative assessments of current and future structural reliability and performance of structures in nuclear power plants. This methodology takes into account the nature of past and future loads, and randomness in strength and in degradation resulting from environmental factors. An adaptive Monte Carlo simulation procedure is used to evaluate time-dependent system reliability. The time-dependent reliability is sensitive to the time-varying load characteristics and to the choice of initial strength and strength degradation models but not to correlation in component strengths within a system. Inspection/maintenance strategies are identified that minimize the expected future costs of keeping the failure probability of a structure at or below an established target failure probability during its anticipated service period

Development of the methodology and approaches to validate safety and accident management

International Nuclear Information System (INIS)

Asmolov, V.G.

1997-01-01

The article compares the development of the methodology and approaches to validate the nuclear power plant safety and accident management in Russia and advanced industrial countries. It demonstrates that the development of methods of safety validation is dialectically related to the accumulation of the knowledge base on processes and events during NPP normal operation, transients and emergencies, including severe accidents. The article describes the Russian severe accident research program (1987-1996), the implementation of which allowed Russia to reach the world level of the safety validation efforts, presents future high-priority study areas. Problems related to possible approaches to the methodological accident management development are discussed. (orig.)

The development of a safety analysis methodology for the optimized power reactor 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun; Yo-Han, Kim

2005-01-01

Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

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)

Methodology for safety classification of PWR type nuclear power plants items

International Nuclear Information System (INIS)

Oliveira, Patricia Pagetti de

1995-01-01

This paper contains the criteria and methodology which define a classification system of structures, systems and components in safety classes according to their importance to nuclear safety. The use of this classification system will provide a set of basic safety requirements associated with each safety class specified. These requirements, when available and applicable, shall be utilized in the design, fabrication and installation of structures, systems and components of Pressurized Water Reactor Nuclear Power Plants. (author). 13 refs, 1 tab

Are automatic systems the future of motorcycle safety? A novel methodology to prioritize potential safety solutions based on their projected effectiveness.

Science.gov (United States)

Gil, Gustavo; Savino, Giovanni; Piantini, Simone; Baldanzini, Niccolò; Happee, Riender; Pierini, Marco

2017-11-17

Motorcycle riders are involved in significantly more crashes per kilometer driven than passenger car drivers. Nonetheless, the development and implementation of motorcycle safety systems lags far behind that of passenger cars. This research addresses the identification of the most effective motorcycle safety solutions in the context of different countries. A knowledge-based system of motorcycle safety (KBMS) was developed to assess the potential for various safety solutions to mitigate or avoid motorcycle crashes. First, a set of 26 common crash scenarios was identified from the analysis of multiple crash databases. Second, the relative effectiveness of 10 safety solutions was assessed for the 26 crash scenarios by a panel of experts. Third, relevant information about crashes was used to weigh the importance of each crash scenario in the region studied. The KBMS method was applied with an Italian database, with a total of more than 1 million motorcycle crashes in the period 2000-2012. When applied to the Italian context, the KBMS suggested that automatic systems designed to compensate for riders' or drivers' errors of commission or omission are the potentially most effective safety solution. The KBMS method showed an effective way to compare the potential of various safety solutions, through a scored list with the expected effectiveness of each safety solution for the region to which the crash data belong. A comparison of our results with a previous study that attempted a systematic prioritization of safety systems for motorcycles (PISa project) showed an encouraging agreement. Current results revealed that automatic systems have the greatest potential to improve motorcycle safety. Accumulating and encoding expertise in crash analysis from a range of disciplines into a scalable and reusable analytical tool, as proposed with the use of KBMS, has the potential to guide research and development of effective safety systems. As the expert assessment of the crash

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

International Nuclear Information System (INIS)

Mateeva, M.

2003-01-01

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

WE-B-BRC-01: Current Methodologies in Risk Assessment

Energy Technology Data Exchange (ETDEWEB)

Rath, F. [University of Wisconsin Madison (United States)

2016-06-15

Prospective quality management techniques, long used by engineering and industry, have become a growing aspect of efforts to improve quality management and safety in healthcare. These techniques are of particular interest to medical physics as scope and complexity of clinical practice continue to grow, thus making the prescriptive methods we have used harder to apply and potentially less effective for our interconnected and highly complex healthcare enterprise, especially in imaging and radiation oncology. An essential part of most prospective methods is the need to assess the various risks associated with problems, failures, errors, and design flaws in our systems. We therefore begin with an overview of risk assessment methodologies used in healthcare and industry and discuss their strengths and weaknesses. The rationale for use of process mapping, failure modes and effects analysis (FMEA) and fault tree analysis (FTA) by TG-100 will be described, as well as suggestions for the way forward. This is followed by discussion of radiation oncology specific risk assessment strategies and issues, including the TG-100 effort to evaluate IMRT and other ways to think about risk in the context of radiotherapy. Incident learning systems, local as well as the ASTRO/AAPM ROILS system, can also be useful in the risk assessment process. Finally, risk in the context of medical imaging will be discussed. Radiation (and other) safety considerations, as well as lack of quality and certainty all contribute to the potential risks associated with suboptimal imaging. The goal of this session is to summarize a wide variety of risk analysis methods and issues to give the medical physicist access to tools which can better define risks (and their importance) which we work to mitigate with both prescriptive and prospective risk-based quality management methods. Learning Objectives: Description of risk assessment methodologies used in healthcare and industry Discussion of radiation oncology

WE-B-BRC-01: Current Methodologies in Risk Assessment

International Nuclear Information System (INIS)

Rath, F.

2016-01-01

Prospective quality management techniques, long used by engineering and industry, have become a growing aspect of efforts to improve quality management and safety in healthcare. These techniques are of particular interest to medical physics as scope and complexity of clinical practice continue to grow, thus making the prescriptive methods we have used harder to apply and potentially less effective for our interconnected and highly complex healthcare enterprise, especially in imaging and radiation oncology. An essential part of most prospective methods is the need to assess the various risks associated with problems, failures, errors, and design flaws in our systems. We therefore begin with an overview of risk assessment methodologies used in healthcare and industry and discuss their strengths and weaknesses. The rationale for use of process mapping, failure modes and effects analysis (FMEA) and fault tree analysis (FTA) by TG-100 will be described, as well as suggestions for the way forward. This is followed by discussion of radiation oncology specific risk assessment strategies and issues, including the TG-100 effort to evaluate IMRT and other ways to think about risk in the context of radiotherapy. Incident learning systems, local as well as the ASTRO/AAPM ROILS system, can also be useful in the risk assessment process. Finally, risk in the context of medical imaging will be discussed. Radiation (and other) safety considerations, as well as lack of quality and certainty all contribute to the potential risks associated with suboptimal imaging. The goal of this session is to summarize a wide variety of risk analysis methods and issues to give the medical physicist access to tools which can better define risks (and their importance) which we work to mitigate with both prescriptive and prospective risk-based quality management methods. Learning Objectives: Description of risk assessment methodologies used in healthcare and industry Discussion of radiation oncology

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.

Knowledge representation in safety assessment: improving transparency and traceability

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Knowledge representation in safety assessment: improving transparency and traceability

International Nuclear Information System (INIS)

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

2011-01-01

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

Compliance strategy for statistically based neutron overpower protection safety analysis methodology

International Nuclear Information System (INIS)

Holliday, E.; Phan, B.; Nainer, O.

2009-01-01

The methodology employed in the safety analysis of the slow Loss of Regulation (LOR) event in the OPG and Bruce Power CANDU reactors, referred to as Neutron Overpower Protection (NOP) analysis, is a statistically based methodology. Further enhancement to this methodology includes the use of Extreme Value Statistics (EVS) for the explicit treatment of aleatory and epistemic uncertainties, and probabilistic weighting of the initial core states. A key aspect of this enhanced NOP methodology is to demonstrate adherence, or compliance, with the analysis basis. This paper outlines a compliance strategy capable of accounting for the statistical nature of the enhanced NOP methodology. (author)

Methodology assessment and recommendations for the Mars science laboratory launch safety analysis.

Energy Technology Data Exchange (ETDEWEB)

Sturgis, Beverly Rainwater; Metzinger, Kurt Evan; Powers, Dana Auburn; Atcitty, Christopher B.; Robinson, David B; Hewson, John C.; Bixler, Nathan E.; Dodson, Brian W.; Potter, Donald L.; Kelly, John E.; MacLean, Heather J.; Bergeron, Kenneth Donald (Sala & Associates); Bessette, Gregory Carl; Lipinski, Ronald J.

2006-09-01

thousands of possible event sequences and to build up a statistical representation of the releases for each accident case. A code to carry out this process will have to be developed or adapted from previous MMRTG missions. Finally, Level C translates the release (or ''source term'') information from Level B into public risk by applying models for atmospheric transport and the health consequences of exposure to the released plutonium dioxide. A number of candidate codes for this level of analysis are available. This report surveys the range of available codes and tools for each of these levels and makes recommendations for which choices are best for the MSL mission. It also identities areas where improvements to the codes are needed. In some cases a second tier of codes may be identified to provide supporting or clarifying insight about particular issues. The main focus of the methodology assessment is to identify a suite of computational tools that can produce a high quality SAR that can be successfully reviewed by external bodies (such as the Interagency Nuclear Safety Review Panel) on the schedule established by NASA and DOE.

Safety assessment of computerized instrumentation and control for nuclear power plants

International Nuclear Information System (INIS)

Fride, B.; Henry, J.Y.; Manners, S.

1996-01-01

France's latest 1400 MWe 'N4' generation of Pressurised Water Reactors (PWR) use distributed programmable control systems interconnected by data networks. The protection system is also software based. IPSN have the task of evaluating the safety demonstration before the government safety authority (DSIN) give the licensee (EDF) permission to fuel the reactor and to raise power. Some of the different aspects of the evaluation carried out and the methodologies used for assessing the C and I are presented. (author)

Assessment of two small-sized innovative nuclear reactors for electricity generation in Brazil using INPRO methodology

International Nuclear Information System (INIS)

Goncalves Filho, Orlando Joao Agostinho; Sefidvash, Farhang

2009-01-01

This paper presents the main results of the assessment study of two small-sized innovative reactors for electricity generation in Brazil using the methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO was initiated in 2001 and has the main objective of helping to ensure that nuclear energy is available to contribute in a sustainable manner to the energy needs of the 21st century. Brazil joined the INPRO project since its beginning and in 2005 submitted a proposal for the assessment using INPRO methodology of two small-sized reactors (IRIS - International Reactor Innovative and Secure, and FBNR - Fixed Bed Nuclear Reactor) as potential components of an innovative nuclear energy system (INS) completed by a conventional open nuclear fuel cycle based on enriched uranium. The scope of this assessment study was restricted to the reactor component of the INS and to the methodology areas of economics and safety for IRIS, and proliferation resistance and safety for FBNR. The results indicate that both IRIS and FBNR innovative designs comply mostly with the basic principles of the areas assessed and have potential to comply with the remaining ones. (author)

A methodology for analysing human errors of commission in accident scenarios for risk assessment

International Nuclear Information System (INIS)

Kim, J. H.; Jung, W. D.; Park, J. K

2003-01-01

As the concern on the impact of the operator's inappropriate interventions, so-called Errors Of Commissions(EOCs), on the plant safety has been raised, the interest in the identification and analysis of EOC events from the risk assessment perspective becomes increasing accordingly. To this purpose, we propose a new methodology for identifying and analysing human errors of commission that might be caused from the failures in situation assessment and decision making during accident progressions given an initiating event. The proposed methodology was applied to the accident scenarios of YGN 3 and 4 NPPs, which resulted in about 10 EOC situations that need careful attention

Safety assessment of HLW geological disposal system

International Nuclear Information System (INIS)

Naito, Morimasa

2006-01-01

that Japan is located in a tectonically active zone. Safety assessment for a disposal system differs from that for other engineered systems such as power stations in terms of: Extremely long timescales must be taken into account. Natural environments, which are heterogeneous and cover large spatial areas, must be evaluated. It is thus impossible to apply conventional engineering approaches, where an entire system is constructed and utilized in such a way as to demonstrate system safety. This is a problem specific to the safety assessment of geological disposal. Taking this into account, this paper describes a general methodology of safety assessment for geological system including presentation of a series of steps for the assessment with examples of JNC's H12 safety assessment. (author)

Exposure data and risk indicators for safety performance assessment in Europe.

NARCIS (Netherlands)

Papadimitriou, E. Yannis, G. Bijleveld, F.D. & Cardoso, J.L.

2013-01-01

The objective of this paper is the analysis of the state-of-the-art in risk indicators and exposure data for safety performance assessment in Europe, in terms of data availability, collection methodologies and use. More specifically, the concepts of exposure and risk are explored, as well as the

A study on the assessment of safety culture impacts on risk of nuclear power plants using common uncertainty source model

International Nuclear Information System (INIS)

Lee, Yong Suk; Bang, Young Suk; Chung, Chang Hyun; Jeong, Ji Hwan

2004-01-01

Since International Safety Advisory Group (INSAG) introduced term 'safety culture', it has been widely recognized that safety culture has an important role in safety of nuclear power plants. Research on the safety culture can be divided in the following two parts. 1) Assessment of safety culture (by interview, questionnaire, etc.) 2) Assessment of link between safety culture and safety of nuclear power plants. There is a substantial body of literature that addresses the first part, but there is much less work that addresses the second part. To address the second part, most work focused on the development of model incorporating safety culture into Probabilistic Safety Assessment (PSA). One of the most advanced methodology in the area of incorporating safety culture quantitatively into PSA is System Dynamics (SD) model developed by Kwak et al. It can show interactions among various factors which affect employees' productivity and job quality. Also various situations in nuclear power plant can be simulated and time-dependent risk can be recalculated with this model. But this model does not consider minimal cut set (MCS) dependency and uncertainty of risk. Another well-known methodology is Work Process Analysis Model (WPAM) developed by Davoudian. It considers MCS dependency by modifying conditional probability values using SLI methodology. But we found that the modified conditional probability values in WPAM are somewhat artificial and have no sound basis. WPAM tend to overestimate conditional probability of hardware failure, because it uses SLI methodology which is normally used in Human Reliability Analysis (HRA). WPAM also does not consider uncertainty of risk. In this study, we proposed methodology to incorporate safety culture into PSA quantitatively that can deal with MCS dependency and uncertainty of risk by applying the Common Uncertainty Source (CUS) model developed by Zhang. CUS is uncertainty source that is common to basic events, and this can be physical

NUSS safety standards: A critical assessment

International Nuclear Information System (INIS)

Minogue, R.B.

1985-01-01

The NUSS safety standards are based on systematic review of safety criteria of many countries in a process carefully defined to assure completeness of coverage. They represent an international consensus of accepted safety principles and practices for regulation and for the design, construction, and operation of nuclear power plants. They are a codification of principles and practices already in use by some Member States. Thus, they are not standards which describe methodologies at their present state of evolution as a result of more recent experience and improvements in technological understanding. The NUSS standards assume an underlying body of national standards and a defined technological base. Detailed design and industrial practices vary between countries and the implementation of basic safety standards within countries has taken approaches that conform with national industrial practices. Thus, application of the NUSS standards requires reconciliation with the standards of the country where the reactor will be built as well as with the country from which procurement takes place. Experience in making that reconciliation will undoubtedly suggest areas of needed improvement. After the TMI accident a reassessment of the NUSS programme was made and it was concluded that, given the information at that time and the then level of technology, the basic approach was sound; the NUSS programme should be continued to completion, and the standards should be brought into use. It was also recognized, however, that in areas such as probabilistic risk assessment, human factors methodology, and consideration of detailed accident sequences, more advanced technology was emerging. As these technologies develop, and become more amenable to practical application, it is anticipated that the NUSS standards will need revision. Ideally those future revisions will also flow from experience in their use

The TRIPOD e-learning Platform for the Training of Earthquake Safety Assessment

International Nuclear Information System (INIS)

Coppari, S.; Di Pasquale, G.; Goretti, A.; Papa, F.; Papa, S.; Paoli, G.; Pizza, A. G.; Severino, M.

2008-01-01

The paper summarizes the results of the in progress EU Project titled TRIPOD (Training Civil Engineers on Post-Earthquake Safety Assessment of Damaged Buildings), funded under the Leonardo Da Vinci program. The main theme of the project is the development of a methodology and a learning platform for the training of technicians involved in post-earthquake building safety inspections. In the event of a catastrophic earthquake, emergency building inspections constitute a major undertaking with severe social impact. Given the inevitable chaotic conditions and the urgent need of a great number of specialized individuals to carry out inspections, past experience indicates that inspection teams are often formed in an adhoc manner, under stressful conditions, at a varying levels of technical expertise and experience, sometime impairing the reliability and consistency of the inspection results. Furthermore each Country has its own building damage and safety assessment methodology, developed according to its experience, laws, building technology and seismicity. This holds also for the partners participating to the project (Greece, Italy, Turkey, Cyprus), that all come from seismically sensitive Mediterranean countries. The project aims at alleviating the above shortcomings by designing and developing a training methodology and e-platform, forming a complete training program targeted at inspection engineers, specialized personnel and civil protection agencies. The e-learning platform will provide flexible and friendly authoring mechanisms, self-teaching and assessment capabilities, course and trainee management, etc. Courses will be also made available as stand-alone multimedia applications on CD and in the form of a complete pocket handbook. Moreover the project will offer the possibility of upgrading different experiences and practices: a first step towards the harmonization of methodologies and tools of different Countries sharing similar problems. Finally, through wide

Track 6: safety and risk management. Plant operational risk management. Plant Configuration Risk Assessment Methodology Development for Periodic Maintenance

International Nuclear Information System (INIS)

Yang, Huichang; Chung, Chang Hyun; Sung, Key Yong

2001-01-01

As the operation experiences of nuclear power plants (NPPs) in Korea accumulate and NPP safety functions become enhanced, the role of stable and optimal NPP operation within acceptable safety criteria becomes important at present. To accomplish the goal of safe and optimal operation, maintenance and its related activities should be regarded as the issues of most concern. Studies of methodologies for maintenance improvement and optimization have focused on system performance rather than on the hardware itself. From this point of view, the probabilistic methods are most useful. In terms of risk including core damage frequency and unavailability, the cause that might impact plant safety during normal maintenance activities can be identified and evaluated effectively. The results from these probabilistic analyses can provide insightful information for the reallocation of risk-contributing maintenance activity. This information can be utilized in a way that separates the significant risk-contributing maintenance activities from each other unless they are timely related. In Korea, the risk-monitoring program for operating NPPs is under development and will be implemented in 2003. To accomplish the risk-monitoring program objectives, suitable risk evaluation methods should be developed before the implementation of the risk-monitoring program. The plant configuration assessment methodology was developed for these reasons, and this method is to incorporate the field experiences into the risk calculation exactly within the limit of probabilistic methods. During normal plant operation, the plant operational risk changes frequently depending on the status of the plant system and the arrangement of the components. Specific plant systems or components are typically removed from service because of random equipment failure, planned preventive/predictive maintenance, corrective maintenance, surveillance testing, and operational bypass activities, and such events usually impact the

A study on safety analysis methodology in spent fuel dry storage facility

Energy Technology Data Exchange (ETDEWEB)

Che, M. S.; Ryu, J. H.; Kang, K. M.; Cho, N. C.; Kim, M. S. [Hanyang Univ., Seoul (Korea, Republic of)

2004-02-15

Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology.

Safety assessment of the disposal of sealed radiation sources in boreholes

International Nuclear Information System (INIS)

Oliveira, Rosana Lagua de; Vicente, Roberto; Hiromoto, Goro

2009-01-01

The Radioactive Waste Management Laboratory (RNML) at the Nuclear Energy Research Institute (NERI) in Sao Paulo, Brazil, is developing the concept of a repository for disused sealed radiation sources in a deep borehole. Several thousands disused sealed radiation sources are stored at NERI awaiting the decision on final disposal and tens of thousands are still under the possession of the licensees. A significant fraction of these sources are long-lived and will require final disposal in a geological repository. The purpose of this paper is to identify and discuss suitable safety assessment strategies for the repository concept and to illustrate a rational approach for a long-term safety assessment methodology. (author)

Hanford Site baseline risk assessment methodology

International Nuclear Information System (INIS)

1993-03-01

This methodology has been developed to prepare human health and environmental evaluations of risk as part of the Comprehensive Environmental Response, Compensation, and Liability Act remedial investigations (RIs) and the Resource Conservation and Recovery Act facility investigations (FIs) performed at the Hanford Site pursuant to the Hanford Federal Facility Agreement and Consent Order referred to as the Tri-Party Agreement. Development of the methodology has been undertaken so that Hanford Site risk assessments are consistent with current regulations and guidance, while providing direction on flexible, ambiguous, or undefined aspects of the guidance. The methodology identifies Site-specific risk assessment considerations and integrates them with approaches for evaluating human and environmental risk that can be factored into the risk assessment program supporting the Hanford Site cleanup mission. Consequently, the methodology will enhance the preparation and review of individual risk assessments at the Hanford Site

Development of Audit Calculation Methodology for RIA Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Joosuk; Kim, Gwanyoung; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

The interim criteria contain more stringent limits than previous ones. For example, pellet-to-cladding mechanical interaction(PCMI) was introduced as a new failure criteria. And both short-term (e.g. fuel-to coolant interaction, rod burst) and long-term(e.g., fuel rod ballooning, flow blockage) phenomena should be addressed for core coolability assurance. For dose calculations, transient-induced fission gas release has to be accounted additionally. Traditionally, the approved RIA analysis methodologies for licensing application are developed based on conservative approach. But newly introduced safety criteria tend to reduce the margins to the criteria. Thereby, licensees are trying to improve the margins by utilizing a less conservative approach. In this situation, to cope with this trend, a new audit calculation methodology needs to be developed. In this paper, the new methodology, which is currently under developing in KINS, was introduced. For the development of audit calculation methodology of RIA safety analysis based on the realistic evaluation approach, preliminary calculation by utilizing the best estimate code has been done on the initial core of APR1400. Followings are main conclusions. - With the assumption of single full-strength control rod ejection in HZP condition, rod failure due to PCMI is not predicted. - And coolability can be assured in view of entalphy and fuel melting. - But, rod failure due to DNBR is expected, and there is possibility of fuel failure at the rated power conditions also.

Safety assessment of computerized instrumentation and control for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Fride, B.; Henry, J.Y.; Manners, S.

1996-12-31

France`s latest 1400 MWe `N4` generation of Pressurised Water Reactors (PWR) use distributed programmable control systems interconnected by data networks. The protection system is also software based. IPSN have the task of evaluating the safety demonstration before the government safety authority (DSIN) give the licensee (EDF) permission to fuel the reactor and to raise power. Some of the different aspects of the evaluation carried out and the methodologies used for assessing the C and I are presented. (author). 3 refs.

Assessment of the factors with significant influence on safety culture

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.

2013-01-01

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

An Assessment of SKB's Performance Assessment Calculations in the Interim Main Report for the Safety Assessment SR-Can

International Nuclear Information System (INIS)

Maul, Philip; Robinson, Peter

2005-03-01

SKB have published their Interim Main Report of the safety assessment SR-Can, which is intended to establish the framework for what will be submitted in 2006 in support of a licence application for construction of the spent fuel encapsulation plant. This follows on from the SR-Can Planning Document published in 2003. The purpose of the Interim Report is stated to be to demonstrate the methodology that will be used for safety assessment. The present report evaluates the information provided in the Interim SR-Can Report that is relevant to the Performance Assessment (PA) calculations that SKB intend to undertake, using independent calculations to facilitate this process. SKB consider that the primary safety function is to isolate completely the fuel within the canisters over the entire assessment period. Should a canister be damaged, the secondary safety function is to ensure that any release is retarded and dispersed sufficiently to ensure that concentrations levels in the accessible environment cannot cause unacceptable consequences. In this report PA calculations are considered to include both a high-level representation of the evolution of the system (relevant to the primary safety function), and any subsequent radionuclide transport (relevant to the secondary safety function). The main conclusions drawn are: 1. The effects of climate evolution on engineered barriers have not been analysed in detail in the Interim Report, and this limits the usefulness of the preliminary calculations that have been undertaken. 2. A key aspect of SKB's approach is the use of an integrated near-field evolution model. The information provided on this model demonstrates its capability efficiently to reproduce calculations from individual process models, but insufficient information is given at the present time to justify statements about interactions between processes. In particular it is assumed that relatively short term thermal and resaturation processes do not affect the

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)

Performance assessment and the safety case: Lessons from recent international projects and areas for further development

International Nuclear Information System (INIS)

Galson, Daniel A.; Bailey, Lucy

2014-01-01

The European Commission (EC) PAMINA project - Performance Assessment Methodologies in Application to Guide the Development of the Safety Case - was conducted over the period 2006-2009 and brought together 27 organisations from 10 countries. PAMINA had the aim of improving and developing a common understanding of performance assessment (PA) methodologies for disposal concepts for spent fuel and other long-lived radioactive wastes in a range of geological environments. This was followed by a Nuclear Energy Agency (NEA) sponsored project on Methods for Safety Assessment of Geological Disposal Facilities for Radioactive Waste (MeSA), which was completed in 2012. This paper presents a selection of conclusions from these projects, in the context of general understanding developed on what would constitute an acceptable safety case for a geological disposal facility, and outlines areas for further development. The paper also introduces a new project on PA that is under consideration within the context of the EC Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP). (authors)

System assessment using modular logic fault tree methodology

International Nuclear Information System (INIS)

Troncoso Fleitas, M.

1996-01-01

In the process of a Probabilistic Safety analysis (PSA) study a large number of fault trees are generated by different specialist. Modular Logic Fault Tree Methodology pave the way the way to systematize the procedures and to unify the criteria in the process of systems modulation. An example of of the application of this methodology is shown

IRIS guidelines. 2014 ed. Integrated Review of Infrastructure for Safety (IRIS) for self-assessment when establishing the safety infrastructure for a nuclear power programme

International Nuclear Information System (INIS)

2014-01-01

The IAEA safety standards reflect an international consensus on what constitutes a high level of safety for protecting people and the environment, and therefore represent what all Member States should achieve, whilst recognizing the ultimate responsibility of each State to ensure safety when implementing a nuclear power programme. IAEA Safety Standards Series No. SSG-16, entitled Establishing the Safety Infrastructure for a Nuclear Power Programme was published in order to provide recommendations, presented in the form of sequential actions, on meeting safety requirements progressively during the initial three phases of the development of safety, as described in INSAG-22, Nuclear Safety Infrastructure for a National Nuclear Power Programme Supported by the IAEA Fundamental Safety Principles. To that end, the 200 safety related actions, which are proposed by SSG-16, constitute a roadmap to establish a foundation for promoting a high level of safety over the entire lifetime of the nuclear power plant. These actions reflect international consensus on good practice in order to achieve full implementation of IAEA safety standards. The IAEA has developed a methodology and tool, the Integrated Review of Infrastructure for Safety (IRIS), to assist States in undertaking self-assessment with respect to SSG-16 recommendations when establishing the safety infrastructure for a nuclear power programme, and to develop an action plan for improvement. The IRIS methodology and the associated tool are fully compatible with the IAEA safety standards and are also used, when appropriate, in the preparation of review missions, such as the Integrated Regulatory Review Service and advisory missions. The present guidelines describe the IRIS methodology for self-assessment against SSG-16 recommendations. Through IRIS implementation, every organization concerned with nuclear safety may gain proper awareness and engage in a continuous progressive process to develop the effective national

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

A consistent approach to assess safety criteria for reactivity initiated accidents

International Nuclear Information System (INIS)

Sartoris, C.; Taisne, A.; Petit, M.; Barre, F.; Marchand, O.

2010-01-01

In the context of more and more demanding reactor managements, the fuel assembly discharge burn-up increases and raises the question of the current safety criteria relevance. In order to assess new safety criteria for reactivity initiated accidents, the IRSN is developing a consistent and original approach to assess safety. This approach is based on: -A thorough understanding of the physical mechanisms involved in each phase (PCMI and post-boiling phases) of the RIA, supported by the interpretation of the experimental database. This experimental data is constituted of global test outcomes, such as CABRI or Nuclear Safety Research Reactor (NSRR) experiments, and analytical program outcomes, such as PATRICIA tests, intending to understand some particular physical phenomena; -The development of computing codes, modelling the physical phenomena. The physical phenomena observed during the tests mentioned above were modelled in the SCANAIR code. SCANAIR is a thermal-mechanical code calculating fuel and clad temperatures and strains during RIA. The CLARIS module is used as a post-calculation tool to evaluate the clad failure risk based on critical flaw depth. These computing codes were validated by global and analytical tests results; -The development of a methodology. The first step of this methodology is the identification of all the parameters affecting the hydride rim depth. Besides, an envelope curve resulting from burst tests giving the hydride rim depth versus oxidation thickness is defined. After that, the critical flaw depth for a given energy pulse is calculated then compared to the hydride rim depth. This methodology results in an energy or enthalpy limit versus burn-up. This approach is planned to be followed for each phase of the RIA. An example of application is presented to evaluate a PCMI limit for a zircaloy-4 cladding UO 2 rod at Hot Zero Power.

Development of assessment methodology for plant configuration control

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Kim, Yoon Ik; Yang, Hui Chang; Huh, Byeong Gill; Lee, Dong Won; Ahn, Gwan Won [Seoul National Univ., Seoul (Korea, Republic of)

2001-03-15

The purpose of this study IS the development of effective and overall assessment methodology which reflects the characteristics of plants for the surveillance, maintenance, repair and operation of nuclear power plants. In this study, recent researches are surveyed and concept definition, procedures, current PSA methodologies, implementation of various models are evaluated. Through this survey, systematic assessment methodology is suggested. Configuration control assessment methodology suggested in this study for the purpose of the development of configuration control methodology reflecting the characteristics of Korean NPPs, can be utilized as the supplement of current PSA methodologies.

Development of probabilistic assessment methodology for geologic disposal of radioactive wastes

International Nuclear Information System (INIS)

Kimura, H.; Takahashi, T.

1998-01-01

The probabilistic assessment methodology is essential to evaluate uncertainties of long-term radiological consequences associated with geologic disposal of radioactive wastes. We have developed a probabilistic assessment methodology to estimate the influences of parameter uncertainties/variabilities. An exposure scenario considered here is based on a groundwater migration scenario. A computer code system GSRW-PSA thus developed is based on a non site-specific model, and consists of a set of sub modules for sampling of model parameters, calculating the release of radionuclides from engineered barriers, calculating the transport of radionuclides through the geosphere, calculating radiation exposures of the public, and calculating the statistical values relating the uncertainties and sensitivities. The results of uncertainty analyses for α-nuclides quantitatively indicate that natural uranium ( 238 U) concentration is suitable for an alternative safety indicator of long-lived radioactive waste disposal, because the estimated range of individual dose equivalent due to 238 U decay chain is narrower that that due to other decay chain ( 237 Np decay chain). It is internationally necessary to have detailed discussion on the PDF of model parameters and the PSA methodology to evaluated the uncertainties due to conceptual models and scenarios. (author)

Methodology and analysis of production safety during Pu recycling at SSC RF RIAR

International Nuclear Information System (INIS)

Kirillovich, A.P.

2000-01-01

The methodology and criteria for estimating safety in technological processes of the nuclear fuel cycle (NFC) are proposed, substantiated and verified during the large-scale Pu recycling (500 kg). The comprehensive investigation results of the radiation-ecological situation are presented during pilot production of the mixed uranium-plutonium fuel and fuel assembly at SSC RF RIAR. The methodology and experimental data bank can be used while estimating safety in the industrial recycling of Pu and minor-actinides (Np, Am, Cm) in NFC. (author)

Scientific Opinion on a Qualified Presumption of Safety (QPS) approach for the safety assessment of botanicals and botanical preparations

DEFF Research Database (Denmark)

Pilegaard, Kirsten

preparation for QPS status, it has been possible to develop a structured assessment scheme that provides a practical method for assessing botanicals and botanical preparations for which an adequate body of knowledge exists and therefore without the need for further testing. Reiterative applications...... in the development of a comprehensive, systematic and transparent methodology. The Scientific Committee recommends its use as an extension of the 2009 EFSA guidance for the safety assessment of botanicals and botanical preparations intended to be used in food supplements....

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

Energy Technology Data Exchange (ETDEWEB)

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01

Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

Human error risk management for engineering systems: a methodology for design, safety assessment, accident investigation and training

International Nuclear Information System (INIS)

Cacciabue, P.C.

2004-01-01

The objective of this paper is to tackle methodological issues associated with the inclusion of cognitive and dynamic considerations into Human Reliability methods. A methodology called Human Error Risk Management for Engineering Systems is presented that offers a 'roadmap' for selecting and consistently applying Human Factors approaches in different areas of application and contains also a 'body' of possible methods and techniques of its own. Two types of possible application are discussed to demonstrate practical applications of the methodology. Specific attention is dedicated to the issue of data collection and definition from specific field assessment

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

The selection of probabilistic safety assessment techniques for non-reactor nuclear facilities

International Nuclear Information System (INIS)

Vail, J.

1992-01-01

Historically, the probabilistic safety assessment (PSA) methodology of choice is the well known event tree/fault tree inductive technique. For reactor facilities is has stood the test of time. Some non-reactor nuclear facilities have found inductive methodologies difficult to apply. The stand-alone fault tree deductive technique has been used effectively to analyze risk in nuclear chemical processing facilities and waste handling facilities. The selection between the two choices suggest benefits from use of the deductive method for non-reactor facilities

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

International Nuclear Information System (INIS)

Crossland, I.G.; Torres, C.

2002-01-01

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

Development of probabilistic methods for safety assessment of nuclear power plants

International Nuclear Information System (INIS)

Schott, H.; Berg, H.P.

1998-01-01

Since its introduction by the German Risk Study, Probabilistic Safety Assessment (PSA) has developed in Germany to a valuable tool in regulatory decision-making. Plant specific PSAs of Level 1+ are now conducted for all nuclear power plants in the frame of Periodic Safety Reviews. This paper is devoted to the description or key elements set out in the regulatory guidelines for PSA-Level 1+ and the corresponding technical documents and the further development of PSA methodology in the Federal Republic of Germany. In the course of the next years it is intended to make progress in the modeling of common cause failures, human reliability evaluation, reduction of uncertainties in PSA modeling techniques and data estimation, analysis of low power and shut down states as well as in reaching a mature methodology for inclusion of external events into the analysis. (author)

Assessment of Integrated Pedestrian Protection Systems with Autonomous Emergency Braking (AEB) and Passive Safety Components.

Science.gov (United States)

Edwards, Mervyn; Nathanson, Andrew; Carroll, Jolyon; Wisch, Marcus; Zander, Oliver; Lubbe, Nils

2015-01-01

Autonomous emergency braking (AEB) systems fitted to cars for pedestrians have been predicted to offer substantial benefit. On this basis, consumer rating programs-for example, the European New Car Assessment Programme (Euro NCAP)-are developing rating schemes to encourage fitment of these systems. One of the questions that needs to be answered to do this fully is how the assessment of the speed reduction offered by the AEB is integrated with the current assessment of the passive safety for mitigation of pedestrian injury. Ideally, this should be done on a benefit-related basis. The objective of this research was to develop a benefit-based methodology for assessment of integrated pedestrian protection systems with AEB and passive safety components. The method should include weighting procedures to ensure that it represents injury patterns from accident data and replicates an independently estimated benefit of AEB. A methodology has been developed to calculate the expected societal cost of pedestrian injuries, assuming that all pedestrians in the target population (i.e., pedestrians impacted by the front of a passenger car) are impacted by the car being assessed, taking into account the impact speed reduction offered by the car's AEB (if fitted) and the passive safety protection offered by the car's frontal structure. For rating purposes, the cost for the assessed car is normalized by comparing it to the cost calculated for a reference car. The speed reductions measured in AEB tests are used to determine the speed at which each pedestrian in the target population will be impacted. Injury probabilities for each impact are then calculated using the results from Euro NCAP pedestrian impactor tests and injury risk curves. These injury probabilities are converted into cost using "harm"-type costs for the body regions tested. These costs are weighted and summed. Weighting factors were determined using accident data from Germany and Great Britain and an independently

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.

Assessment of methodologies for radioactive waste management

International Nuclear Information System (INIS)

Hoos, I.R.

1978-01-01

No quantitative methodology is adequate to encompass and assess all the risks, no risk/benefit calculation is fine-tuned enough to supply decision-makers with the full range and all of the dimensions. Quality assurance cannot be conceived in terms of systems design alone, but must be maintained vigilantly and with integrity throughout the process. The responsibility of the NRC is fairly well established with respect to overall reactor safety. With respect to the management of radioactive wastes, its mission is not yet so clearly delineated. Herein lies a challenge and an opportunity. Where the known quantitative methodologies are restrictive and likely to have negative feedback effect on authority and public support, the broader lens and the bolder thrust are called for. The cozy cocoon of figures ultimately protects no one. The Commission, having acknowledged that the management of radioactive wastes is not merely a technological matter can now take the socially responsible position of exploring as fully and confronting as candidly as possible the total range of dimensions involved. Paradoxically, it is Charles J. Hitch, intellectual progenitor of the methodology, who observes that we may be missing the meaning of his message by relying too heavily on quantitative analysis and thus defining our task too narrowly. We live in a closed system, in which science and technology, politics and economics, and, above all, social and human elements interact, sometimes to create the problems, sometimes to articulate the questions, and sometimes to find viable solutions

Safety-related operator actions: methodology for developing criteria

International Nuclear Information System (INIS)

Kozinsky, E.J.; Gray, L.H.; Beare, A.N.; Barks, D.B.; Gomer, F.E.

1984-03-01

This report presents a methodology for developing criteria for design evaluation of safety-related actions by nuclear power plant reactor operators, and identifies a supporting data base. It is the eleventh and final NUREG/CR Report on the Safety-Related Operator Actions Program, conducted by Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The operator performance data were developed from training simulator experiments involving operator responses to simulated scenarios of plant disturbances; from field data on events with similar scenarios; and from task analytic data. A conceptual model to integrate the data was developed and a computer simulation of the model was run, using the SAINT modeling language. Proposed is a quantitative predictive model of operator performance, the Operator Personnel Performance Simulation (OPPS) Model, driven by task requirements, information presentation, and system dynamics. The model output, a probability distribution of predicted time to correctly complete safety-related operator actions, provides data for objective evaluation of quantitative design criteria

A progressive methodology for seismic safety evaluation of gravity dams

International Nuclear Information System (INIS)

Ghrib, F.; Leger, P.; Tinawi, R.; Lupien, R.; Veilleux, M.

1995-01-01

A progressive methodology for the seismic safety evaluation of existing concrete gravity dams was described. The methodology was based on five structural analysis levels with increasing complexity to represent inertia forces, dam-foundation and dam-interaction mechanisms, as well as concrete cracking. The five levels were (1) preliminary screening, (2) pseudo-static method, (3) pseudo-dynamic method, (4) linear time history analysis, and (5) non-linear history analysis. The first four levels of analysis were applied for the seismic safety evaluation of Paugan gravity dam (Quebec). Results showed that internal forces from pseudo-dynamic, response spectra and transient finite element analyses could be used to interpret the dynamic stability of dams from familiar strength-based criteria. However, as soon as the base was cracked, the seismically induced forces were modified, and level IV analyses proved more suitable to handle rationally these complexities. 8 refs., 7 figs., 1 tab

The DYLAM approach to systems safety and reliability assessment

International Nuclear Information System (INIS)

Amendola, A.

1988-01-01

A survey of the principal features and applications of DYLAM (Dynamic Logical Analytical Methodology) is presented, whose basic principles can be summarized as follows: after a particular modelling of the component states, computerized heuristical procedures generate stochastic configurations of the system, whereas the resulting physical processes are simultaneously simulated to give account of the possible interactions between physics and states and, on the other hand, to search for system dangerous configurations and related probabilities. The association of probabilistic techniques for describing the states with physical equations for describing the process results in a very powerful tool for safety and reliability assessment of systems potentially subjected to dangerous incidental transients. A comprehensive picture of DYLAM capability for manifold applications can be obtained by the review of the study cases analyzed (LMFBR core accident, systems reliability assessment, accident simulation, man-machine interaction analysis, chemical reactors safety, etc.)

Defining safety culture and the nexus between safety goals and safety culture. 3. A Methodology for Identifying Deficiencies in Safety Culture

International Nuclear Information System (INIS)

Apostolakis, George; Weil, Rick

2001-01-01

At present, the drivers of performance problems at nuclear power plants (NPPs) are organizational in nature. Organizational deficiencies and other 'latent' conditions cause human errors, resulting in incidents that impact the performance of NPPs. Therefore, the human reliability community, regulators, and others concerned with NPP safety express the view that safety culture and organizational factors play an important role in plant safety. However, we have yet to identify one complete set of organizational factors, establish links between deficient safety culture and performance, or develop adequate tools to measure safety culture. This paper will contribute to the resolution of these issues. Safety culture is not a single factor but rather is a collection of several distinct factors. This paper asserts that in order to pro-actively manage safety culture at NPPs, leading indicators and appropriate measurements must be identified and developed. Central to this effort are the identification of the distinct factors comprising safety culture and the relationships between those factors and performance. We have identified several factors important to safety culture. We have developed a methodology that is a combination of traditional root-cause analysis and theories of human error, most notably Reason's theory of accident causation. In addition to this methodology's usefulness in identifying deficiencies in safety culture, it could also be used as a starting point to identify leading indicators of deteriorating safety performance. We have identified six organizational factors as being important: communication, formalization, goal prioritization, problem identification, roles and responsibilities, and technical knowledge. In addition, we have found that certain organizational factors, although pervasive throughout the organization, have a much greater influence on the successful outcome of particular tasks of work processes, rather than being equally important to all

A Reliability Assessment Method for the VHTR Safety Systems

International Nuclear Information System (INIS)

Lee, Hyung Sok; Jae, Moo Sung; Kim, Yong Wan

2011-01-01

The Passive safety system by very high temperature reactor which has attracted worldwide attention in the last century is the reliability safety system introduced for the improvement in the safety of the next generation nuclear power plant design. The Passive system functionality does not rely on an external source of energy, but on an intelligent use of the natural phenomena, such as gravity, conduction and radiation, which are always present. Because of these features, it is difficult to evaluate the passive safety on the risk analysis methodology having considered the existing active system failure. Therefore new reliability methodology has to be considered. In this study, the preliminary evaluation and conceptualization are tried, applying the concept of the load and capacity from the reliability physics model, designing the new passive system analysis methodology, and the trial applying to paper plant.

Development of a Methodology for VHTR Accident Consequence Assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-05-15

The substitution of the VHTR for burning fossil fuels conserves these hydrocarbon resources for other uses and eliminates the emissions of greenhouse. In Korea, for these reasons, constructing the VHTR plan for hydrogen production is in progress. In this study, the consequence analysis for the off-site releases of radioactive materials during severe accidents has been performed using the level 3 PRA technology. The offsite consequence analysis for a VHTR using the MACCS code has been performed. Since the passive system such as the RCCS(Reactor Cavity Cooling System) are equipped, the frequency of occurrence of accidents has been evaluated to be very low. For further study, the assessment for characteristic of VHTR safety system and precise quantification of its accident scenarios is expected to conduct more certain consequence analysis. This methodology shown in this study might contribute to enhancing the safety of VHTR design by utilizing the results having far lower effect on the environment than the LWRs.

LFR safety approach and main ELFR safety analysis results

International Nuclear Information System (INIS)

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

2013-01-01

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

The 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

Safety assessment of multi-unit NPP sites subject to external events

International Nuclear Information System (INIS)

Samaddar, Sujit; Hibino, Kenta; Coman, Ovidiu

2014-01-01

This paper presents a framework for conducting a probabilistic safety assessment of multi-unit sites against external events. The treatment of multiple hazard on a unit, interaction between units, implementation of severe accident measures, human reliability, environmental conditions, metric of risk for both reactor and non-reactor sources, integration of risk and responses and many such important factors need to be addressed within the context of this framework. The framework facilitates the establishment of a comprehensive methodology that can be applied internationally to the peer review of safety assessment of multi-unit sites under the impact of multiple external hazards. In summary, it can be said that the site safety assessment for a multi-unit site will be quite complex and need to start with individual unit risk assessments, these need to be combined considering the interactions between units and their responses, and the fragilities of the installations established considering the combined demands from all interactions. Using newly established risk metric the risk can then be integrated for the overall site. Fig. 2 shows schematically such a proposal. Much work has to done and the IAEA has established a working group that is systematically establishing the structure and process to incorporate the many issues that are a part of a multi-unit site safety assessment. (authors)

Safety Assessment of Multi Purpose Small Payload Rack(MSPR)

Science.gov (United States)

Mizutani, Yoshinobu; Takada, Satomi; Murata, Kosei; Ozawa, Daisaku; Kobayashi, Ryoji; Nakamura, Yasuhiro

2010-09-01

We are reporting summary of preliminary safety assessment for Multi Purpose Small Payload Rack(MSPR), which is one of the micro gravity experiment facilities that are being developed for the 2nd phase JEM utilization(JEM: Japanese Experiment Module) that will be launched on H-II Transfer Vehicle(HTV) 2nd flight in 2011. MSPR is used for multi-purpose micro-g experiment providing experimental spaces and work stations. MSPR has three experimental spaces; first, there is a space called Work Volume(WV) with capacity volume of approximately 350 litters, in which multiple resources including electricity, communication, and moving image functions can be used. Within this space, installation of devices can be done by simple, prompt attachment by Velcro and pins with high degree of flexibility. Second, there is Small Experiment Area(SEA), with capacity volume of approximately 70 litters, in which electricity, communication, and moving image functions can also be used in the same way as WV. These spaces protect experiment devices and specimens from contingent loads by the crewmembers. Third, there is Work Bench with area of 0.5 square meters, on which can be used for maintenance, inspection and data operations of installed devices, etc. This bench can be stored in the rack during contingency. Chamber for Combustion Experiment(CCE) that is planned to be installed in WV is a pressure-resistant experimental container that can be used to seal hazardous materials from combustion experiments. This CCE has double sealing design in chamber itself, which resist gas leakage under normal the temperature and pressure. Electricity, communication, moving image function can be used in the same way as WV. JAXA Phase 2 Safety Review Panel(SRP) has been held in April, 2010. For safety analysis of MSPR, hazards were identified based on Fault Tree Analysis methodology and then these hazards were classified into either eight ISS standard-type hazards or eight unique-type hazards that requires

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

scope of the quantitative safety assessment. These arguments include: Support from natural and anthropogenic analogues for both key process understanding and total system performance. Comparison of the methodology and results with the earlier TILA-99 and SR-Can safety assessments, as well as other international safety assessments, to ensure completeness, consistency and reasonableness of the present assessment. Use of safety indicators other than dose and activity to avoid uncertainties in future human lifestyles and also in geological processes on very long timescales. Consideration of the calculation results from a wider perspective to consider significance of their impact compared to other risks

Methodological quality of meta-analyses on treatments for chronic obstructive pulmonary disease: a cross-sectional study using the AMSTAR (Assessing the Methodological Quality of Systematic Reviews) tool.

Science.gov (United States)

Ho, Robin S T; Wu, Xinyin; Yuan, Jinqiu; Liu, Siya; Lai, Xin; Wong, Samuel Y S; Chung, Vincent C H

2015-01-08

Meta-analysis (MA) of randomised trials is considered to be one of the best approaches for summarising high-quality evidence on the efficacy and safety of treatments. However, methodological flaws in MAs can reduce the validity of conclusions, subsequently impairing the quality of decision making. To assess the methodological quality of MAs on COPD treatments. A cross-sectional study on MAs of COPD trials. MAs published during 2000-2013 were sampled from the Cochrane Database of Systematic Reviews and Database of Abstracts of Reviews of Effect. Methodological quality was assessed using the validated AMSTAR (Assessing the Methodological Quality of Systematic Reviews) tool. Seventy-nine MAs were sampled. Only 18% considered the scientific quality of primary studies when formulating conclusions and 49% used appropriate meta-analytic methods to combine findings. The problems were particularly acute among MAs on pharmacological treatments. In 48% of MAs the authors did not report conflict of interest. Fifty-eight percent reported harmful effects of treatment. Publication bias was not assessed in 65% of MAs, and only 10% had searched non-English databases. The methodological quality of the included MAs was disappointing. Consideration of scientific quality when formulating conclusions should be made explicit. Future MAs should improve on reporting conflict of interest and harm, assessment of publication bias, prevention of language bias and use of appropriate meta-analytic methods.

Developing new methodology for nuclear power plants vulnerability assessment

International Nuclear Information System (INIS)

Kostadinov, Venceslav

2011-01-01

Research highlights: → Paper presents new methodology for vulnerability assessment of nuclear power plants. → First universal quantitative risks assessment model for terrorist attack on a NPPs. → New model enhance security, reliability and safe operation of all energy infrastructure. → Significant research benefits: increased NPPs security, reliability and availability. → Useful new tool for PRA application to evaluation of terrorist threats on NPPs. - Abstract: The fundamental aim of an efficient regulatory emergency preparedness and response system is to provide sustained emergency readiness and to prevent emergency situations and accidents. But when an event occurs, the regulatory mission is to mitigate consequences and to protect people and the environment against nuclear and radiological damage. The regulatory emergency response system, which would be activated in the case of a nuclear and/or radiological emergency and release of radioactivity to the environment, is an important element of a comprehensive national regulatory system of nuclear and radiation safety. In the past, national emergency systems explicitly did not include vulnerability assessments of the critical nuclear infrastructure as an important part of a comprehensive preparedness framework. But after the huge terrorist attack on 11/09/2001, decision-makers became aware that critical nuclear infrastructure could also be an attractive target to terrorism, with the purpose of using the physical and radioactive properties of the nuclear material to cause mass casualties, property damage, and detrimental economic and/or environmental impacts. The necessity to evaluate critical nuclear infrastructure vulnerability to threats like human errors, terrorist attacks and natural disasters, as well as preparation of emergency response plans with estimation of optimized costs, are of vital importance for assurance of safe nuclear facilities operation and national security. In this paper presented

An Integrated Approach for Characterization of Uncertainty in Complex Best Estimate Safety Assessment

International Nuclear Information System (INIS)

Pourgol-Mohamad, Mohammad; Modarres, Mohammad; Mosleh, Ali

2013-01-01

This paper discusses an approach called Integrated Methodology for Thermal-Hydraulics Uncertainty Analysis (IMTHUA) to characterize and integrate a wide range of uncertainties associated with the best estimate models and complex system codes used for nuclear power plant safety analyses. Examples of applications include complex thermal hydraulic and fire analysis codes. In identifying and assessing uncertainties, the proposed methodology treats the complex code as a 'white box', thus explicitly treating internal sub-model uncertainties in addition to the uncertainties related to the inputs to the code. The methodology accounts for uncertainties related to experimental data used to develop such sub-models, and efficiently propagates all uncertainties during best estimate calculations. Uncertainties are formally analyzed and probabilistically treated using a Bayesian inference framework. This comprehensive approach presents the results in a form usable in most other safety analyses such as the probabilistic safety assessment. The code output results are further updated through additional Bayesian inference using any available experimental data, for example from thermal hydraulic integral test facilities. The approach includes provisions to account for uncertainties associated with user-specified options, for example for choices among alternative sub-models, or among several different correlations. Complex time-dependent best-estimate calculations are computationally intense. The paper presents approaches to minimize computational intensity during the uncertainty propagation. Finally, the paper will report effectiveness and practicality of the methodology with two applications to a complex thermal-hydraulics system code as well as a complex fire simulation code. In case of multiple alternative models, several techniques, including dynamic model switching, user-controlled model selection, and model mixing, are discussed. (authors)

The JAERI program for development of safety assessment models and acquisition of data needed for assessment of geological disposal of high-level radioactive wastes

International Nuclear Information System (INIS)

Matsuzuru, H.

1991-01-01

The JAERI is conducting R and D program for the development of safety assessment methodologies and the acquisition of data needed for the assessment of geologic disposal of high-level radioactive wastes, aiming at the elucidation of feasibility of geologic disposal in Japan. The paper describes current R and D activities to develop interim versions of both a deterministic and a probabilistic methodologies based on a normal evolution scenario, to collect data concerning engineered barriers and geologic media through field and laboratory experiments, and to validate the models used in the methodologies. 2 figs., 2 refs

MODELS AND METHODS OF SAFETY-ORIENTED PROJECT MANAGEMENT OF DEVELOPMENT OF COMPLEX SYSTEMS: METHODOLOGICAL APPROACH

Directory of Open Access Journals (Sweden)

Олег Богданович ЗАЧКО

2016-03-01

Full Text Available The methods and models of safety-oriented project management of the development of complex systems are proposed resulting from the convergence of existing approaches in project management in contrast to the mechanism of value-oriented management. A cognitive model of safety oriented project management of the development of complex systems is developed, which provides a synergistic effect that is to move the system from the original (pre condition in an optimal one from the viewpoint of life safety - post-project state. The approach of assessment the project complexity is proposed, which consists in taking into account the seasonal component of a time characteristic of life cycles of complex organizational and technical systems with occupancy. This enabled to take into account the seasonal component in simulation models of life cycle of the product operation in complex organizational and technical system, modeling the critical points of operation of systems with occupancy, which forms a new methodology for safety-oriented management of projects, programs and portfolios of projects with the formalization of the elements of complexity.

Assessing digital control system dependability using the dynamic flowgraph methodology

International Nuclear Information System (INIS)

Garrett, C.J.; Guarro, S.B.; Apostolakis, G.E.

1993-01-01

Dynamic Flowgraph Methodology (DFM) is a methodological approach to modeling and analyzing the behavior of software-driven embedded systems for the purpose of reliability/safety assessment and verification. The methodology has two fundamental goals: (a) to identify how certain postulated events may occur in a system and (b) to identify an appropriate testing strategy based on an analysis of system functional behavior. To achieve these goals, the methodology employs a modeling framework in which system models are developed in terms of causal relationships between physical variables and temporal characteristics of the execution of software modules. These models are then analyzed to determine how a certain state (desirable or undesirable) can be reached. This is done by developing timed fault trees, which take the form of logical combinations of static trees relating system parameters at different points in time. The prime implicants (multistate analog of minimal cut sets) of the fault trees can be used to identify and eliminate system faults resulting from unanticipated combinations of software logic errors, hardware failures, and adverse environmental conditions and to direct testing activity to more efficiently eliminate implementation errors by focusing on the neighborhood of potential failure modes arising from these combinations of system conditions

Assessment and management of aging of nuclear power plant safety-related structures

International Nuclear Information System (INIS)

Naus, D.J.; Graves, H.L. III; Ellingwood, B.R.

2003-01-01

Background information and data have been developed for improving existing and developing new methods to assist in quantifying the effects of age-related degradation on the performance of nuclear power plant (NPP) safety-related structures. Factors that can lead to age-related degradation of safety-related structures are identified and their manifestations described. Current regulatory testing and inspection requirements are reviewed and a summary of degradation experience presented. Techniques commonly used to inspect NPP concrete structures to assess and quantify age-related degradation are summarized. An approach for conduct of condition assessments of structures in NPPs is presented. Criteria, based primarily on visual indications, are provided for use in classification and assessment of concrete degradation. Materials and techniques for repair of degraded structures are noted and guidance provided on repair options available for various forms of degradation. A probabilistic methodology for condition assessment and reliability-based life prediction has been developed and applied to structures subject to combinations of structural load processes and to structural systems. The methodology has also been used to investigate optimization of in-service inspection and maintenance strategies to maintain failure probability below a specified target value as well as to minimize costs. Fragility assessments involving analytical solutions and finite-element methods have been utilized to predict the effect of aging degradation on structural component performance. (author)

Waste isolation safety assessment program

International Nuclear Information System (INIS)

Brandstetter, A.; Harwell, M.A.

1979-05-01

Associated with commercial nuclear power production in the United States is the generation of potentially hazardous radioactive wastes. The Department of Energy (DOE), through the National Waste Terminal Storage (NWTS) Program, is seeking to develop nuclear waste isolation systems in geologic formations that will preclude contact with the biosphere of waste radionuclides in concentrations which are sufficient to cause deleterious impact on humans or their environments. Comprehensive analyses of specific isolation systems are needed to assess the expectations of meeting that objective. The Waste Isolation Safety Assessment Program (WISAP) has been established at the Pacific Northwest Laboratory (operated by Battelle Memorial Institute) for developing the capability of making those analyses. Among the analyses required for isolation system evaluation is the detailed assessment of the post-closure performance of nuclear waste repositories in geologic formations. This assessment is essential, since it is concerned with aspects of the nuclear power program which previously have not been addressed. Specifically, the nature of the isolation systems (e.g., involving breach scenarios and transport through the geosphere), and the time-scales necessary for isolation, dictate the development, demonstration and application of novel assessment capabilities. The assessment methodology needs to be thorough, flexible, objective, and scientifically defensible. Further, the data utilized must be accurate, documented, reproducible, and based on sound scientific principles

Level 1 and 2 PSA methodology taking into account new design, operating and safety factors. Rev. 1

International Nuclear Information System (INIS)

Jirsa, P.; Patrik, M.

2000-11-01

The status of probabilistic safety assessment (PSA) is discussed (i) in relation to the expected nature of 'revolutionary' innovations and (ii) in the light of the EUR document, summarizing requirements put by European NPP operators on the future NPP design. The aims included: (1) analysis of limitations to the current PSA methodology; (2) specification of physical and operation processes the knowledge of which is necessary to ensure the safety criteria of advanced reactors; (3) summarisation of existing knowledge and description formats of the processes; (4) identification of theoretical and experimental work required to address the problem, preparation of data and computer codes, ensuring traceability to EU developmental programs. (P.A.)

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

USGS Methodology for Assessing Continuous Petroleum Resources

Science.gov (United States)

Charpentier, Ronald R.; Cook, Troy A.

2011-01-01

The U.S. Geological Survey (USGS) has developed a new quantitative methodology for assessing resources in continuous (unconventional) petroleum deposits. Continuous petroleum resources include shale gas, coalbed gas, and other oil and gas deposits in low-permeability ("tight") reservoirs. The methodology is based on an approach combining geologic understanding with well productivities. The methodology is probabilistic, with both input and output variables as probability distributions, and uses Monte Carlo simulation to calculate the estimates. The new methodology is an improvement of previous USGS methodologies in that it better accommodates the uncertainties in undrilled or minimally drilled deposits that must be assessed using analogs. The publication is a collection of PowerPoint slides with accompanying comments.

Applicability and feasibility of systematic review for performing evidence-based risk assessment in food and feed safety

DEFF Research Database (Denmark)

Aiassa, E.; Higgins, J.P.T.; Frampton, G. K.

2015-01-01

for answering questions in health care, and can be implemented to minimise biases in food and feed safety risk assessment. However, no methodological frameworks exist for refining risk assessment multi-parameter models into questions suitable for systematic review, and use of meta-analysis to estimate all......Food and feed safety risk assessment uses multi-parameter models to evaluate the likelihood of adverse events associated with exposure to hazards in human health, plant health, animal health, animal welfare and the environment. Systematic review and meta-analysis are established methods...... parameters in the risk model. This approach to planning and prioritising systematic review seems to have useful implications for producing evidence-based food and feed safety risk assessment....

Development of a methodology for safety classification on a non-reactor nuclear facility illustrated using an specific example

International Nuclear Information System (INIS)

Scheuermann, F.; Lehradt, O.; Traichel, A.

2015-01-01

To realize the safety of personnel and environment systems and components of nuclear facilities are classified according to their potential danger into safety classes. Based on this classification different demands on the manufacturing quality result. The objective of this work is to present the standardized method developed by NUKEM Technologies Engineering Services for the categorization into the safety classes restricted to Non-reactor nuclear facilities (NRNF). Exemplary the methodology is used on the complex Russian normative system (four safety classes). For NRNF only the lower two safety classes are relevant. The classification into the lowest safety class 4 is accordingly if the maximum resulting dose following from clean-up actions in case of incidents/accidents remains below 20 mSv and the volume activity restrictions of set in NRB-99/2009 are met. The methodology is illustrated using an example. In short the methodology consists of: - Determination of the working time to remove consequences of incidents, - Calculation of the dose resulting from direct radiation and due to inhalation during these works. The application of this methodology avoids over-conservative approaches. As a result some previously higher classified equipment can be classified into the lower safety class.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Assessment of safety culture from the INB organization: A case study for nuclear fuel cycle industry

International Nuclear Information System (INIS)

Goncalves, J.S.; Barreto, A.C.

2002-01-01

The present article describes strategies, methodologies and first results on the Safety Culture Self-assessment Project under way at INB since August 2001. As a Brazilian Government company in charge of the nuclear fuel cycle activities,. the main purposes of the Project is to evaluate the present status of its safety culture and to propose actions to ensure continuous safety improvement at management level of its industrial processes. The proposed safety culture assessment describes INB's various production sites taking into account the different aspects of their activities, such as regional, social and technical issues. The survey was performed in March/2002 very good attendance (about 80%) the employees. The first global survey results are presented in item 4. (author)

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

Science.gov (United States)

Ancel, Ersin; Shih, Ann T.

2014-01-01

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

Sensitivity and Uncertainty Analyses Applied to Neutronics Calculations for Safety Assessment at IRSN

International Nuclear Information System (INIS)

Ivanov, Evgeny; Ivanova, Tatiana; Pignet, Sophie

2013-01-01

Objective of the presentation: • Present IRSN vision relevant to validation of stand-alone neutronics codes on support of the fuel cycle and reactor safety assessment for fast neutron reactors. • Provide work status, future developments and needs for R&D working program on validation methodology for neutronics of fast systems

Methodology, status and plans for development and assessment of Cathare code

Energy Technology Data Exchange (ETDEWEB)

Bestion, D.; Barre, F.; Faydide, B. [CEA - Grenoble (France)

1997-07-01

This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests or integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.

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)

Contribution to the methodology of safety evaluation - and licensing of reloading cycle for PWR type reactors

International Nuclear Information System (INIS)

Esteves, R.G.

1981-01-01

A simplified methodology for evaluating a reload safety cycle is presented. This methodology consists in selecting for each foreseen accident, the nuclear key reload safety parameters which determine the accident evolution. So, each key reload parameter is calculated and compared with its value for the first cycle. Those accidents, which have their key reload parameter bounded by the values of the first cycle do not need reanalise. Extension of the validity of this methodology when there exists change of fuel supplier is commented. (Author) [pt

Hanford Site Risk Assessment Methodology. Revision 3

International Nuclear Information System (INIS)

1995-05-01

This methodology has been developed to prepare human health and ecological evaluations of risk as part of the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) remedial investigations (RI) and the Resource conservation and Recovery Act of 1976 (RCRA) facility investigations (FI) performed at the Hanford Site pursuant to the hanford Federal Facility Agreement and Consent Order (Ecology et al. 1994), referred to as the Tri-Party Agreement. Development of the methodology has been undertaken so that Hanford Site risk assessments are consistent with current regulations and guidance, while providing direction on flexible, ambiguous, or undefined aspects of the guidance. The methodology identifies site-specific risk assessment considerations and integrates them with approaches for evaluating human and ecological risk that can be factored into the risk assessment program supporting the Hanford Site cleanup mission. Consequently, the methodology will enhance the preparation and review of individual risk assessments at the Hanford Site

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

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

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

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)

Methodology for safety and security of radioactive sources and materials. The Israeli approach

International Nuclear Information System (INIS)

Keren, M.

1998-01-01

About 10 Radioactive incidents occurred in Israel during 1996-1997. Some of them were theft or lost of Radioactive equipment or sources, some happened because misuse of Radioactive equipment and some of other reasons. Part of them could be eliminated if a better methodological attitude to the subject existed. A new methodology for notification, registration and licensing is described. Hopefully this methodology will increase defense in depth and the Safety and Security of Radioactive sources and materials. Information on the inventory of Radioactive sources and materials is essential. Where they are situated, what is the supply rate or all history from berth to grave. Persons involved are important: Who are the Radiation Safety Officers (RSO), what is their training and updating programs. As much as possible information on the site and places where those Radioactive sources and materials are used. Procedures for security of sources and materials is part of site information, beside safety precautions. Users are obliged to inform on any changes and to ask for confirmation to those changes. The same is when high activity sources are moved across the country. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-12-15

that lie outside the scope of the quantitative safety assessment. These arguments include: Support from natural and anthropogenic analogues for both key process understanding and total system performance. Comparison of the methodology and results with the earlier TILA-99 and SR-Can safety assessments, as well as other international safety assessments, to ensure completeness, consistency and reasonableness of the present assessment. Use of safety indicators other than dose and activity to avoid uncertainties in future human lifestyles and also in geological processes on very long timescales. Consideration of the calculation results from a wider perspective to consider significance of their impact compared to other risks

Health economic assessment: a methodological primer.

Science.gov (United States)

Simoens, Steven

2009-12-01

This review article aims to provide an introduction to the methodology of health economic assessment of a health technology. Attention is paid to defining the fundamental concepts and terms that are relevant to health economic assessments. The article describes the methodology underlying a cost study (identification, measurement and valuation of resource use, calculation of costs), an economic evaluation (type of economic evaluation, the cost-effectiveness plane, trial- and model-based economic evaluation, discounting, sensitivity analysis, incremental analysis), and a budget impact analysis. Key references are provided for those readers who wish a more advanced understanding of health economic assessments.

Health Economic Assessment: A Methodological Primer

Directory of Open Access Journals (Sweden)

Steven Simoens

2009-11-01

Full Text Available This review article aims to provide an introduction to the methodology of health economic assessment of a health technology. Attention is paid to defining the fundamental concepts and terms that are relevant to health economic assessments. The article describes the methodology underlying a cost study (identification, measurement and valuation of resource use, calculation of costs, an economic evaluation (type of economic evaluation, the cost-effectiveness plane, trial- and model-based economic evaluation, discounting, sensitivity analysis, incremental analysis, and a budget impact analysis. Key references are provided for those readers who wish a more advanced understanding of health economic assessments.

Draft report: a selection methodology for LWR safety R and D programs and proposals

International Nuclear Information System (INIS)

Husseiny, A.A.; Ritzman, R.L.

1980-03-01

The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application

Draft report: a selection methodology for LWR safety R and D programs and proposals

Energy Technology Data Exchange (ETDEWEB)

Husseiny, A. A.; Ritzman, R. L.

1980-03-01

The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application.

The Use and Development of Probabilistic Safety Assessment in NEA Member Countries

International Nuclear Information System (INIS)

2002-01-01

The mission of the CSNI is to assist Member countries in maintaining and further developing the scientific and technical knowledge base required to assess the safety of nuclear reactors and fuel cycle facilities. The mission of the Working Group on Risk Assessment (WGRisk) is to advance the understanding and utilisation of Probabilistic Safety Assessment (PSA) in ensuring continued safety of nuclear installations in Member countries. In pursuing this goal, the Working Group shall recognize the different methodologies for identifying contributors to risk and assessing their importance. While the Working Group shall continue to focus on the more mature PSA methodologies for Level 1, Level 2, internal, external, shutdown, etc. It shall also consider the applicability and maturity of PSA methods for considering evolving issues such as human reliability, software reliability, ageing issues, etc., as appropriate. This report provides descriptions of the current status of PSA programmes in Member countries including basic background information, guidelines, various PSA applications, major results in recent studies, PSA based plant modifications and research and development topics. While the compilation is a not complete compilation it provides a 'snapshot' of the current situation in the Member countries and hence it provides reference information and various insights to both the PSA practician and others involved in the nuclear industry. The terms PSA (Probabilistic Safety Assessment) and PRA (Probabilistic Risk Assessment) are utilised to denote this subject. In each of the chapters the objective is to present a 'snapshot' of the current status. The main issues considered in the different chapters are Background Information, Quantitative Safety Guidelines, Status of PSA Programmes, PSA Applications, PSA Related Research and Development and PSA Based Plant Modifications. It is important to note that the information contained in this report represents current practices in

Prototype application of best estimate and uncertainty safety analysis methodology to large LOCA analysis

International Nuclear Information System (INIS)

Luxat, J.C.; Huget, R.G.

2001-01-01

Development of a methodology to perform best estimate and uncertainty nuclear safety analysis has been underway at Ontario Power Generation for the past two and one half years. A key driver for the methodology development, and one of the major challenges faced, is the need to re-establish demonstrated safety margins that have progressively been undermined through excessive and compounding conservatism in deterministic analyses. The major focus of the prototyping applications was to quantify the safety margins that exist at the probable range of high power operating conditions, rather than the highly improbable operating states associated with Limit of the Envelope (LOE) assumptions. In LOE, all parameters of significance to the consequences of a postulated accident are assumed to simultaneously deviate to their limiting values. Another equally important objective of the prototyping was to demonstrate the feasibility of conducting safety analysis as an incremental analysis activity, as opposed to a major re-analysis activity. The prototype analysis solely employed prior analyses of Bruce B large break LOCA events - no new computer simulations were undertaken. This is a significant and novel feature of the prototyping work. This methodology framework has been applied to a postulated large break LOCA in a Bruce generating unit on a prototype basis. This paper presents results of the application. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

European project SARGEN IV: safety approach and assessment of GEN IV reactors

International Nuclear Information System (INIS)

Ammirabile, L.

2013-01-01

• SARGEN I V has elaborated a proposal for the harmonization of safety assessment practices for GEN IV NPP. • An overall reinforcement of DiD is expected for GEN I V NPP, including improved independence between all levels of DiD. • An inherent approach should reinforce the fulfillment of fundamental safety functions e.g. the consequences for some situations should be reduced and the grace periods should be extended. For the same reason, the use of passive systems can be envisaged. • The need of complementary and integrated deterministic and probabilistic approaches is reiterated. • Methodologies: Some of them are not yet applied. • Assessment of hazards would be a challenging aspect of next generation of NPP safety assessment and should be improved, which is confirmed by the first insights of Fukushima Daiichi TEPCO reactors accidents. • Provisions to cope with extreme events notably to improve the grace period before cliff-edge effects and thus allowing back-up measures to be implemented have to be defined and should be considered as hardened equipments

Characterising influences on safety culture in military aviation:a methodologically grounded approach

OpenAIRE

Bennett, Anthea; Hellier, Elizabeth; Weyman, Andrew

2015-01-01

Historically, much effort has been expended in safety culture / climate research toward identifying a generic core set of components, predominately using the self-administered questionnaire approach. However, no stable unified model has emerged, and much of this research has taken a methodologically top-down approach to depicting organisational safety culture. In light of this, the benefits of qualitative exploration as a precursor to and foundation for the development of quantitative climate...

A generic standard for assessing and managing activities with significant risk to health and safety

International Nuclear Information System (INIS)

Wilde, T.S.; Sandquist, G.M.

2005-01-01

Some operations and activities in industry, business, and government can present an unacceptable risk to health and safety if not performed according to established safety practices and documented procedures. The nuclear industry has extensive experience and commitment to assessing and controlling such risks. This paper provides a generic standard based upon DOE Standard DOE-STD-3007- 93, Nov 1993, Change Notice No. 1, Sep 1998. This generic standard can be used to assess practices and procedures employed by any industrial and government entity to ensure that an acceptable level of safety and control prevail for such operations. When any activity and operation is determined to involve significant risk to health and safety to workers or the public, the organization should adopt and establish an appropriate standard and methodology to ensure that adequate health and safety prevail. This paper uses DOE experience and standards to address activities with recognized potential for impact upon health and safety. Existing and future assessments of health and safety issues can be compared and evaluated against this generic standard for insuring that proper planning, analysis, review, and approval have been made. (authors)

Risk assessment of cryogenic installations – implementation, applicability of methodologies and challenges at CERN

CERN Multimedia

CERN. Geneva

2016-01-01

For the safe design of a cryogenic installation, it is essential to carry out a comprehensive hazard identification and risk estimate in order to put in place the necessary control measures for an adequate risk mitigation. According to CERN Safety Rules, it is mandatory that the organic unit owning a cryogenic facility conducts and documents a risk assessment. This requirement is also given by the European Directive 2014/68/EU to manufacturers of pressure equipment. During the talk, some of the challenges CERN faces in the development of risk assessments across the broad array of activities involving cryogenic equipment in the organization will be discussed. Challenges such as the choice of the best-suited risk assessment methodology based on the features and complexity of the installation/activities, the efforts to develop tools to facilitate hazard identification, risk analysis and definition of related measures to protect the health and safety of workers, such as streamlined guidelines, forms and check...

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

Methodology for flood risk analysis for nuclear power plants

International Nuclear Information System (INIS)

Wagner, D.P.; Casada, M.L.; Fussell, J.B.

1984-01-01

The methodology for flood risk analysis described here addresses the effects of a flood on nuclear power plant safety systems. Combining the results of this method with the probability of a flood allows the effects of flooding to be included in a probabilistic risk assessment. The five-step methodology includes accident sequence screening to focus the detailed analysis efforts on the accident sequences that are significantly affected by a flood event. The quantitative results include the flood's contribution to system failure probability, accident sequence occurrence frequency and consequence category occurrence frequency. The analysis can be added to existing risk assessments without a significant loss in efficiency. The results of two example applications show the usefulness of the methodology. Both examples rely on the Reactor Safety Study for the required risk assessment inputs and present changes in the Reactor Safety Study results as a function of flood probability

Assessment of the long-term safety of repositories. Scientific basis

International Nuclear Information System (INIS)

Noseck, Ulrich; Becker, Dirk; Fahrenholz, Christine

2008-12-01

The project contributed to increase the scientific knowledge on the long-term safety assessment and the safety cases of a radioactive waste repository. International guidelines and more recent safety cases from other countries were evaluated. The feasibility study of the three safety indicators ''individual dose rate'', ''radiotoxicity concentration in the biosphere water'' and ''radiotoxicity flux from the geosphere'' showed that due to the independently derived corresponding reference values these indicators describe three different safety statements. The combination of the three values can give a stronger argument for the safety of the repository system. Another important methodological aspect of the safety cases is the definition and selection of scenarios, one of these the human intrusion scenario. Various human intrusion scenarios are considered in the different nations, which differ significantly with respect to type and time scale, the exposition type and exposition pathway. Further progress has been achieved in how to treat human intrusion scenarios in a German post-closure safety case. Another port of the project dealt with the impact of specific geochemical processes on the long-term safety of the repository. The impact of climate changes on the long-term safety of a radioactive waste repository in rock salt was investigated with respect to processes in the overburden and the biosphere where highest impact is expected. Sofa simplified models and only discrete climate estates have been considered

Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

Energy Technology Data Exchange (ETDEWEB)

Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

2015-11-15

Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

International Nuclear Information System (INIS)

Song, Wei; Wu, Yuanyu; Hu, Wenjun; Zuo, Jiaxu

2015-01-01

Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

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

Directory of Open Access Journals (Sweden)

Ernest S. Fung

2018-06-01

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

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)

The current CEA/DRN safety approach for the design and the assessment of future nuclear installations

International Nuclear Information System (INIS)

Fiorini, G.L.; Pinto, P.L.; Costa, M.

1999-01-01

The purpose of the document is to present the basis of the safety approach currently implemented by the CEA/DRN, both for the design and the assessment of innovative systems and future nuclear installations. This approach is the result of the experience maturated, within the context of the CEA/DRN Innovative Programme through practical applications over several future concepts, both for fission and fusion reactors, as well as for waste disposal. The background of this experience is structured coherently with the European Safety Authorities recommendations and the European Utilities Requirements (EUR). The Defence In Depth principle and its application, by means, among others, of the barrier concept, remains the basis of the safety design process of future nuclear installations. Its adequacy is checked through the safety assessment. The methodology for Lines Of Defence (LOD) implementation as well as the one for the LOD architecture assessment is shown and motivated. The document shows that the clear and unambiguous definition of the safety approach provides an essential base for the organisation of the design tasks, being sure that the safety aspects are correctly taken into account and implemented, and for an adequate safety assessment of the final design, both from qualitative point of view as well as for the quantitative safety analysis. (author)

Hanford Site baseline risk assessment methodology. Revision 2

Energy Technology Data Exchange (ETDEWEB)

1993-03-01

This methodology has been developed to prepare human health and environmental evaluations of risk as part of the Comprehensive Environmental Response, Compensation, and Liability Act remedial investigations (RIs) and the Resource Conservation and Recovery Act facility investigations (FIs) performed at the Hanford Site pursuant to the Hanford Federal Facility Agreement and Consent Order referred to as the Tri-Party Agreement. Development of the methodology has been undertaken so that Hanford Site risk assessments are consistent with current regulations and guidance, while providing direction on flexible, ambiguous, or undefined aspects of the guidance. The methodology identifies Site-specific risk assessment considerations and integrates them with approaches for evaluating human and environmental risk that can be factored into the risk assessment program supporting the Hanford Site cleanup mission. Consequently, the methodology will enhance the preparation and review of individual risk assessments at the Hanford Site.

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)

Probabilistic safety assessment - regulatory perspective

International Nuclear Information System (INIS)

Solanki, R.B.; Paul, U.K.; Hajra, P.; Agarwal, S.K.

2002-01-01

Full text: Nuclear power plants (NPPs) have been designed, constructed and operated mainly based on deterministic safety analysis philosophy. In this approach, a substantial amount of safety margin is incorporated in the design and operational requirements. Additional margin is incorporated by applying the highest quality engineering codes, standards and practices, and the concept of defence-in-depth in design and operating procedures, by including conservative assumptions and acceptance criteria in plant response analysis of postulated initiating events (PIEs). However, as the probabilistic approach has been improved and refined over the years, it is possible for the designer, operator and regulator to get a more detailed and realistic picture of the safety importance of plant design features, operating procedures and operational practices by using probabilistic safety assessment (PSA) along with the deterministic methodology. At present, many countries including USA, UK and France are using PSA insights in their decision making along with deterministic basis. India has also made substantial progress in the development of methods for carrying out PSA. However, consensus on the use of PSA in regulatory decision-making has not been achieved yet. This paper emphasises on the requirements (e.g.,level of details, key modelling assumptions, data, modelling aspects, success criteria, sensitivity and uncertainty analysis) for improving the quality and consistency in performance and use of PSA that can facilitate meaningful use of the PSA insights in the regulatory decision-making in India. This paper also provides relevant information on international scenario and various application areas of PSA along with progress made in India. The PSA perspective presented in this paper may help in achieving consensus on the use of PSA for regulatory / utility decision-making in design and operation of NPPs

Thinking of the safety assessment of HLW disposal

International Nuclear Information System (INIS)

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

2014-01-01

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

ALARP considerations in criticality safety assessments

International Nuclear Information System (INIS)

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

2003-01-01

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

Safety culture assessment developed by JANTI

International Nuclear Information System (INIS)

Hamada, Jun

2009-01-01

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

Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

Energy Technology Data Exchange (ETDEWEB)

Kumar, Mukesh, E-mail: mukeshd@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravarty, Aranyak [School of Nuclear Studies and Application, Jadavpur University, Kolkata 700032 (India); Nayak, A.K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Prasad, Hari; Gopika, V. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

2014-10-15

Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

International Nuclear Information System (INIS)

Kumar, Mukesh; Chakravarty, Aranyak; Nayak, A.K.; Prasad, Hari; Gopika, V.

2014-01-01

Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components

Research on Methodology to Prioritize Critical Digital Assets based on Nuclear Risk Assessment

Energy Technology Data Exchange (ETDEWEB)

Kim, Wonjik; Kwon, Kookheui; Kim, Hyundoo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2016-10-15

Digital systems are used in nuclear facilities to monitor and control various types of field devices, as well as to obtain and store vital information. Therefore, it is getting important for nuclear facilities to protect digital systems from cyber-attack in terms of safety operation and public health since cyber compromise of these systems could lead to unacceptable radiological consequences. Based on KINAC/RS-015 which is a cyber security regulatory standard, regulatory activities for cyber security at nuclear facilities generally focus on critical digital assets (CDAs) which are safety, security, and emergency preparedness related digital assets. Critical digital assets are estimated over 60% among all digital assets in a nuclear power plant. Therefore, it was required to prioritize critical digital assets to improve efficiency of regulation and implementation. In this paper, the research status on methodology development to prioritize critical digital assets based on nuclear risk assessment will be introduced. In this paper, to derive digital asset directly affect accident, PRA results (ET, FT, and minimal cut set) are analyzed. According to result of analysis, digital systems related to CD are derived ESF-CCS (safety-related component control system) and Process-CCS (non-safety-related component control system) as well as Engineered Safety Features Actuation System (ESFAS). These digital assets can be identified Vital Digital Asset (VDA). Hereafter, to develop general methodology which was identified VDA related to accident among CDAs, (1) method using result of minimal cut set in PRA model will be studied and (2) method quantifying result of Digital I and C PRA which is performed to reflect all digital cabinet related to system in FT will be studied.

Research on Methodology to Prioritize Critical Digital Assets based on Nuclear Risk Assessment

International Nuclear Information System (INIS)

Kim, Wonjik; Kwon, Kookheui; Kim, Hyundoo

2016-01-01

Digital systems are used in nuclear facilities to monitor and control various types of field devices, as well as to obtain and store vital information. Therefore, it is getting important for nuclear facilities to protect digital systems from cyber-attack in terms of safety operation and public health since cyber compromise of these systems could lead to unacceptable radiological consequences. Based on KINAC/RS-015 which is a cyber security regulatory standard, regulatory activities for cyber security at nuclear facilities generally focus on critical digital assets (CDAs) which are safety, security, and emergency preparedness related digital assets. Critical digital assets are estimated over 60% among all digital assets in a nuclear power plant. Therefore, it was required to prioritize critical digital assets to improve efficiency of regulation and implementation. In this paper, the research status on methodology development to prioritize critical digital assets based on nuclear risk assessment will be introduced. In this paper, to derive digital asset directly affect accident, PRA results (ET, FT, and minimal cut set) are analyzed. According to result of analysis, digital systems related to CD are derived ESF-CCS (safety-related component control system) and Process-CCS (non-safety-related component control system) as well as Engineered Safety Features Actuation System (ESFAS). These digital assets can be identified Vital Digital Asset (VDA). Hereafter, to develop general methodology which was identified VDA related to accident among CDAs, (1) method using result of minimal cut set in PRA model will be studied and (2) method quantifying result of Digital I and C PRA which is performed to reflect all digital cabinet related to system in FT will be studied

Using proliferation assessment methodologies for Safeguards-by-Design

International Nuclear Information System (INIS)

Van der Meer, K.; Rossa, R.; Turcanu, C.; Borella, A.

2013-01-01

MYRRHA, an accelerator driven system (ADS) is designed as a proton accelerator coupled to a liquid Pb-Bi spallation target, surrounded by a Pb-Bi cooled sub-critical neutron multiplying medium in a pool type configuration. An assessment based on three methodologies was made of the proliferation risks of the MYRRHA ADS in comparison with the BR2 MTR, an existing research reactor at the Belgian Nuclear Research Centre SCK-CEN. The used methodologies were the TOPS (Technical Opportunities to Increase the Proliferation Resistance of Nuclear Power Systems), the PR-PP and the INPRO methodologies. The various features of the methodologies are described and the results of the assessments are given and discussed. It is concluded that it would be useful to define one single methodology with two options to perform a quick and a more detailed assessment. The paper is followed by the slides of the presentation

Feasibility studies of safety assessment methods for programmable automation systems. Final report of the AVV project

International Nuclear Information System (INIS)

Haapanen, P.; Maskuniitty, M.; Pulkkinen, U.; Heikkinen, J.; Korhonen, J.; Tuulari, E.

1995-10-01

Feasibility studies of two different groups of methodologies for safety assessment of programmable automation systems has been executed at the Technical Research Centre of Finland (VTT). The studies concerned the dynamic testing methods and the fault tree (FT) and failure mode and effects analysis (FMEA) methods. In order to get real experience in the application of these methods, an experimental testing of two realistic pilot systems were executed and a FT/FMEA analysis of a programmable safety function accomplished. The purpose of the studies was not to assess the object systems, but to get experience in the application of methods and assess their potentials and development needs. (46 refs., 21 figs.)

Safety and performance assessment of geologic disposal systems for nuclear wastes

International Nuclear Information System (INIS)

Peltonen, E.

1987-01-01

This thesis presents a methodology for the safety and performance assesment of final disposal of nuclear wastes into crystalline bedrock. The applicability of radiation protection objectives is discussed, as well as the goals of the assessment in the various repository system development phases. Due consideration is given to the description of the pertinent analysis methods and to the comprehensive model system. The methodology has been applied to assess the acceptability of the basic disposal concepts and to study the possibilities for the optimization of protection. Furthermore, performance of different components in the multiple barrier disposal systems is estimated. The waste types dealt with are low- and intermediate-level waste as well as high-level spent nuclear fuel from a nuclear power plant. In addition, an option of high-level vitrified waste from reprocessing of spent fuel is taken into account. On the basis of the various analyses carried out it can be concluded that the disposal of different nuclear wastes in the Finnish bedrock in properly designed repositories meets the radiation protection objectives with good confidence. In addition, the studies indicate that the safety margins are considerable. This is due to the fact that the overall performance of the multiple barrier disposal systems analysed is not sensitive to possible unfavourable changes in barrier properties. From the optimization of protection point of view it can be concluded that there is no need to develop more effective repository designs than those analysed in this thesis. In fact, the results indicate that the most sophisticated designs have already gone beyond an optimal level of safety

The waste isolation safety assessment programme