WorldWideScience

Sample records for safety analysis methodology

  1. 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)

  2. Comprehensive Safety Analysis 2010 Safety Measurement System (SMS) Methodology, Version 2.1 Revised December 2010

    Science.gov (United States)

    2010-12-01

    This report documents the Safety Measurement System (SMS) methodology developed to support the Comprehensive Safety Analysis 2010 (CSA 2010) Initiative for the Federal Motor Carrier Safety Administration (FMCSA). The SMS is one of the major tools for...

  3. Incorporation of advanced accident analysis methodology into safety analysis reports

    International Nuclear Information System (INIS)

    2003-05-01

    as structural analysis codes and computational fluid dynamics codes (CFD) are applied. The initial code development took place in the sixties and seventies and resulted in a set of quite conservative codes for the reactor dynamics, thermal-hydraulics and containment analysis. The most important limitations of these codes came from insufficient knowledge of the physical phenomena and of the limited computer memory and speed. Very significant advances have been made in the development of the code systems during the last twenty years in all of the above areas. If the data for the physical models of the code are sufficiently well established and allow quite a realistic analysis, these newer versions are called advanced codes. The assumptions used in the deterministic safety analysis vary from very pessimistic to realistic assumptions. In the accident analysis terminology, it is customary to call the pessimistic assumptions 'conservative' and the realistic assumptions 'best estimate'. The assumptions can refer to the selection of physical models, the introduction of these models into the code, and the initial and boundary conditions including the performance and failures of the equipment and human action. The advanced methodology in the present report means application of advanced codes (or best estimate codes), which sometimes represent a combination of various advanced codes for separate stages of the analysis, and in some cases in combination with experiments. The Safety Analysis Reports are required to be available before and during the operation of the plant in most countries. The contents, scope and stages of the SAR vary among the countries. The guide applied in the USA, i.e. the Regulatory Guide 1.70 is representative for the way in which the SARs are made in many countries. During the design phase, a preliminary safety analysis report (PSAR) is requested in many countries and the final safety analysis report (FSAR) is required for the operating licence. There is

  4. 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

  5. Using of BEPU methodology in a final safety analysis report

    International Nuclear Information System (INIS)

    Menzel, Francine; Sabundjian, Gaiane; D'auria, Francesco; Madeira, Alzira A.

    2015-01-01

    The Nuclear Reactor Safety (NRS) has been established since the discovery of nuclear fission, and the occurrence of accidents in Nuclear Power Plants worldwide has contributed for its improvement. The Final Safety Analysis Report (FSAR) must contain complete information concerning safety of the plant and plant site, and must be seen as a compendium of NRS. The FSAR integrates both the licensing requirements and the analytical techniques. The analytical techniques can be applied by using a realistic approach, addressing the uncertainties of the results. This work aims to show an overview of the main analytical techniques that can be applied with a Best Estimated Plus Uncertainty (BEPU) methodology, which is 'the best one can do', as well as the ALARA (As Low As Reasonably Achievable) principle. Moreover, the paper intends to demonstrate the background of the licensing process through the main licensing requirements. (author)

  6. Using of BEPU methodology in a final safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Menzel, Francine; Sabundjian, Gaiane, E-mail: fmenzel@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); D' auria, Francesco, E-mail: f.dauria@ing.unipi.it [Universita degli Studi di Pisa, Gruppo di Ricerca Nucleare San Piero a Grado (GRNSPG), Pisa (Italy); Madeira, Alzira A., E-mail: alzira@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The Nuclear Reactor Safety (NRS) has been established since the discovery of nuclear fission, and the occurrence of accidents in Nuclear Power Plants worldwide has contributed for its improvement. The Final Safety Analysis Report (FSAR) must contain complete information concerning safety of the plant and plant site, and must be seen as a compendium of NRS. The FSAR integrates both the licensing requirements and the analytical techniques. The analytical techniques can be applied by using a realistic approach, addressing the uncertainties of the results. This work aims to show an overview of the main analytical techniques that can be applied with a Best Estimated Plus Uncertainty (BEPU) methodology, which is 'the best one can do', as well as the ALARA (As Low As Reasonably Achievable) principle. Moreover, the paper intends to demonstrate the background of the licensing process through the main licensing requirements. (author)

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

    International Nuclear Information System (INIS)

    2003-01-01

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

  8. SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

    Directory of Open Access Journals (Sweden)

    WOLFGANG HARTMANN

    2013-10-01

    Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

  9. Development of Safety Margin Analysis Methodology on Aging Effect for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Yoo, Jun Soo; Choi, Yong Won; Park, Chang Hwan [Seoul National Univ., Seoul (Korea, Republic of)

    2007-07-01

    Considering that operating year of Wolsong Unit 1 gets close to the design life, 30 years, the aging effect due to the component degradation takes into consideration as an important safety issue. However, since the thermalhydraulic effect due to the aging did not identify clearly, the safety analysis methodology is not be well established so far. Therefore, in this study, the aging effect affected by thermal-hydraulic characteristics was investigated and a safety margin analysis methodology considering aging effect was proposed.

  10. Development of Safety Margin Analysis Methodology on Aging Effect for CANDU Reactors (II)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Choi, Yong Won; Lee, Un Chul [Nuclear Engr. Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2007-10-15

    Considering that operating year of Wolsong Unit 1 gets close to the design life, 30 years, the aging effect due to the component degradation takes into consideration as an important safety issue. However, since the thermal hydraulic effect due to the aging did not identify clearly, the safety analysis methodology is not be well established so far. Therefore, in this study, the aging effect affected by thermal-hydraulic characteristics was investigated and a safety margin analysis methodology considering aging effect was proposed.

  11. 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)

  12. 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)

  13. 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

  14. 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

  15. 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

  16. Development of Non-LOCA Safety Analysis Methodology with RETRAN-3D and VIPRE-01/K

    International Nuclear Information System (INIS)

    Kim, Yo-Han; Cheong, Ae-Ju; Yang, Chang-Keun

    2004-01-01

    Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report

  17. 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.

  18. CANDU safety analysis system establishment; development of trip coverage and multi-dimensional hydrogen analysis methodology

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jong Ho; Ohn, M. Y.; Cho, C. H. [KOPEC, Taejon (Korea)

    2002-03-01

    The trip coverage analysis model requires the geometry network for primary and secondary circuit as well as the plant control system to simulate all the possible plant operating conditions throughout the plant life. The model was validated for the power maneuvering and the Wolsong 4 commissioning test. The trip coverage map was produced for the large break loss of coolant accident and the complete loss of class IV power event. The reliable multi-dimensional hydrogen analysis requires the high capability for thermal hydraulic modelling. To acquire such a basic capability and verify the applicability of GOTHIC code, the assessment of heat transfer model, hydrogen mixing and combustion model was performed. Also, the assessment methodology for flame acceleration and deflagration-to-detonation transition is established. 22 refs., 120 figs., 31 tabs. (Author)

  19. 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

  20. 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.

  1. Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Hyun, E-mail: jhkim@actbest.com [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); ACT Co., Ltd, 705 Gwanpyeong-dong, Yuseong-gu, Daejeon 305-509 (Korea, Republic of); Jin, Dong Sik [ACT Co., Ltd, 705 Gwanpyeong-dong, Yuseong-gu, Daejeon 305-509 (Korea, Republic of); Chang, Soon Heung [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2013-10-15

    Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively.

  2. AGR core safety assessment methodologies

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  3. Geo-ethical dimension of community's safety: rural and urban population vulnerability analysis methodology

    Science.gov (United States)

    Kostyuchenko, Yuriy; Movchan, Dmytro; Kopachevsky, Ivan; Yuschenko, Maxim

    2016-04-01

    Modern world based on relations more than on causalities, so communicative, socio-economic, and socio-cultural issues are important to understand nature of risks and to make correct, ethical decisions. Today major part of risk analysts declared new nature of modern risks. We faced coherent or systemic risks, realization of which leads to domino effect, unexpected growing of losses and fatalities. This type of risks originated by complicated nature of heterogeneous environment, close interconnection of engineering networks, and changing structure of society. Heterogeneous multi-agent environment generates systemic risks, which requires analyze multi-source data with sophisticated tools. Formal basis for analysis of this type of risks is developed during last 5-7 years. But issues of social fairness, ethics, and education require further development. One aspect of analysis of social issues of risk management is studied in this paper. Formal algorithm for quantitative analysis of multi-source data analysis is proposed. As it was demonstrated, using proposed methodological base and the algorithm, it is possible to obtain regularized spatial-temporal distribution of investigated parameters over whole observation period with rectified reliability and controlled uncertainty. The result of disaster data analysis demonstrates that about half of direct disaster damage might be caused by social factors: education, experience and social behaviour. Using data presented also possible to estimate quantitative parameters of the losses distributions: a relation between education, age, experience, and losses; as well as vulnerability (in terms of probable damage) toward financial status in current social density. It is demonstrated that on wide-scale range an education determines risk perception and so vulnerability of societies. But on the local level there are important heterogeneities. Land-use and urbanization structure influencing to vulnerability essentially. The way to

  4. CSNI Status summary on utilization of best-estimate methodology in safety analysis and licensing

    International Nuclear Information System (INIS)

    1996-10-01

    The PWG 2 Task Group on Thermal Hydraulic System Behavior has discussed the subject of the use of best-estimate codes in the licensing process (codes that model thermal hydraulic processes are important to assessing safety system performance). The Task Group set out to determine the prevailing practices in member countries, concerning safety assessment and safety review of transients affecting the reactor coolant system. A summary of information provided by member countries in response to eleven questions is given: Who is Responsible for Safety Analysis? Who is Responsible for Review and Evaluation of Safety Analysis? Do the Regulations Permit the use of Best-Estimate Codes? What are the Requirements for What Constitutes a Best Estimate Code? What are the Requirements Concerning Code Documentation? What are the Requirements for Review of Code Models and Correlations? What are the Requirements Concerning Code Assessment? What are the Requirements Concerning Initial and Boundary Conditions? What are the Requirements Concerning Operability of Active Equipment? What are the Requirements Concerning Operator Actions?

  5. Incident analysis methodology

    International Nuclear Information System (INIS)

    Libmann, J.

    1986-05-01

    The number of French nuclear power stations in operation and their division into standardized plant series very soon led to the requirement for a precise organization, within both the nuclear safety authorities and the operator, Electricite de France. The methods of analysis have been gradually extended and diversified and we shall speak of them, but it is evident that a very precise definition of the boundaries between what concerns safety and what does not concern it, is needed. This report first deals with the criteria on which declarations are based before outlining the main guidelines of analysis methodology [fr

  6. Indirect Lightning Safety Assessment Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Ong, M M; Perkins, M P; Brown, C G; Crull, E W; Streit, R D

    2009-04-24

    Lightning is a safety hazard for high-explosives (HE) and their detonators. In the However, the current flowing from the strike point through the rebar of the building The methodology for estimating the risk from indirect lighting effects will be presented. It has two parts: a method to determine the likelihood of a detonation given a lightning strike, and an approach for estimating the likelihood of a strike. The results of these two parts produce an overall probability of a detonation. The probability calculations are complex for five reasons: (1) lightning strikes are stochastic and relatively rare, (2) the quality of the Faraday cage varies from one facility to the next, (3) RF coupling is inherently a complex subject, (4) performance data for abnormally stressed detonators is scarce, and (5) the arc plasma physics is not well understood. Therefore, a rigorous mathematical analysis would be too complex. Instead, our methodology takes a more practical approach combining rigorous mathematical calculations where possible with empirical data when necessary. Where there is uncertainty, we compensate with conservative approximations. The goal is to determine a conservative estimate of the odds of a detonation. In Section 2, the methodology will be explained. This report will discuss topics at a high-level. The reasons for selecting an approach will be justified. For those interested in technical details, references will be provided. In Section 3, a simple hypothetical example will be given to reinforce the concepts. While the methodology will touch on all the items shown in Figure 1, the focus of this report is the indirect effect, i.e., determining the odds of a detonation from given EM fields. Professor Martin Uman from the University of Florida has been characterizing and defining extreme lightning strikes. Using Professor Uman's research, Dr. Kimball Merewether at Sandia National Laboratory in Albuquerque calculated the EM fields inside a Faraday-cage type

  7. Methodology and development of instruments for the safety analysis of a nuclear reprocessing plant

    International Nuclear Information System (INIS)

    Markett, J.

    1987-01-01

    Characteristics and overlapping aspects in the elaboration of safety analyses for the nuclear and conventional units are presented. The current methods are presented and their limits of applicability characterized. The transferability of individual methods or their elements to the analysis of the reference plant of Wackersdorf is examined and the procedure for the systems analysis is determined. It is of great importance to prove that the essential kinds of incidents and possibilities of release with potential effects in the environment are completely identified. The incidents are divided into basic incidents, which are characterized by superior physical/chemical release mechanisms. An essential objective is to systematize the safety analysis and to summarize the presentation of results. Selection criteria are presented, which allow a limitation of the analysis to essential influencing parameters without removing aspects from the overall safety-relevant statement. Besides the selection criteria, instruments and mathematical models are explained with the help of which the representative and possible incidents covering all potential risks for all areas of the plant, systems and components can be selected. These design-basis accidents (criticality, self-heating, fire, explosion, leakages, earth quakes) are decisive for the determination of potential damaging effects in the environment and thus for the overall statement on the licensability. (orig./HP) [de

  8. Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin

    2015-06-28

    Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

  9. Towards the Development of a Methodology for the Cyber Security Analysis of Safety Related Nuclear Digital I and C Systems

    International Nuclear Information System (INIS)

    Khand, Parvaiz Ahmed; Seong, Poong Hyun

    2007-01-01

    In nuclear power plants the redundant safety related systems are designed to take automatic action to prevent and mitigate accident conditions if the operators and the non-safety systems fail to maintain the plant within normal operating conditions. In case of an event, the failure of these systems has catastrophic consequences. The tendency in the industry over the past 10 years has been to use of commercial of the shelf (COTS) technologies in these systems. COTS software was written with attention to function and performance rather than security. COTS hardware usually designed to fail safe, but security vulnerabilities could be exploited by an attacker to disable the fail safe mechanisms. Moreover, the use of open protocols and operating systems in these technologies make the plants to become vulnerable to a host of cyber attacks. An effective security analysis process is required during all life cycle phases of these systems in order to ensure the security from cyber attacks. We are developing a methodology for the cyber security analysis of safety related nuclear digital I and C Systems. This methodology will cover all phases of development, operation and maintenance processes of software life cycle. In this paper, we will present a security analysis process for the concept stage of software development life cycle

  10. Overview of the ISAM safety assessment methodology

    International Nuclear Information System (INIS)

    Simeonov, G.

    2003-01-01

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

  11. 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.

  12. Safety balance: Analysis of safety systems

    International Nuclear Information System (INIS)

    Delage, M.; Giroux, C.

    1990-12-01

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

  13. Preparation of MARS-LMR Best-Estimate Input for Evaluating the Applicability of Safety Analysis Methodology based on BEPU Technique

    International Nuclear Information System (INIS)

    Kang, Doo-hyuk; Kang, Seok-ju; Suh, Jae-seung; Bae, Sung-won

    2016-01-01

    There is a need to re-establish the methodology at the level of PIRT development for the applicability of an analysis of the uncertainty quantification. Besides, we need to develop the input deck for the MARS-LMR best-estimate safety analysis in order to assess the applicability of safety analysis methodology based on BEPU technique and show the completion of the input deck throughout the results of preliminary computation. The objective of this study is to develop the MARSLMR best-estimate input of the EBR-II nuclear power plant for evaluating the applicability of safety analysis methodology based on BEPU technique for DEC. This code is developed to simulate the sodium thermalhydraulic and neutronic behavior such as reactivity feedbacks for liquid metal cooled fast reactor. A preliminary analysis of the SHRT-45R test was performed using the MARS-LMR code to understand the general behavior of EBR-II SHRT-45R test. The MARS-LMR best-estimate input of the EBR-II was prepared for evaluating the applicability of safety analysis methodology based on BEPU technique for DEC under support from the KAERI. Also, a preliminary analysis of the SHRT-45R test was performed using the MARS-LMR code to understand the general behavior of EBR-II SHRT-45R test

  14. Application of hazard analysis and critical control point methodology and risk-based grading to consumer food safety surveys.

    Science.gov (United States)

    Røssvoll, Elin Halbach; Ueland, Øydis; Hagtvedt, Therese; Jacobsen, Eivind; Lavik, Randi; Langsrud, Solveig

    2012-09-01

    Traditionally, consumer food safety survey responses have been classified as either "right" or "wrong" and food handling practices that are associated with high risk of infection have been treated in the same way as practices with lower risks. In this study, a risk-based method for consumer food safety surveys has been developed, and HACCP (hazard analysis and critical control point) methodology was used for selecting relevant questions. We conducted a nationally representative Web-based survey (n = 2,008), and to fit the self-reported answers we adjusted a risk-based grading system originally developed for observational studies. The results of the survey were analyzed both with the traditional "right" and "wrong" classification and with the risk-based grading system. The results using the two methods were very different. Only 5 of the 10 most frequent food handling violations were among the 10 practices associated with the highest risk. These 10 practices dealt with different aspects of heat treatment (lacking or insufficient), whereas the majority of the most frequent violations involved storing food at room temperature for too long. Use of the risk-based grading system for survey responses gave a more realistic picture of risks associated with domestic food handling practices. The method highlighted important violations and minor errors, which are performed by most people and are not associated with significant risk. Surveys built on a HACCP-based approach with risk-based grading will contribute to a better understanding of domestic food handling practices and will be of great value for targeted information and educational activities.

  15. Review of Overall Safety Manual for space nuclear systems. An evaluation of a nuclear safety analysis methodology for plutonium-fueled space nuclear systems

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, J.; Inhaber, H.

    1984-02-01

    As part of its duties in connection with space missions involving nuclear power sources, the Office of Nuclear Safety (ONS) of the Office of Assistant Secretary for Environmental Protection, Safety, and Emergency Preparedness has been assigned the task of reviewing the Overall Safety Manual (OSM) (memo from B.J. Rock to J.R. Maher, December 1, 1982). The OSM, dated July 1981 and in four volumes, was prepared by NUS Corporation, Rockville, Maryland, for the US Department of Energy. The OSM provides many of the technical models and much of the data which are used by (1) space launch contractors in safety analysis reports and (2) the broader Interagency Nuclear Safety Review Panel (INSRP) safety evaluation reports. If fhs interaction between the OSM, contractors, and INSRP is to work effectively, the OSM must be accurate, comprehensive, understandable, and usable.

  16. Methodology for deriving hydrogeological input parameters for safety-analysis models - application to fractured crystalline rocks of Northern Switzerland

    International Nuclear Information System (INIS)

    Vomvoris, S.; Andrews, R.W.; Lanyon, G.W.; Voborny, O.; Wilson, W.

    1996-04-01

    Switzerland is one of many nations with nuclear power that is seeking to identify rock types and locations that would be suitable for the underground disposal of nuclear waste. A common challenge among these programs is to provide engineering designers and safety analysts with a reasonably representative hydrogeological input dataset that synthesizes the relevant information from direct field observations as well as inferences and model results derived from those observations. Needed are estimates of the volumetric flux through a volume of rock and the distribution of that flux into discrete pathways between the repository zones and the biosphere. These fluxes are not directly measurable but must be derived based on understandings of the range of plausible hydrogeologic conditions expected at the location investigated. The methodology described in this report utilizes conceptual and numerical models at various scales to derive the input dataset. The methodology incorporates an innovative approach, called the geometric approach, in which field observations and their associated uncertainty, together with a conceptual representation of those features that most significantly affect the groundwater flow regime, were rigorously applied to generate alternative possible realizations of hydrogeologic features in the geosphere. In this approach, the ranges in the output values directly reflect uncertainties in the input values. As a demonstration, the methodology is applied to the derivation of the hydrogeological dataset for the crystalline basement of Northern Switzerland. (author) figs., tabs., refs

  17. Analysis of Software Development Methodologies to Build Safety Software Applications for the SATEX-II: A Mexican Experimental Satellite

    Science.gov (United States)

    Aguilar Cisneros, Jorge; Vargas Martinez, Hector; Pedroza Melendez, Alejandro; Alonso Arevalo, Miguel

    2013-09-01

    Mexico is a country where the experience to build software for satellite applications is beginning. This is a delicate situation because in the near future we will need to develop software for the SATEX-II (Mexican Experimental Satellite). SATEX- II is a SOMECyTA's project (the Mexican Society of Aerospace Science and Technology). We have experienced applying software development methodologies, like TSP (Team Software Process) and SCRUM in other areas. Then, we analyzed these methodologies and we concluded: these can be applied to develop software for the SATEX-II, also, we supported these methodologies with SSP-05-0 Standard in particular with ESA PSS-05-11. Our analysis was focusing on main characteristics of each methodology and how these methodologies could be used with the ESA PSS 05-0 Standards. Our outcomes, in general, may be used by teams who need to build small satellites, but, in particular, these are going to be used when we will build the on board software applications for the SATEX-II.

  18. Regional Shelter Analysis Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Dillon, Michael B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dennison, Deborah [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kane, Jave [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Walker, Hoyt [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Miller, Paul [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-08-01

    The fallout from a nuclear explosion has the potential to injure or kill 100,000 or more people through exposure to external gamma (fallout) radiation. Existing buildings can reduce radiation exposure by placing material between fallout particles and exposed people. Lawrence Livermore National Laboratory was tasked with developing an operationally feasible methodology that could improve fallout casualty estimates. The methodology, called a Regional Shelter Analysis, combines the fallout protection that existing buildings provide civilian populations with the distribution of people in various locations. The Regional Shelter Analysis method allows the consideration of (a) multiple building types and locations within buildings, (b) country specific estimates, (c) population posture (e.g., unwarned vs. minimally warned), and (d) the time of day (e.g., night vs. day). The protection estimates can be combined with fallout predictions (or measurements) to (a) provide a more accurate assessment of exposure and injury and (b) evaluate the effectiveness of various casualty mitigation strategies. This report describes the Regional Shelter Analysis methodology, highlights key operational aspects (including demonstrating that the methodology is compatible with current tools), illustrates how to implement the methodology, and provides suggestions for future work.

  19. Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Hao-Tzu [Institute of Nuclear Energy Research, Taoyuan, Taiwan (China). Research Atomic Energy Council; Li, Wan-Yun; Wang, Jong-Rong; Tseng, Yung-Shin; Chen, Hsiung-Chih; Shih, Chunkuan; Chen, Shao-Wen [National Tsing Hua Univ., HsinChu, Taiwan (China). Inst. of Nuclear Engineering and Science

    2017-03-15

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

  20. Application of disturbance analysis methodology to safety related transients in the electrical systems of a nuclear power plant. Report UCLA-ENG-8056

    Energy Technology Data Exchange (ETDEWEB)

    Guarro, S.; Okrent, D.

    1981-08-01

    The present study tries to address the question of whether or not the computerized on-line procedures known under the name of DAS (Disturbance Analysis System) can be usefully and successfully applied to provide timely diagnostics and operational suggestions during the occurrence of a major electrical transient in the auxiliary systems of a nuclear power plant. The perspective of the study is from the plant-safety point of view. A short definition of DAS methodology features and capabilities is presented. A discussion of some of the problems of a general nature that are encountered in DAS safety-oriented applications are also included. The event insufficient power on both emergency buses, with reference to a particular plant dsign (San Onofre 1), is presented. Some transients that have recently occurred in the power supply systems of operating plants are examined. Whether or not a DAS could have successfully dealt with such occurrences is considered.

  1. Safety analysis fundamentals

    International Nuclear Information System (INIS)

    Wright, A.C.D.

    2002-01-01

    This paper discusses the safety analysis fundamentals in reactor design. This study includes safety analysis done to show consequences of postulated accidents are acceptable. Safety analysis is also used to set design of special safety systems and includes design assist analysis to support conceptual design. safety analysis is necessary for licensing a reactor, to maintain an operating license, support changes in plant operations

  2. Safety assessment of adjuvanted vaccines: Methodological considerations

    Science.gov (United States)

    Da Silva, Fernanda Tavares; Di Pasquale, Alberta; Yarzabal, Juan P; Garçon, Nathalie

    2015-01-01

    Adjuvants mainly interact with the innate immune response and are used to enhance the quantity and quality of the downstream adaptive immune response to vaccine antigens. Establishing the safety of a new adjuvant-antigen combination is achieved through rigorous evaluation that begins in the laboratory, and that continues throughout the vaccine life-cycle. The strategy for the evaluation of safety pre-licensure is guided by the disease profile, vaccine indication, and target population, and it is also influenced by available regulatory guidelines. In order to allow meaningful interpretation of clinical data, clinical program methodology should be optimized and standardized, making best use of all available data sources. Post-licensure safety activities are directed by field experience accumulated pre- and post-licensure clinical trial data and spontaneous adverse event reports. Continued evolution of safety evaluation processes that keep pace with advances in vaccine technology and updated communication of the benefit-risk profile is necessary to maintain public confidence in vaccines. PMID:26029975

  3. Safety assessment of adjuvanted vaccines: Methodological considerations.

    Science.gov (United States)

    Tavares Da Silva, Fernanda; Di Pasquale, Alberta; Yarzabal, Juan P; Garçon, Nathalie

    2015-01-01

    Adjuvants mainly interact with the innate immune response and are used to enhance the quantity and quality of the downstream adaptive immune response to vaccine antigens. Establishing the safety of a new adjuvant-antigen combination is achieved through rigorous evaluation that begins in the laboratory, and that continues throughout the vaccine life-cycle. The strategy for the evaluation of safety pre-licensure is guided by the disease profile, vaccine indication, and target population, and it is also influenced by available regulatory guidelines. In order to allow meaningful interpretation of clinical data, clinical program methodology should be optimized and standardized, making best use of all available data sources. Post-licensure safety activities are directed by field experience accumulated pre- and post-licensure clinical trial data and spontaneous adverse event reports. Continued evolution of safety evaluation processes that keep pace with advances in vaccine technology and updated communication of the benefit-risk profile is necessary to maintain public confidence in vaccines.

  4. A Risk Analysis Methodology to Address Human and Organizational Factors in Offshore Drilling Safety: With an Emphasis on Negative Pressure Test

    Science.gov (United States)

    Tabibzadeh, Maryam

    According to the final Presidential National Commission report on the BP Deepwater Horizon (DWH) blowout, there is need to "integrate more sophisticated risk assessment and risk management practices" in the oil industry. Reviewing the literature of the offshore drilling industry indicates that most of the developed risk analysis methodologies do not fully and more importantly, systematically address the contribution of Human and Organizational Factors (HOFs) in accident causation. This is while results of a comprehensive study, from 1988 to 2005, of more than 600 well-documented major failures in offshore structures show that approximately 80% of those failures were due to HOFs. In addition, lack of safety culture, as an issue related to HOFs, have been identified as a common contributing cause of many accidents in this industry. This dissertation introduces an integrated risk analysis methodology to systematically assess the critical role of human and organizational factors in offshore drilling safety. The proposed methodology in this research focuses on a specific procedure called Negative Pressure Test (NPT), as the primary method to ascertain well integrity during offshore drilling, and analyzes the contributing causes of misinterpreting such a critical test. In addition, the case study of the BP Deepwater Horizon accident and their conducted NPT is discussed. The risk analysis methodology in this dissertation consists of three different approaches and their integration constitutes the big picture of my whole methodology. The first approach is the comparative analysis of a "standard" NPT, which is proposed by the author, with the test conducted by the DWH crew. This analysis contributes to identifying the involved discrepancies between the two test procedures. The second approach is a conceptual risk assessment framework to analyze the causal factors of the identified mismatches in the previous step, as the main contributors of negative pressure test

  5. 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)

  6. Review of methodologies for analysis of safety incidents at NPPs. Final report of a co-ordinated research project 1998-2001

    International Nuclear Information System (INIS)

    2002-03-01

    The safe operation of nuclear power plants around the world and the prevention of incidents in these installations remain key concerns for the nuclear community. In this connection, the feedback of operating experience plays a major role: every nuclear power plant or nuclear utility needs to have a system in place for collecting information on unusual events, whether these are incidents or merely deviations from normal operation. Reporting to the regulatory body of important events and lessons learned is normally carried out through the national reporting schemes based on regulatory reporting requirements. The most important lessons learned are further shared internationally, through, for example, the Joint IAEA/NEA Incident Reporting System (IRS) or the event information exchange of the World Association of Nuclear Operators (WANO). In order to properly assess the event, an adequate event investigation methodology has to be applied, which leads to the identification of correct root causes. Once these root causes have been ascertained, appropriate corrective actions can be established and corresponding lessons can be drawn. The overall goal of root cause analysis is the prevention of events or their recurrence and thus the overall improvement in plant safety. In 1998, the IAEA established a co-ordinated research project with the objective of exploring root cause methodologies and techniques currently in use in Member States, evaluating their strengths and limitations and developing criteria for appropriate event investigation methodologies. This report is the outcome of four years of co-ordinated research which involved 15 national and international research organizations

  7. RPP-PRT-58489, Revision 1, One Systems Consistent Safety Analysis Methodologies Report. 24590-WTP-RPT-MGT-15-014

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, Mukesh [URS Professional Solutions LLC, Aiken, SC (United States); Niemi, Belinda [Washington River Protection Solutions, LLC, Richland, WA (United States); Paik, Ingle [Washington River Protection Solutions, LLC, Richland, WA (United States)

    2015-09-02

    In 2012, One System Nuclear Safety performed a comparison of the safety bases for the Tank Farms Operations Contractor (TOC) and Hanford Tank Waste Treatment and Immobilization Plant (WTP) (RPP-RPT-53222 / 24590-WTP-RPT-MGT-12-018, “One System Report of Comparative Evaluation of Safety Bases for Hanford Waste Treatment and Immobilization Plant Project and Tank Operations Contract”), and identified 25 recommendations that required further evaluation for consensus disposition. This report documents ten NSSC approved consistent methodologies and guides and the results of the additional evaluation process using a new set of evaluation criteria developed for the evaluation of the new methodologies.

  8. 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.

  9. Methodology of containment response analysis for the Probabilistic Safety Assessment -PSA of the CAREM-25 nuclear power plant

    International Nuclear Information System (INIS)

    Baron, Jorge

    1996-01-01

    This work is part of the Probabilistic Safety Assessment actually under development for the CAREM-25 Nuclear Power Station, and departs from the accident sequences already obtained and quantified by the Event Trees/Fault Trees techniques. At first, the potential containment failure modes for nuclear stations are listed, based on the experience. Then, the CAREM-25 design peculiarities are analyzed, on their possible influence on the containment behavior during, severe accidents. Then Plan Damage States are then defined. Furthermore, Containment Damage States are also defined, and Containment Event Trees are built for each Plant Damage State. Those sequences considered representative from the annual probability (those which exceed or equal a probability of 1E-09 per year, are used to quantify the combinations of Plant Damage States/Containment Damage States, based on the estimation of a Vulnerability Matrix. (author)

  10. 14 CFR Appendix A to Part 417 - Flight Safety Analysis Methodologies and Products for a Launch Vehicle Flown With a Flight Safety...

    Science.gov (United States)

    2010-01-01

    ... would occur as either a specific value or a probability distribution for time until breakup. (vi... applicability of this appendix during the licensing process. A417.5General A launch operator's flight safety... produce the following dispersed trajectories and describe the distribution of a launch vehicle's position...

  11. 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.

  12. 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

  13. Simplified methodology for Angra 1 containment analysis

    International Nuclear Information System (INIS)

    Neves Conti, T. das; Souza, A.L. de; Sabundjian, G.

    1991-08-01

    A simplified methodology of analysis was developed to simulate a Large Break Loss of Coolant Accident in the Angra 1 Nuclear Power Station. Using the RELAP5/MOD1, RELAP4/MOD5 and CONTEMPT-LT Codes, the time variation of pressure and temperature in the containment was analysed. The obtained data was compared with the Angra 1 Final Safety Analysis Report, and too those calculated by a Detailed Model. The results obtained by this new methodology such as the small computational time of simulation, were satisfactory when getting the preliminary evaluation of the Angra 1 global parameters. (author)

  14. Methodology and applications for organizational safety culture

    International Nuclear Information System (INIS)

    Sakaue, Takeharu; Makino, Maomi

    2004-01-01

    The mission of our activity is making 'guidance of safety culture for understanding and evaluations' which comes in much more useful and making it substantial by clarifying positioning of safety culture within evaluation of the quality management. This is pointed out by 'Discussion on how to implement safety culture sufficiently and possible recommendation' last year by falsification issue of TEPCO (Tokyo Electric Power Company). We have been developing the safety culture evaluation structured by three elements. One is safety culture evaluation support tool (SCET), another is organizational reliability model (ORM), third is system for safety. This paper describes mainly organizational reliability model (ORM) and its applications as well as ticking the system for safety culture within quality management. (author)

  15. Risk analysis methodology survey

    Science.gov (United States)

    Batson, Robert G.

    1987-01-01

    NASA regulations require that formal risk analysis be performed on a program at each of several milestones as it moves toward full-scale development. Program risk analysis is discussed as a systems analysis approach, an iterative process (identification, assessment, management), and a collection of techniques. These techniques, which range from simple to complex network-based simulation were surveyed. A Program Risk Analysis Handbook was prepared in order to provide both analyst and manager with a guide for selection of the most appropriate technique.

  16. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Energy Technology Data Exchange (ETDEWEB)

    Salesa, F.

    2014-02-01

    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  17. Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities

    International Nuclear Information System (INIS)

    Batandjieva, B.; Torres-Vidal, C.

    2002-01-01

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

  18. Methodology of assessment of the impact on the risk of changes of monitoring requirements of safety equipment integrating treatment and analysis of uncertainty of model and parameter

    International Nuclear Information System (INIS)

    Martorell, S.; Villamizar, M.; Marton, I.; Villanueva, J. f.; Carlos, S.; Sanchez, A. I.

    2013-01-01

    The feasibility of the proposed methodology has been demonstrated with the results obtained in a case of application for the analysis of changes in the requirements of maximum time of unavailability of system batteries using an APS level 1. The proposed methodology is consistent with the proposal by the American regulatory guide RG 1,174, applicable to LWR designs in Spanish, although large plants their level of technical development.

  19. METHODOLOGICAL ELEMENTS OF SITUATIONAL ANALYSIS

    Directory of Open Access Journals (Sweden)

    Tetyana KOVALCHUK

    2016-07-01

    Full Text Available The article deals with the investigation of theoretical and methodological principles of situational analysis. The necessity of situational analysis is proved in modern conditions. The notion “situational analysis” is determined. We have concluded that situational analysis is a continuous system study which purpose is to identify dangerous situation signs, to evaluate comprehensively such signs influenced by a system of objective and subjective factors, to search for motivated targeted actions used to eliminate adverse effects of the exposure of the system to the situation now and in the future and to develop the managerial actions needed to bring the system back to norm. It is developed a methodological approach to the situational analysis, its goal is substantiated, proved the expediency of diagnostic, evaluative and searching functions in the process of situational analysis. The basic methodological elements of the situational analysis are grounded. The substantiation of the principal methodological elements of system analysis will enable the analyst to develop adaptive methods able to take into account the peculiar features of a unique object which is a situation that has emerged in a complex system, to diagnose such situation and subject it to system and in-depth analysis, to identify risks opportunities, to make timely management decisions as required by a particular period.

  20. METHODOLOGICAL ELEMENTS OF SITUATIONAL ANALYSIS

    OpenAIRE

    Tetyana KOVALCHUK

    2016-01-01

    The article deals with the investigation of theoretical and methodological principles of situational analysis. The necessity of situational analysis is proved in modern conditions. The notion “situational analysis” is determined. We have concluded that situational analysis is a continuous system study which purpose is to identify dangerous situation signs, to evaluate comprehensively such signs influenced by a system of objective and subjective factors, to search for motivated targeted action...

  1. 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

  2. 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.

  3. K Basin safety analysis

    International Nuclear Information System (INIS)

    Porten, D.R.; Crowe, R.D.

    1994-01-01

    The purpose of this accident safety analysis is to document in detail, analyses whose results were reported in summary form in the K Basins Safety Analysis Report WHC-SD-SNF-SAR-001. The safety analysis addressed the potential for release of radioactive and non-radioactive hazardous material located in the K Basins and their supporting facilities. The safety analysis covers the hazards associated with normal K Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. After a review of the Criticality Safety Evaluation of the K Basin activities, the following postulated events were evaluated: Crane failure and casks dropped into loadout pit; Design basis earthquake; Hypothetical loss of basin water accident analysis; Combustion of uranium fuel following dryout; Crane failure and cask dropped onto floor of transfer area; Spent ion exchange shipment for burial; Hydrogen deflagration in ion exchange modules and filters; Release of Chlorine; Power availability and reliability; and Ashfall

  4. A progressive methodology for structural safety evaluation of gravity dams subjected to floods

    International Nuclear Information System (INIS)

    Leger, P.; Tinawi, R.; Leclerc, M.; Rheault, S.

    1996-01-01

    Failure mechanisms that could be expected for gravity dams subjected to hydrostatic overloads, and flood safety performance criteria for gravity dams, were reviewed. A progressive structural analysis methodology, divided into four structural analysis levels of increasing complexity to evaluate the flood safety of gravity dams was proposed. Application of the proposed methodology was illustrated by parametric analysis performed on a typical 52 m gravity dam. Results of a coupled hydro-fracture analysis of an old masonry dam was also presented to provide qualitative correlations between numerical analysis procedures and traditional observations of failure mechanisms of gravity dams under hydrostatic overload. 9 refs., 8 figs

  5. 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

  6. 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.

  7. Recent Methodology in Ginseng Analysis

    Science.gov (United States)

    Baek, Seung-Hoon; Bae, Ok-Nam; Park, Jeong Hill

    2012-01-01

    As much as the popularity of ginseng in herbal prescriptions or remedies, ginseng has become the focus of research in many scientific fields. Analytical methodologies for ginseng, referred to as ginseng analysis hereafter, have been developed for bioactive component discovery, phytochemical profiling, quality control, and pharmacokinetic studies. This review summarizes the most recent advances in ginseng analysis in the past half-decade including emerging techniques and analytical trends. Ginseng analysis includes all of the leading analytical tools and serves as a representative model for the analytical research of herbal medicines. PMID:23717112

  8. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

    Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...... the same system model and that this model is formalized in a real-time, interval logic, based on a conventional dynamic systems model with a state over time. The three safety analysis techniques are interpreted in this model and it is shown how to derive safety requirements for components of a system....

  9. Safety analysis for 'Fugen'

    International Nuclear Information System (INIS)

    1997-10-01

    The improvement of safety in nuclear power stations is an important proposition. Therefore also as to the safety evaluation, it is important to comprehensively and systematically execute it by referring to the operational experience and the new knowledge which is important for the safety throughout the period of use as well as before the construction and the start of operation of nuclear power stations. In this report, the results when the safety analysis for ''Fugen'' was carried out by referring to the newest technical knowledge are described. As the result, it was able to be confirmed that the safety of ''Fugen'' has been secured by the inherent safety and the facilities which were designed for securing the safety. The basic way of thinking on the safety analysis including the guidelines to be conformed to is mentioned. As to the abnormal transient change in operation and accidents, their definition, the events to be evaluated and the standards for judgement are reported. The matters which were taken in consideration at the time of the analysis are shown. The computation programs used for the analysis were REACT, HEATUP, LAYMON, FATRAC, SENHOR, LOTRAC, FLOOD and CONPOL. The analyses of the abnormal transient change in operation and accidents are reported on the causes, countermeasures, protective functions and results. (K.I.)

  10. Methodology of safety assessment for radioactive waste disposal

    International Nuclear Information System (INIS)

    Matsuzuru, Hideo; Kimura, Hideo

    1991-01-01

    The Japan Atomic Energy Research Institute (JAERI) is conducting an extensive R and D program to develop a safety assessment methodology to evaluate environmental consequences associated with geological disposal of a high-level radioactive waste, and also to elucidate a generic feasibility of the geological disposal in Japan. The paper describes the current R and D activities in the JAERI to develop an interim version of the methodology based on a normal evolution scenario, and also to validate models used in the methodology. (author)

  11. Integrated Response Time Evaluation Methodology for the Nuclear Safety Instrumentation System

    Science.gov (United States)

    Lee, Chang Jae; Yun, Jae Hee

    2017-06-01

    Safety analysis for a nuclear power plant establishes not only an analytical limit (AL) in terms of a measured or calculated variable but also an analytical response time (ART) required to complete protective action after the AL is reached. If the two constraints are met, the safety limit selected to maintain the integrity of physical barriers used for preventing uncontrolled radioactivity release will not be exceeded during anticipated operational occurrences and postulated accidents. Setpoint determination methodologies have actively been developed to ensure that the protective action is initiated before the process conditions reach the AL. However, regarding the ART for a nuclear safety instrumentation system, an integrated evaluation methodology considering the whole design process has not been systematically studied. In order to assure the safety of nuclear power plants, this paper proposes a systematic and integrated response time evaluation methodology that covers safety analyses, system designs, response time analyses, and response time tests. This methodology is applied to safety instrumentation systems for the advanced power reactor 1400 and the optimized power reactor 1000 nuclear power plants in South Korea. The quantitative evaluation results are provided herein. The evaluation results using the proposed methodology demonstrate that the nuclear safety instrumentation systems fully satisfy corresponding requirements of the ART.

  12. Advanced bridge safety initiative, task 3 : slab bridge load rating using AASHTO methodology and finite element analysis - an analysis of 20 bridges.

    Science.gov (United States)

    2011-12-01

    Current AASHTO provisions for the conventional load rating of flat slab bridges rely on the equivalent strip method : of analysis for determining live load effects, this is generally regarded as overly conservative by many professional : engineers. A...

  13. Methodology of Credit Analysis Development

    Directory of Open Access Journals (Sweden)

    Slađana Neogradi

    2017-12-01

    Full Text Available The subject of research presented in this paper refers to the definition of methodology for the development of credit analysis in companies and its application in lending operations in the Republic of Serbia. With the developing credit market, there is a growing need for a well-developed risk and loss prevention system. In the introduction the process of bank analysis of the loan applicant is presented in order to minimize and manage the credit risk. By examining the subject matter, the process of processing the credit application is described, the procedure of analyzing the financial statements in order to get an insight into the borrower's creditworthiness. In the second part of the paper, the theoretical and methodological framework is presented applied in the concrete company. In the third part, models are presented which banks should use to protect against exposure to risks, i.e. their goal is to reduce losses on loan operations in our country, as well as to adjust to market conditions in an optimal way.

  14. METHODOLOGY OF MATHEMATICAL ANALYSIS IN POWER NETWORK

    OpenAIRE

    Jerzy Szkutnik; Mariusz Kawecki

    2008-01-01

    Power distribution network analysis is taken into account. Based on correlation coefficient authors establish methodology of mathematical analysis useful in finding substations bear responsibility for power stoppage. Also methodology of risk assessment will be carried out.

  15. Stakeholder analysis methodologies resource book

    Energy Technology Data Exchange (ETDEWEB)

    Babiuch, W.M.; Farhar, B.C.

    1994-03-01

    Stakeholder analysis allows analysts to identify how parties might be affected by government projects. This process involves identifying the likely impacts of a proposed action and stakeholder groups affected by that action. Additionally, the process involves assessing how these groups might be affected and suggesting measures to mitigate any adverse effects. Evidence suggests that the efficiency and effectiveness of government actions can be increased and adverse social impacts mitigated when officials understand how a proposed action might affect stakeholders. This report discusses how to conduct useful stakeholder analyses for government officials making decisions on energy-efficiency and renewable-energy technologies and their commercialization. It discusses methodological issues that may affect the validity and reliability of findings, including sampling, generalizability, validity, ``uncooperative`` stakeholder groups, using social indicators, and the effect of government regulations. The Appendix contains resource directories and a list of specialists in stakeholder analysis and involvement.

  16. Safety studies on Korean fusion DEMO plant using Integrated Safety Assessment Methodology: Part 2

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    Highlights: • The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. • The concepts of Integrated Safety Assessment Methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. • Phenomena Identification Ranking Table (PIRT) and Objective Provision Tree (OPT) were performed and updated. • This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. - Abstract: The purpose of this paper is to investigate safety issues using Integrated Safety Assessment Methodology (ISAM) proposed by Generation IV Forum Risk and Safety Working Group (RSWG) for Korean fusion DEMO plant (K-DEMO). In ongoing nuclear energy research such as Generation IV fission power plant (GEN-IV), new methodology based on Technology-Neutral Framework (TNF) has been applied for safety assessment. In this methodology, design and regulatory requirements for safety of nuclear power plants are considered simultaneously. The design based on regulatory requirements can save resource, time, and manpower while maintaining high level safety. ISAM is one of the options to apply TNF in K-DEMO. We have performed safety studies for K-DEMO using Phenomena Identification and Ranking Table (PIRT) and Objective Provision Tree (OPT) which are constitutive part of ISAM. Considering the design phase of K-DEMO, the current study focused on PIRT and OPT for K-DEMO. Results have been reviewed and updated by Korean fusion advisory group after considering the views of specialists from domestic universities, industries, and national institutes in South Korea.

  17. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

  18. 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

  19. 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

  20. 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.

  1. Review of methodologies used in safety analyses of nuclear waste disposal

    International Nuclear Information System (INIS)

    Robinson, P.C.

    1984-06-01

    The methodologies used in safety analyses of nuclear waste disposal are reviewed. Particular reference is made to the KBS-3 proposals and to the WP-Cave concept. These methodologies are aimed mainly at reducing uncertainties in the predictions of health consequences from nuclear waste repositories, and quantifying the uncertainties that remain. The steps of a safety analysis are described in turn and the currently used techniques are discussed. Recommendations are made as to the use and development of the techniques and the need for new techniques. (author)

  2. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  3. Software safety hazard analysis

    International Nuclear Information System (INIS)

    Lawrence, J.D.

    1996-02-01

    Techniques for analyzing the safety and reliability of analog-based electronic protection systems that serve to mitigate hazards in process control systems have been developed over many years, and are reasonably well understood. An example is the protection system in a nuclear power plant. The extension of these techniques to systems which include digital computers is not well developed, and there is little consensus among software engineering experts and safety experts on how to analyze such systems. One possible technique is to extend hazard analysis to include digital computer-based systems. Software is frequently overlooked during system hazard analyses, but this is unacceptable when the software is in control of a potentially hazardous operation. In such cases, hazard analysis should be extended to fully cover the software. A method for performing software hazard analysis is proposed in this paper

  4. Methodology of assessment of the impact on the risk of changes in peak times of unavailability of safety equipment integrating treatment and analysis of uncertainties

    International Nuclear Information System (INIS)

    Martorell, S.; Villamizar, M.; Marton, I.; Villanueva, J. f.; Carlos, S.; Sanchez, A. I.

    2013-01-01

    Surveillance of safety equipment requirements form parts of the technical specifications of Performance included in the bases of operation of nuclear power plants license, so they are center of attention in the study of improvements to the safety of the operation of the plant using different methods, both traditional deterministic as probabilistic.

  5. Bowtie Risk Management methodology and Modern Nuclear Safety Reports

    International Nuclear Information System (INIS)

    Ilizastigui Pérez, F.

    2016-01-01

    The Safety Report (SR) plays a crucial role within the nuclear licensing regime as the principal means for demonstrating the adequacy of safety analysis for a nuclear facility to ensure that it can be constructed, operated, maintained, shut down, and decommissioned safely and in compliance with applicable laws and regulations. It serves as the basis for granting authorizations for the commencement of the main stages of the facility’s life cycle as well as decision-making processes related to safety. Historically, the majority of nuclear safety reports have operated under rather prescriptive regimes, with emphasis placed on demonstrations of the robustness of the facility’s design (design safety) against prescriptive technical requirements set by the regulatory body, and less attention paid to demonstrating the adequacy and effectiveness of Operator’s management system for managing risks to daily operation.

  6. Statistical considerations on safety analysis

    International Nuclear Information System (INIS)

    Pal, L.; Makai, M.

    2004-01-01

    statement is true. In some cases statistical aspects of safety are misused, where the number of runs for several outputs is correct only for statistically independent outputs, or misinterpreted. We do not know the probability distribution of the output variables subjected to safety limitations. At the same time in some asymmetric distributions the 0.95/0.95 methodology simply fails: if we repeat the calculations in many cases we would get a value higher than the basic value, which means the limit violation in the calculation becomes more and more probable in the repeated analysis. Consequent application of order statistics or the application of the sign test may offer a way out of the present situation. The authors are also convinced that efforts should be made to study the statistics of the output variables, and to study the occurrence of chaos in the analyzed cases. All these observations should influence, in safety analysis, the application of best estimate methods, and underline the opinion that any realistic modeling and simulation of complex systems must include the probabilistic features of the system and the environment

  7. Causal Meta-Analysis : Methodology and Applications

    NARCIS (Netherlands)

    Bax, L.J.

    2009-01-01

    Meta-analysis is a statistical method to summarize research data from multiple studies in a quantitative manner. This dissertation addresses a number of methodological topics in causal meta-analysis and reports the development and validation of meta-analysis software. In the first (methodological)

  8. 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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. 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.

  11. 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

  12. A methodology for radiological accidents analysis in industrial gamma radiography

    International Nuclear Information System (INIS)

    Silva, F.C.A. da.

    1990-01-01

    A critical review of 34 published severe radiological accidents in industrial gamma radiography, that happened in 15 countries, from 1960 to 1988, was performed. The most frequent causes, consequences and dose estimation methods were analysed, aiming to stablish better procedures of radiation safety and accidents analysis. The objective of this work is to elaborate a radiological accidents analysis methodology in industrial gamma radiography. The suggested methodology will enable professionals to determine the true causes of the event and to estimate the dose with a good certainty. The technical analytical tree, recommended by International Atomic Energy Agency to perform radiation protection and nuclear safety programs, was adopted in the elaboration of the suggested methodology. The viability of the use of the Electron Gamma Shower 4 Computer Code System to calculate the absorbed dose in radiological accidents in industrial gamma radiography, mainly at sup(192)Ir radioactive source handling situations was also studied. (author)

  13. 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

  14. 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.

  15. 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.

  16. Additional methodology development for statistical evaluation of reactor safety analyses

    International Nuclear Information System (INIS)

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

    1977-03-01

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

  17. Preclosure Safety Analysis Guide

    International Nuclear Information System (INIS)

    D.D. Orvis

    2003-01-01

    A preclosure safety analysis (PSA) is a required element of the License Application (LA) for the high- level radioactive waste repository at Yucca Mountain. This guide provides analysts and other Yucca Mountain Repository Project (the Project) personnel with standardized methods for developing and documenting the PSA. The definition of the PSA is provided in 10 CFR 63.2, while more specific requirements for the PSA are provided in 10 CFR 63.112, as described in Sections 1.2 and 2. The PSA requirements described in 10 CFR Part 63 were developed as risk-informed performance-based regulations. These requirements must be met for the LA. The PSA addresses the safety of the Geologic Repository Operations Area (GROA) for the preclosure period (the time up to permanent closure) in accordance with the radiological performance objectives of 10 CFR 63.111. Performance objectives for the repository after permanent closure (described in 10 CFR 63.113) are not mentioned in the requirements for the PSA and they are not considered in this guide. The LA will be comprised of two phases: the LA for construction authorization (CA) and the LA amendment to receive and possess (R and P) high-level radioactive waste (HLW). PSA methods must support the safety analyses that will be based on the differing degrees of design detail in the two phases. The methods described herein combine elements of probabilistic risk assessment (PRA) and deterministic analyses that comprise a risk-informed performance-based safety analysis. This revision to the PSA guide was prepared for the following objectives: (1) To correct factual and typographical errors. (2) To provide additional material suggested from reviews by the Project, the U.S. Department of Energy (DOE), and U.S. Nuclear Regulatory Commission (NRC) Staffs. (3) To update material in accordance with approaches and/or strategies adopted by the Project. In addition, a principal objective for the planned revision was to ensure that the methods and

  18. Methodology for Validating Building Energy Analysis Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Judkoff, R.; Wortman, D.; O' Doherty, B.; Burch, J.

    2008-04-01

    The objective of this report was to develop a validation methodology for building energy analysis simulations, collect high-quality, unambiguous empirical data for validation, and apply the validation methodology to the DOE-2.1, BLAST-2MRT, BLAST-3.0, DEROB-3, DEROB-4, and SUNCAT 2.4 computer programs. This report covers background information, literature survey, validation methodology, comparative studies, analytical verification, empirical validation, comparative evaluation of codes, and conclusions.

  19. Development of a Long Term Cooling Analysis Methodology Using Rappel

    International Nuclear Information System (INIS)

    Lee, S. I.; Jeong, J. H.; Ban, C. H.; Oh, S. J.

    2012-01-01

    Since the revision of the 10CFR50.46 in 1988, which allowed BE (Best-Estimate) method in analyzing the safety performance of a nuclear power plant, safety analysis methodologies have been changed continuously from conservative EM (Evaluation Model) approaches to BE ones. In this context, LSC (Long-Term core Cooling) methodologies have been reviewed by the regulatory bodies of USA and Korea. Some non-conservatism and improperness of the old methodology have been identified, and as a result, USNRC suspended the approval of CENPD-254-P-A which is the old LSC methodology for CE-designed NPPs. Regulatory bodies requested to remove the non-conservatisms and to reflect system transient behaviors in all the LSC methodologies used. In the present study, a new LSC methodology using RELAP5 is developed. RELAP5 and a newly developed code, BACON (Boric Acid Concentration Of Nuclear power plant) are used to calculate the transient behavior of the system and the boric acid concentration, respectively. Full range of break spectrum is considered and the applicability is confirmed through plant demonstration calculations. The result shows a good comparison with the old-fashioned ones, therefore, the methodology could be applied with no significant changes of current LSC plans

  20. METHODOLOGICAL STRATEGIES FOR TEXTUAL DATA ANALYSIS:

    Directory of Open Access Journals (Sweden)

    Juan Carlos Rincón-Vásquez

    2011-12-01

    Full Text Available This paper presents a classification methodology for studies of textual data. Thisclassification is based on the two predominant methodologies for social scienceresearch: qualitative and quantitative. The basic assumption is that the researchprocess involves three main features: 1 Structure Research, 2 Collection of informationand, 3 Analysis and Interpretation of Data. In each, there are generalguidelines for textual studies.

  1. Constructive Analysis : A Study in Epistemological Methodology

    DEFF Research Database (Denmark)

    Ahlström, Kristoffer

    The present study is concerned the viability of the primary method in contemporary philosophy, i.e., conceptual analysis. Starting out by tracing the roots of this methodology to Platonic philosophy, the study questions whether such a methodology makes sense when divorced from Platonic philosophy...

  2. Simplified methodology for analysis of Angra-1 containing

    International Nuclear Information System (INIS)

    Neves Conti, T. das; Souza, A.L. de; Sabundjian, G.

    1988-01-01

    A simplified methodology of analysis was developed to simulate a Large Break Loss of Coolant Accident in the Angra 1 Nuclear Power Station. Using the RELAP5/MOD1, RELAP4/MOD5 and CONTEMPT-LT Codes, the time the variation of pressure and temperature in the containment was analysed. The obtained data was compared with the Angra 1 Final Safety Analysis Report, and too those calculated by a Detailed Model. The results obtained by this new methodology such as the small computational time of simulation, were satisfactory when getting the preliminar avaliation of the Angra 1 global parameters. (author) [pt

  3. Deep Borehole Disposal Safety Analysis.

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  4. Methodology for calculating guideline concentrations for safety shot sites

    International Nuclear Information System (INIS)

    1997-06-01

    Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination

  5. Methodology for calculating guideline concentrations for safety shot sites

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination.

  6. Rat sperm motility analysis: methodologic considerations

    Science.gov (United States)

    The objective of these studies was to optimize conditions for computer-assisted sperm analysis (CASA) of rat epididymal spermatozoa. Methodologic issues addressed include sample collection technique, sampling region within the epididymis, type of diluent medium used, and sample c...

  7. Aging evaluation methodology of periodic safety review in Korea

    International Nuclear Information System (INIS)

    Park, Heung-Bae; Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In Korea plant lifetime management (PLIM) study for Kori Unit 1 has been performed since 1993. Meanwhile, periodic safety review (PSR) for all operating nuclear power plants (NPPs) has been started with Kori Unit 1 since 2000 per IAEA recommendation. The evaluation period is 10 years, and safety (evaluation) factors are 11 per IAEA guidelines as represented in table 1. The relationship between PSR factors and PLIM is also represented. Among these factors evaluation of 'management of aging' is one of the most important and difficult factor. This factor is related to 'actual condition of the NPP', 'use of experience from other nuclear NPPs and of research findings', and 'management of aging'. The object of 'management of aging' is to obtain plant safety through identifying actual condition of system, structure and components (SSCs) and evaluating aging phenomena and residual life of SSCs using operating experience and research findings. The paper describes the scope and procedure of valuation of 'management of aging', such as, screening criteria of SSCs, Code and Standards, evaluation of SSCs and safety issues as represented. Evaluating SSCs are determined using final safety analysis report (FSAR) and power unit maintenance system for Nuclear Ver. III (PUMAS/N-III). The screening criteria of SSCs are safety-related items (quality class Q), safety-impact items (quality class T), backfitting rule items (fire protection (10CFR50.48), environmental qualification (10CFR50.49), pressurized thermal shock (10CFR50.61), anticipated transient without scram (10CFR50.62), and station blackout (10CFR50.63)) and regulating authority requiring items[1∼3]. The purpose of review of Code and Standards is identifying actual condition of the NPP and evaluating aging management using effective Code and Standards corresponding to reactor facilities. Code and Standards is composed of regulating laws, FSAR items, administrative actions, regulating actions, agreement items, and other

  8. Application of the integrated safety assessment methodology to sequences with loss of component cooling water system

    International Nuclear Information System (INIS)

    Queral, C.; Ibanez, L.; Hortal, J.; Izquierdo, J.M.; Sanchez-Perea, M.; Melendez, E.

    2011-01-01

    The Integrated Safety Assessment (ISA) methodology, developed by the Consejo de Seguridad Nuclear (CSN), Spanish Nuclear Regulatory Body, has been applied to a thermo-hydraulic analysis of Zion NPP for sequences with loss of the Component Cooling Water System (CCWS) in the context of the NEA application exercise SM2A (Safety Margin Application and Assessment). The ISA methodology allows among others to obtain the damage domain (the region where the PCT limit is exceeded) for each sequence of the dynamic event tree as a function of the uncertain times of operator actuations (secondary side cooling and recovery of CCWS). Sequences with available and unavailable accumulators (ACC) have been analyzed in order to distinguish all the possibilities (ACC demanded and successful, ACC demanded and failed and ACC not demanded). In this work, this damage domain and some parameters of every sequence have been obtained from the results of the simulations performed with TRACE code, these data as well as the time-density probability distributions of the considered uncertain parameters (manual actions) are used to obtain the exceedance frequency of the particular safety limit or damage limit. . The results show the feasibility of ISA methodology in order to obtain accurate enough regions of uncertain parameters (time delays or physical parameters) where the particular safety limit of interest is exceeded, as well as the frequency of exceeding this limit. (author)

  9. Reliability analysis of PLC safety equipment

    Energy Technology Data Exchange (ETDEWEB)

    Yu, J.; Kim, J. Y. [Chungnam Nat. Univ., Daejeon (Korea, Republic of)

    2006-06-15

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system.

  10. Reliability analysis of PLC safety equipment

    International Nuclear Information System (INIS)

    Yu, J.; Kim, J. Y.

    2006-06-01

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system

  11. Methodology of human factor analysis

    International Nuclear Information System (INIS)

    Griffon-Fouco, M.

    1988-01-01

    The paper describes the manner in which the Heat Production Department of Electricite de France analyses the human factors in nuclear power plants. After describing the teams and structures set up to deal with this subject, the paper emphasizes two types of methods which are used, most often in complementary fashion: (1) an a posteriori analysis, which consists in studying the events which have taken place at nuclear power plants and in seeking the deepseated causes so as to prevent their reoccurrence in future; (2) an a priori analysis, which consists in analysing a work situation and in detecting all its potential failure factors so as to prevent their resulting once again in dysfunctions of the facility. To illustrate these two types of analysis, two examples are given: first, a study of the telephonic communications between operators in one plant (in which the a posteriori and a priori analysis are developed) and, second, a study of stress in a plant (in which only the a priori analysis is used). (author). 1 tab

  12. Proposal of Integrated Safety Assessment Methodology for Embedded System

    International Nuclear Information System (INIS)

    Sun, Wei; Kageyama, Makoto; Kanemoto, Shigeru

    2011-01-01

    To do risk analysis and risk evaluation for complicated safety critical embedded systems, there are three things should be paid a good attention: 1) an efficient and integrated model expression of embedded systems: 2) systematic risk analysis based on integrated system model: 3) quantitative risk evaluation for software and hardware integrated system. In this paper, taken electric water boiler as a target system, a proposal of risk analysis and risk evaluation for the embedded system is presented to meet these three purposes. In risk analysis, MFM is used and FT is generated automatically from MFM following some rules: And in risk evaluation, GO-FLOW is used to evaluate the reliability of sensors. And furthermore, FIT is applied to evaluate the safety software logic based on the diversity design concept. Although the electric water boiler is a simple example, it includes the key components of the embedded system like sensors, actuators, and software component. So, the process of modeling, analysis, and evaluation could be applied to other kinds of complicated embedded systems

  13. Disposal criticality analysis methodology's principal isotope burnup credit

    International Nuclear Information System (INIS)

    Doering, T.W.; Thomas, D.A.

    2001-01-01

    This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)

  14. 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

  15. The influence of sodium fires on LMFBRs safety analysis

    International Nuclear Information System (INIS)

    Justin, F.

    1979-01-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs

  16. Disposal Criticality Analysis Methodology Topical Report

    International Nuclear Information System (INIS)

    D.G. Horton

    1998-01-01

    The fundamental objective of this topical report is to present the planned risk-informed disposal criticality analysis methodology to the NRC to seek acceptance that the principles of the methodology and the planned approach to validating the methodology are sound. The design parameters and environmental assumptions within which the waste forms will reside are currently not fully established and will vary with the detailed waste package design, engineered barrier design, repository design, and repository layout. Therefore, it is not practical to present the full validation of the methodology in this report, though a limited validation over a parameter range potentially applicable to the repository is presented for approval. If the NRC accepts the methodology as described in this section, the methodology will be fully validated for repository design applications to which it will be applied in the License Application and its references. For certain fuel types (e.g., intact naval fuel), any processes, criteria, codes or methods different from the ones presented in this report will be described in separate addenda. These addenda will employ the principles of the methodology described in this report as a foundation. Departures from the specifics of the methodology presented in this report will be described in the addenda

  17. Methodology for assessing the safety of vehicles on the Technical level and term of operation

    OpenAIRE

    Komarov, V.

    2007-01-01

    The paper presents the results of research in the field of safety evaluation and safety control of vehicles. Different approaches to determining the safety of vehicles are analyzed. Safety is considered as a complex property of vehicle's quality. The concept of stability of properties is introduced. The concept of technical level of vehicle and its impact on safety is clarified. The paper presents the methodology of safety evaluation of vehicle according to its technical level and operation l...

  18. Safety analysis of the nuclear chemistry Building 151

    International Nuclear Information System (INIS)

    Kvam, D.

    1984-01-01

    This report summarizes the results of a safety analysis that was done on Building 151. The report outlines the methodology, the analysis, and the findings that led to the low hazard classification. No further safety evaluation is indicated at this time. 5 tables

  19. Comparative analysis of proliferation resistance assessment methodologies

    International Nuclear Information System (INIS)

    Takaki, Naoyuki; Kikuchi, Masahiro; Inoue, Naoko; Osabe, Takeshi

    2005-01-01

    Comparative analysis of the methodologies was performed based on the discussions in the international workshop on 'Assessment Methodology of Proliferation Resistance for Future Nuclear Energy Systems' held in Tokyo, on March 2005. Through the workshop and succeeding considerations, it is clarified that the proliferation resistance assessment methodologies are affected by the broader nuclear options being pursued and also by the political situations of the state. Even the definition of proliferation resistance, despite the commonality of fundamental issues, derives from perceived threat and implementation circumstances inherent to the larger programs. Deep recognitions of the 'difference' among communities would help us to make further essential and progressed discussion with harmonization. (author)

  20. The current level safety analysis of the nuclear power plant units based on the safety control charts using

    International Nuclear Information System (INIS)

    Aronov, I.Z.; Biryukova, N.F.; Grozovskij, G.I.

    1994-01-01

    The methodology of graphic analysis of current safery of a NPP unit using the safety control charts, on which the values of unit safety reiting and its statistical characteristics are plotted, is suggested. The conclusions about the current safety level are made basing on the reiting change dynamics analysis. Examples of the analysis are discussed

  1. Nondestructive assay methodologies in nuclear forensics analysis

    International Nuclear Information System (INIS)

    Tomar, B.S.

    2016-01-01

    In the present chapter, the nondestructive assay (NDA) methodologies used for analysis of nuclear materials as a part of nuclear forensic investigation have been described. These NDA methodologies are based on (i) measurement of passive gamma and neutrons emitted by the radioisotopes present in the nuclear materials, (ii) measurement of gamma rays and neutrons emitted after the active interrogation of the nuclear materials with a source of X-rays, gamma rays or neutrons

  2. Update of Part 61 impacts analysis methodology

    International Nuclear Information System (INIS)

    Oztunali, O.I.; Roles, G.W.

    1986-01-01

    The US Nuclear Regulatory Commission is expanding the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of costs and impacts of disposal of waste that exceeds Class C concentrations. The project includes updating the computer codes that comprise the methodology, reviewing and updating data assumptions on waste streams and disposal technologies, and calculation of costs for small as well as large disposal facilities. This paper outlines work done to date on this project

  3. Update of Part 61 impacts analysis methodology

    International Nuclear Information System (INIS)

    Oztunali, O.I.; Roles, G.W.; US Nuclear Regulatory Commission, Washington, DC 20555)

    1985-01-01

    The US Nuclear Regulatory Commission is expanding the impacts analysis methodology used during the development of the 10 CFR Part 61 regulation to allow improved consideration of costs and impacts of disposal of waste that exceeds Class C concentrations. The project includes updating the computer codes that comprise the methodology, reviewing and updating data assumptions on waste streams and disposal technologies, and calculation of costs for small as well as large disposal facilities. This paper outlines work done to date on this project

  4. Safety analysis and synthesis using fuzzy sets and evidential reasoning

    International Nuclear Information System (INIS)

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

    1995-01-01

    This paper presents a new methodology for safety analysis and synthesis of a complex engineering system with a structure that is capable of being decomposed into a hierarchy of levels. In this methodology, fuzzy set theory is used to describe each failure event and an evidential reasoning approach is then employed to synthesise the information thus produced to assess the safety of the whole system. Three basic parameters--failure likelihood, consequence severity and failure consequence probability, are used to analyse a failure event. These three parameters are described by linguistic variables which are characterised by a membership function to the defined categories. As safety can also be clearly described by linguistic variables referred to as the safety expressions, the obtained fuzzy safety score can be mapped back to the safety expressions which are characterised by membership functions over the same categories. This mapping results in the identification of the safety of each failure event in terms of the degree to which the fuzzy safety score belongs to each of the safety expressions. Such degrees represent the uncertainty in safety evaluations and can be synthesised using an evidential reasoning approach so that the safety of the whole system can be evaluated in terms of these safety expressions. Finally, a practical engineering example is presented to demonstrate the proposed safety analysis and synthesis methodology

  5. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

    Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

    1999-03-01

    Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

  6. Current status and new trends in the methodology of safety assessment for near surface disposal facilities

    International Nuclear Information System (INIS)

    Ilie, Petre; Didita, Liana; Danchiv, Alexandru

    2008-01-01

    The main goal of this paper is to present the status of the safety assessment methodology at the end of IAEA CRP 'Application of Safety Assessment Methodology for Near-Surface Radioactive Waste Disposal Facilities (ASAM)', and the new trends outlined at the launch of the follow-up project 'Practical Implementation of Safety Assessment Methodologies in a Context of Safety Case of Near-Surface Facilities (PRISM)'. Over the duration of the ASAM project, the ISAM methodology was confirmed as providing a good framework for conducting safety assessment calculations. In contrast, ASAM project identified the limitations of the ISAM methodology as currently formulated. The major limitations are situated in the area of the use of safety assessment for informing practical decisions about alternative waste and risk management strategies for real disposal sites. As a result of the limitation of the ISAM methodology, the PRISM project is established as an extension of the ISAM and ASAM projects. Based on the outcomes of the ASAM project, the main objective of the PRISM project are: 1 - to develop an overview of what constitutes an adequate safety case and safety assessment with a view to supporting decision making processes; 2 - to provide practical illustrations of how the safety assessment methodology could be used for addressing some specific issues arising from the ASAM project and national cases; 3 - to support harmonization with the IAEA's international safety standards. (authors)

  7. New Methodology for a Comprehensive Modular Safety Control System in a Cyclotron Site

    International Nuclear Information System (INIS)

    Kaufman, Y.; Kravitz, M.; Arad, M.; Osovizky, A.; Paran, J.; Sarussi, B.; Ellenbogen, M.; Tal, N.

    2004-01-01

    This Paper describes a new methodology for a comprehensive modular Safety Control System (SCS), for a cyclotron site. The developed SCS is a modular approach for controlling the production procedures, safety conditions and documentation aspects in the Cyclotron site. Usually, the safety conditions in cyclotron sites are maintained by a variety of sensors. The cyclotron is supplied from the manufacturer with a self-integrated control system for its operation, yet the comprehensive SCS has to be defined and setup by the customer. Therefore, customers face a lot of integration problems in trying to combine all the signals from the different safety systems such as radiation monitoring, environmental and access control, in order to maintain proper safety working conditions. The presented SCS design provides main user interface and the complete safety solution required by including preset control logic definitions and open logic for specific user applications. The knowledge for the preset control logic definitions was gathered in previous projects. Failure Mode and Effects Analysis (FMEA) method has been implemented on the SCS to analyze the potential failure modes and their impact on the product reliability

  8. Exploring participatory methodologies in organizational discourse analysis

    DEFF Research Database (Denmark)

    Plotnikof, Mie

    2014-01-01

    Recent debates in the field of organizational discourse analysis stress contrasts in approaches as single-level vs. multi-level, critical vs. participatory, discursive vs. material methods. They raise methodological issues of combining such to embrace multimodality in order to enable new contribu......Recent debates in the field of organizational discourse analysis stress contrasts in approaches as single-level vs. multi-level, critical vs. participatory, discursive vs. material methods. They raise methodological issues of combining such to embrace multimodality in order to enable new......-and practices by dealing with challenges of methodological overview, responsive creativity and identity-struggle. The potentials hereof are demonstrated and discussed with cases of two both critical and co-creative practices, namely ‘organizational modelling’ and ‘fixed/unfixed positioning’ from fieldwork...

  9. 2007 motor vehicle occupant safety survey. Volume 1, Methodology report

    Science.gov (United States)

    2008-07-01

    The 2007 Motor Vehicle Occupant Safety Survey was the sixth in a series of periodic national telephone surveys on occupant protection issues conducted for the National Highway Traffic Safety Administration (NHTSA). Data collection was conducted by Sc...

  10. Application of human reliability analysis methodology of second generation

    International Nuclear Information System (INIS)

    Ruiz S, T. de J.; Nelson E, P. F.

    2009-10-01

    The human reliability analysis (HRA) is a very important part of probabilistic safety analysis. The main contribution of HRA in nuclear power plants is the identification and characterization of the issues that are brought together for an error occurring in the human tasks that occur under normal operation conditions and those made after abnormal event. Additionally, the analysis of various accidents in history, it was found that the human component has been a contributing factor in the cause. Because of need to understand the forms and probability of human error in the 60 decade begins with the collection of generic data that result in the development of the first generation of HRA methodologies. Subsequently develop methods to include in their models additional performance shaping factors and the interaction between them. So by the 90 mid, comes what is considered the second generation methodologies. Among these is the methodology A Technique for Human Event Analysis (ATHEANA). The application of this method in a generic human failure event, it is interesting because it includes in its modeling commission error, the additional deviations quantification to nominal scenario considered in the accident sequence of probabilistic safety analysis and, for this event the dependency actions evaluation. That is, the generic human failure event was required first independent evaluation of the two related human failure events . So the gathering of the new human error probabilities involves the nominal scenario quantification and cases of significant deviations considered by the potential impact on analyzed human failure events. Like probabilistic safety analysis, with the analysis of the sequences were extracted factors more specific with the highest contribution in the human error probabilities. (Author)

  11. Bow tie methodology: a tool to enhance the visibility and understanding of nuclear safety cases

    International Nuclear Information System (INIS)

    Vannerem, Marc

    2013-01-01

    improve the visibility and accessibility of complex safety cases. The bow tie method is a purely qualitative technique, which could be successfully introduced (or similar methodologies) to the nuclear industry as an additional tool to improve the visibility and understanding of the safety case, and thus complement (not substitute) the more rigorous safety analysis techniques which are the norm in this industry. By making the diagrams readily accessible in the control room, the operators of nuclear facilities could further improve their understanding of the safety significance of their role in preventing major accidents and mitigating consequences. (authors)

  12. Methodology of safety operating range determination for the PWR plant

    International Nuclear Information System (INIS)

    Kostadinov, V.; Mavko, B.

    1991-01-01

    The results of NPP Krsko core thermal power design limits investigation, which set bounds to the maximum allowable fuel temperature during normal operation and incidents of moderate frequency, are presented. In addition, allowable reactor coolant temperatures limited by the pressure of the steam generator safety valves opening are calculated. The range of a PWR plant safe operation imposed by the thermal overpower, the steam generator safety valves opening and DNBR safety limits is determined. (author)

  13. 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

  14. Radwaste Disposal Safety Analysis

    International Nuclear Information System (INIS)

    Hwang, Yong Soo; Kang, C. H.; Lee, Y. M.; Lee, S. H.; Jeong, J. T.; Choi, J. W.; Park, S. W.; Lee, H. S.; Kim, J. H.; Jeong, M. S.

    2010-02-01

    For the purpose of evaluating annual individual doses from a potential repository disposing of radioactive wastes from the operation of the prospective advanced nuclear fuel cycle facilities in Korea, the new safety assessment approaches are developed such as PID methods. The existing KAERI FEP list was reviewed. Based on these new reference and alternative scenarios are developed along with a new code based on the Goldsim. The code based on the compartment theory can be applied to assess both normal and what if scenarios. In addition detailed studies on THRC coupling is studied. The oriental biosphere study ends with great success over the completion of code V and V with JAEA. The further development of quality assurance, in the form of the CYPRUS+ enables handy use of it for information management

  15. 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)

  16. 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)

  17. Nuclear methodology development for clinical analysis

    International Nuclear Information System (INIS)

    Oliveira, Laura Cristina de

    2003-01-01

    In the present work the viability of using the neutron activation analysis to perform urine and blood clinical analysis was checked. The aim of this study is to investigate the biological behavior of animals that has been fed with chow doped by natural uranium for a long period. Aiming at time and cost reduction, the absolute method was applied to determine element concentration on biological samples. The quantitative results of urine sediment using NAA were compared with the conventional clinical analysis and the results were compatible. This methodology was also used on bone and body organs such as liver and muscles to help the interpretation of possible anomalies. (author)

  18. Safety assessment of a borehole type disposal facility using the ISAM methodology

    International Nuclear Information System (INIS)

    Blerk, J.J. van; Yucel, V.; Kozak, M.W.; Moore, B.A.

    2002-01-01

    As part of the IAEA's Co-ordinated Research Project (CRP) on Improving Long-term of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ISAM), three example cases were developed. The aim was to test the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Borehole Test Case (BTC), related to a proposed future disposal option for disused sealed radioactive sources. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the BTC and provides some general conclusions that can be drawn from the findings of their work. (author)

  19. 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

  20. Safety analysis SFR 1. Long-term safety

    International Nuclear Information System (INIS)

    2008-12-01

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

  1. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the

  2. Indian reservation safety improvement program : a methodology and case study.

    Science.gov (United States)

    2015-11-01

    Improving roadway safety on Indian reservations requires a comprehensive approach. Limited : resources, lack of crash data, and few cross-jurisdictions coordination has made it difficult for : Native American communities to address their roadway safe...

  3. Development of mass and energy release analysis methodology

    International Nuclear Information System (INIS)

    Kim, Cheol Woo; Song, Jeung Hyo; Park, Seok Jeong; Kim, Tech Mo; Han, Kee Soo; Choi, Han Rim

    2009-01-01

    Recently, new approaches to the accident analysis using the realistic evaluation have been attempted. These new approaches provide more margins to the plant safety, design, operation and maintenance. KREM (KEPRI Realistic Evaluation Methodology) for a large break loss-of-coolant accident (LOCA) is performed using RELAP5/MOD3 computer code including realistic evaluation models. KOPEC has developed KIMERA (KOPEC Improved Mass and Energy Release Analysis methodology) based on the realistic evaluation to improve the analysis method for the mass and energy (M/E) release and to obtain the adequate margin. KIMERA uses a simplified single code system unlike conventional M/E release analysis methodologies. This simple code system reduces the computing efforts especially for LOCA analysis. The computer code systems of this methodology are RELAP5K/CONTEMPT4 (or RELAP5-ME) like KREM methodology which couples RELAP5/MOD3.1/K and CONTEMPT4/MOD5. The new methodology, KIMERA based on the same engine as KREM, adopted conservative approaches for the M/E release such as break spillage model, multiplier on heat transfer coefficient (HTC), and long-term cooling model. KIMERA is developed based on a LOCA and applied to a main steam line break (MSLB) and approved by Korea Government. KIMERA has an ability of one-through calculation of the various transient stages of LOCAs in a single code system and calculate the M/E release analysis during the long term cooling period with the containment pressure and temperature (P/T) response. The containment P/T analysis results are compared with those of the Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3 and 4) FSAR which is the OPR1000 (Optimized Power Reactor 1000) type nuclear power plant. The results of a large break LOCA and an MSLB are similar to those of FSAR for UCN 3 and 4. However, the containment pressure during the post-blowdown period of a large break LOCA has much lower second peak than the first peak. The resultant containment peak

  4. A Practical Risk Assessment Methodology for Safety-Critical Train Control Systems

    Science.gov (United States)

    2009-07-01

    This project proposes a Practical Risk Assessment Methodology (PRAM) for analyzing railroad accident data and assessing the risk and benefit of safety-critical train control systems. This report documents in simple steps the algorithms and data input...

  5. Integrated vehicle-based safety systems light-vehicle field operational test, methodology and results report.

    Science.gov (United States)

    2010-12-01

    "This document presents the methodology and results from the light-vehicle field operational test conducted as part of the Integrated Vehicle-Based Safety Systems program. These findings are the result of analyses performed by the University of Michi...

  6. Methodology for dimensional variation analysis of ITER integrated systems

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes, F. Javier, E-mail: FranciscoJavier.Fuentes@iter.org [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France); Trouvé, Vincent [Assystem Engineering & Operation Services, rue J-M Jacquard CS 60117, 84120 Pertuis (France); Cordier, Jean-Jacques; Reich, Jens [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France)

    2016-11-01

    Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

  7. Methodology for dimensional variation analysis of ITER integrated systems

    International Nuclear Information System (INIS)

    Fuentes, F. Javier; Trouvé, Vincent; Cordier, Jean-Jacques; Reich, Jens

    2016-01-01

    Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

  8. Methodology for Determination of the Upper Safety Limit for Criticality Calculations for Criticality Safety Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.H.; Keener, H.J.; DeClue, J.F.; Krass, A.W.

    2001-04-01

    This report considers the methods for determination of an upper safety limit, and incorporating uncertainty and margin into the safety limit, provides comparisons, and recommends a preferred method for determining the Upper Safety Limit (USL). A USL is developed for CSAS25 from SCALE4.4a. The USL is applicable for the CSAS25 control module from the SCALE 4.4a computer code system for use in evaluating nuclear criticality safety of enriched uranium systems. The benchmark calculation results used for this report are documented in Y/DD-896. The statistical evaluation is documented in CCG-380. The 27-group ENDF/B-IV, 44-group ENDF/B-V, and 238-group ENDF/B-V cross-section libraries were used. Numerical methods for applying margins are described, but the determination of appropriate correlating parameters and values for additional margin, applicable to a particular analysis, must be determined as part of a process analysis. As such, this document does not specify final upper subcritical limits as has been done in the past. No correlation between calculation results and neutron energy causing fission was found for the critical experiment results. Analysts using these results are responsible for exercising sound engineering judgment using strong technical arguments to develop ''a margin in k{sub eff} or other correlating parameter that is sufficiently large to ensure that conditions (calculated by this method to be subcritical by this margin) will actually be subcritical.'' Documentation of area of applicability and determination and justification of the appropriate margin in the analyst's evaluation, in conjunction with this report, will constitute the complete Validation Report in accordance with ANSI/ANS-8.1-1998, Section 4.3.6(4).

  9. Chemical plant innovative safety investments decision-support methodology.

    Science.gov (United States)

    Reniers, G L L; Audenaert, A

    2009-01-01

    This article examines the extent to which investing in safety during the creation of a new chemical installation proves profitable. The authors propose a management supporting cost-benefit model that identifies and evaluates investments in safety within a chemical company. This innovative model differentiates between serious accidents and less serious accidents, thus providing an authentic image of prevention-related costs and benefits. In classic cost-benefit analyses, which do not make such differentiations, only a rudimentary image of potential profitability resulting from investments in safety is obtained. The resulting management conclusions that can be drawn from such classical analyses are of a very limited nature. The proposed model, however, is applied to a real case study and the proposed investments in safety at an appointed chemical installation are weighed against the estimated hypothetical benefits resulting from the preventive measures to be installed at the installation. In the case-study carried out in question, it would appear that the proposed prevention investments are justified. Such an economic exercise may be very important to chemical corporations trying to (further) improve their safety investments.

  10. 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)

  11. Requirements Analysis in the Value Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Conner, Alison Marie

    2001-05-01

    The Value Methodology (VM) study brings together a multidisciplinary team of people who own the problem and have the expertise to identify and solve it. With the varied backgrounds and experiences the team brings to the study, come different perspectives on the problem and the requirements of the project. A requirements analysis step can be added to the Information and Function Analysis Phases of a VM study to validate whether the functions being performed are required, either regulatory or customer prescribed. This paper will provide insight to the level of rigor applied to a requirements analysis step and give some examples of tools and techniques utilized to ease the management of the requirements and functions those requirements support for highly complex problems.

  12. Cost analysis methodology of spent fuel storage

    International Nuclear Information System (INIS)

    1994-01-01

    The report deals with the cost analysis of interim spent fuel storage; however, it is not intended either to give a detailed cost analysis or to compare the costs of the different options. This report provides a methodology for calculating the costs of different options for interim storage of the spent fuel produced in the reactor cores. Different technical features and storage options (dry and wet, away from reactor and at reactor) are considered and the factors affecting all options defined. The major cost categories are analysed. Then the net present value of each option is calculated and the levelized cost determined. Finally, a sensitivity analysis is conducted taking into account the uncertainty in the different cost estimates. Examples of current storage practices in some countries are included in the Appendices, with description of the most relevant technical and economic aspects. 16 figs, 14 tabs

  13. RAMS (Risk Analysis - Modular System) methodology

    Energy Technology Data Exchange (ETDEWEB)

    Stenner, R.D.; Strenge, D.L.; Buck, J.W. [and others

    1996-10-01

    The Risk Analysis - Modular System (RAMS) was developed to serve as a broad scope risk analysis tool for the Risk Assessment of the Hanford Mission (RAHM) studies. The RAHM element provides risk analysis support for Hanford Strategic Analysis and Mission Planning activities. The RAHM also provides risk analysis support for the Hanford 10-Year Plan development activities. The RAMS tool draws from a collection of specifically designed databases and modular risk analysis methodologies and models. RAMS is a flexible modular system that can be focused on targeted risk analysis needs. It is specifically designed to address risks associated with overall strategy, technical alternative, and `what if` questions regarding the Hanford cleanup mission. RAMS is set up to address both near-term and long-term risk issues. Consistency is very important for any comparative risk analysis, and RAMS is designed to efficiently and consistently compare risks and produce risk reduction estimates. There is a wide range of output information that can be generated by RAMS. These outputs can be detailed by individual contaminants, waste forms, transport pathways, exposure scenarios, individuals, populations, etc. However, they can also be in rolled-up form to support high-level strategy decisions.

  14. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    International Nuclear Information System (INIS)

    E.N. Lindner

    2004-01-01

    The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

  15. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    E.N. Lindner

    2004-12-03

    The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

  16. 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

  17. 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.)

  18. 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

  19. Status of safety analysis reports

    International Nuclear Information System (INIS)

    Cserhati, A.

    1999-01-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  20. 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 ...

  1. 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

  2. Methodology for Evaluating Safety System Operability using Virtual Parameter Network

    International Nuclear Information System (INIS)

    Park, Sukyoung; Heo, Gyunyoung; Kim, Jung Taek; Kim, Tae Wan

    2014-01-01

    KAERI (Korea Atomic Energy Research Institute) and UTK (University of Tennessee Knoxville) are working on the I-NERI project to suggest complement of this problem. This research propose the methodology which provide the alternative signal in case of unable guaranteed reliability of some instrumentation with KAERI. Proposed methodology is assumed that several instrumentations are working normally under the power supply condition because we do not consider the instrumentation survivability itself. Thus, concept of the Virtual Parameter Network (VPN) is used to identify the associations between plant parameters. This paper is extended version of the paper which was submitted last KNS meeting by changing the methodology and adding the result of the case study. In previous research, we used Artificial Neural Network (ANN) inferential technique for estimation model but every time this model showed different estimate value due to random bias each time. Therefore Auto-Associative Kernel Regression (AAKR) model which have same number of inputs and outputs is used to estimate. Also the importance measures in the previous method depend on estimation model but importance measure of improved method independent on estimation model. Also importance index of previous method depended on estimation model but importance index of improved method is independent on estimation model. In this study, we proposed the methodology to identify the internal state of power plant when severe accident happens also it has been validated through case study. SBLOCA which has large contribution to severe accident is considered as initiating event and relationship amongst parameter has been identified. VPN has ability to identify that which parameter has to be observed and which parameter can be alternative to the missing parameter when some instruments are failed in severe accident. In this study we have identified through results that commonly number 2, 3, 4 parameter has high connectivity while

  3. EUROCONTROL-Systemic Occurrence Analysis Methodology (SOAM)-A 'Reason'-based organisational methodology for analysing incidents and accidents

    International Nuclear Information System (INIS)

    Licu, Tony; Cioran, Florin; Hayward, Brent; Lowe, Andrew

    2007-01-01

    The Safety Occurrence Analysis Methodology (SOAM) developed for EUROCONTROL is an accident investigation methodology based on the Reason Model of organisational accidents. The purpose of a SOAM is to broaden the focus of an investigation from human involvement issues, also known as 'active failures of operational personnel' under Reason's original model, to include analysis of the latent conditions deeper within the organisation that set the context for the event. Such an approach is consistent with the tenets of Just Culture in which people are encouraged to provide full and open information about how incidents occurred, and are not penalised for errors. A truly systemic approach is not simply a means of transferring responsibility for a safety occurrence from front-line employees to senior managers. A consistent philosophy must be applied, where the investigation process seeks to correct deficiencies wherever they may be found, without attempting to apportion blame or liability

  4. ASSESSMENT OF SEISMIC ANALYSIS METHODOLOGIES FOR DEEPLY EMBEDDED NPP STRUCTURES

    International Nuclear Information System (INIS)

    XU, J.; MILLER, C.; COSTANTINO, C.; HOFMAYER, C.; GRAVES, H. NRC.

    2005-01-01

    Several of the new generation nuclear power plant designs have structural configurations which are proposed to be deeply embedded. Since current seismic analysis methodologies have been applied to shallow embedded structures (e.g., ASCE 4 suggest that simple formulations may be used to model embedment effect when the depth of embedment is less than 30% of its foundation radius), the US Nuclear Regulatory Commission is sponsoring a program at the Brookhaven National Laboratory with the objective of investigating the extent to which procedures acceptable for shallow embedment depths are adequate for larger embedment depths. This paper presents the results of a study comparing the response spectra obtained from two of the more popular analysis methods for structural configurations varying from shallow embedment to complete embedment. A typical safety related structure embedded in a soil profile representative of a typical nuclear power plant site was utilized in the study and the depths of burial (DOB) considered range from 25-100% the height of the structure. Included in the paper are: (1) the description of a simplified analysis and a detailed approach for the SSI analyses of a structure with various DOB, (2) the comparison of the analysis results for the different DOBs between the two methods, and (3) the performance assessment of the analysis methodologies for SSI analyses of deeply embedded structures. The resulting assessment from this study has indicated that simplified methods may be capable of capturing the seismic response for much deeper embedded structures than would be normally allowed by the standard practice

  5. Positive lists of cosmetic ingredients: Analytical methodology for regulatory and safety controls – A review

    International Nuclear Information System (INIS)

    Lores, Marta; Llompart, Maria; Alvarez-Rivera, Gerardo; Guerra, Eugenia; Vila, Marlene; Celeiro, Maria; Lamas, J. Pablo; Garcia-Jares, Carmen

    2016-01-01

    Cosmetic products placed on the market and their ingredients, must be safe under reasonable conditions of use, in accordance to the current legislation. Therefore, regulated and allowed chemical substances must meet the regulatory criteria to be used as ingredients in cosmetics and personal care products, and adequate analytical methodology is needed to evaluate the degree of compliance. This article reviews the most recent methods (2005–2015) used for the extraction and the analytical determination of the ingredients included in the positive lists of the European Regulation of Cosmetic Products (EC 1223/2009): comprising colorants, preservatives and UV filters. It summarizes the analytical properties of the most relevant analytical methods along with the possibilities of fulfilment of the current regulatory issues. The cosmetic legislation is frequently being updated; consequently, the analytical methodology must be constantly revised and improved to meet safety requirements. The article highlights the most important advances in analytical methodology for cosmetics control, both in relation to the sample pretreatment and extraction and the different instrumental approaches developed to solve this challenge. Cosmetics are complex samples, and most of them require a sample pretreatment before analysis. In the last times, the research conducted covering this aspect, tended to the use of green extraction and microextraction techniques. Analytical methods were generally based on liquid chromatography with UV detection, and gas and liquid chromatographic techniques hyphenated with single or tandem mass spectrometry; but some interesting proposals based on electrophoresis have also been reported, together with some electroanalytical approaches. Regarding the number of ingredients considered for analytical control, single analyte methods have been proposed, although the most useful ones in the real life cosmetic analysis are the multianalyte approaches. - Highlights:

  6. Positive lists of cosmetic ingredients: Analytical methodology for regulatory and safety controls – A review

    Energy Technology Data Exchange (ETDEWEB)

    Lores, Marta, E-mail: marta.lores@usc.es; Llompart, Maria; Alvarez-Rivera, Gerardo; Guerra, Eugenia; Vila, Marlene; Celeiro, Maria; Lamas, J. Pablo; Garcia-Jares, Carmen

    2016-04-07

    Cosmetic products placed on the market and their ingredients, must be safe under reasonable conditions of use, in accordance to the current legislation. Therefore, regulated and allowed chemical substances must meet the regulatory criteria to be used as ingredients in cosmetics and personal care products, and adequate analytical methodology is needed to evaluate the degree of compliance. This article reviews the most recent methods (2005–2015) used for the extraction and the analytical determination of the ingredients included in the positive lists of the European Regulation of Cosmetic Products (EC 1223/2009): comprising colorants, preservatives and UV filters. It summarizes the analytical properties of the most relevant analytical methods along with the possibilities of fulfilment of the current regulatory issues. The cosmetic legislation is frequently being updated; consequently, the analytical methodology must be constantly revised and improved to meet safety requirements. The article highlights the most important advances in analytical methodology for cosmetics control, both in relation to the sample pretreatment and extraction and the different instrumental approaches developed to solve this challenge. Cosmetics are complex samples, and most of them require a sample pretreatment before analysis. In the last times, the research conducted covering this aspect, tended to the use of green extraction and microextraction techniques. Analytical methods were generally based on liquid chromatography with UV detection, and gas and liquid chromatographic techniques hyphenated with single or tandem mass spectrometry; but some interesting proposals based on electrophoresis have also been reported, together with some electroanalytical approaches. Regarding the number of ingredients considered for analytical control, single analyte methods have been proposed, although the most useful ones in the real life cosmetic analysis are the multianalyte approaches. - Highlights:

  7. Axiomatic safety-critical assessment process (ASCAP) simulation methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kaufman, L.M.; Giras, T.C. [Virginia Univ., Charlottesville, VA (United States). Center for Safety-Critical Ssytems

    2000-07-01

    The ASCAP simulation methodology models continuous train movement using a hybrid simulation consisting of both time and event driven portions. The actual train movement is predicated upon the solution of a series of discrete equations to determine its velocity, acceleration and braking profile at a specific time in tandem with the behavior of the train crew and the dispatcher agents in relationship to the behavioral states of the various stationary and mobile objects that a given train encounters. From this hybrid simulation, it is possible to generate accident scenarios. By performing multiple experiments to collect data for various accident scenarios, it will be possible to statistically quantify the risk associated with the various accident scenarios that are identified. (orig.)

  8. Reactivity parameters for safety analysis

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1985-01-01

    The reactor core model in the most commonly used computer programs for safety analysis is a point kinetics model. The core average fission rate is calculated knowing the reactivity, neutron generation time and delayed-neutron parameters. The reactivity is a time dependent function taking account of the effect of changes in water density and temperature, fuel temperature, control rod position and soluble boron concentration. In this presentation some of the alternative ways of representing this reactivity function are reviewed

  9. Methodologies for risk analysis in slope instability

    International Nuclear Information System (INIS)

    Bernabeu Garcia, M.; Diaz Torres, J. A.

    2014-01-01

    This paper is an approach to the different methodologies used in conducting landslide risk maps so that the reader can get a basic knowledge about how to proceed in its development. The landslide hazard maps are increasingly demanded by governments. This is because due to climate change, deforestation and the pressure exerted by the growth of urban centers, damage caused by natural phenomena is increasing each year, making this area of work a field of study with increasing importance. To explain the process of mapping a journey through each of the phases of which it is composed is made: from the study of the types of slope movements and the necessary management of geographic information systems (GIS) inventories and landslide susceptibility analysis, threat, vulnerability and risk. (Author)

  10. Studying creativity training programs: A methodological analysis

    DEFF Research Database (Denmark)

    Valgeirsdóttir, Dagný; Onarheim, Balder

    2017-01-01

    Throughout decades of creativity research, a range of creativity training programs have been developed, tested, and analyzed. In 2004 Scott and colleagues published a meta‐analysis of all creativity training programs to date, and the review presented here sat out to identify and analyze studies...... published since the seminal 2004 review. Focusing on quantitative studies of creativity training programs for adults, our systematic review resulted in 22 publications. All studies were analyzed, but comparing the reported effectiveness of training across studies proved difficult due to methodological...... inconsistencies, variations in reporting of results as well as types of measures used. Thus a consensus for future studies is called for to answer the question: Which elements make one creativity training program more effective than another? This is a question of equal relevance to academia and industry...

  11. Methodological considerations for improving Western blot analysis.

    Science.gov (United States)

    MacPhee, Daniel J

    2010-01-01

    The need for a technique that could allow the determination of antigen specificity of antisera led to the development of a method that allowed the production of a replica of proteins, which had been separated electrophoretically on polyacrylamide gels, on to a nitrocellulose membrane. This method was coined Western blotting and is very useful to study the presence, relative abundance, relative molecular mass, post-translational modification, and interaction of specific proteins. As a result it is utilized routinely in many fields of scientific research such as chemistry, biology and biomedical sciences. This review serves to touch on some of the methodological conditions that should be considered to improve Western blot analysis, particularly as a guide for graduate students but also scientists who wish to continue adapting this now fundamental research tool. Copyright 2009 Elsevier Inc. All rights reserved.

  12. Reload safety analysis automation tools

    International Nuclear Information System (INIS)

    Havlůj, F.; Hejzlar, J.; Vočka, R.

    2013-01-01

    Performing core physics calculations for the sake of reload safety analysis is a very demanding and time consuming process. This process generally begins with the preparation of libraries for the core physics code using a lattice code. The next step involves creating a very large set of calculations with the core physics code. Lastly, the results of the calculations must be interpreted, correctly applying uncertainties and checking whether applicable limits are satisfied. Such a procedure requires three specialized experts. One must understand the lattice code in order to correctly calculate and interpret its results. The next expert must have a good understanding of the physics code in order to create libraries from the lattice code results and to correctly define all the calculations involved. The third expert must have a deep knowledge of the power plant and the reload safety analysis procedure in order to verify, that all the necessary calculations were performed. Such a procedure involves many steps and is very time consuming. At ÚJV Řež, a.s., we have developed a set of tools which can be used to automate and simplify the whole process of performing reload safety analysis. Our application QUADRIGA automates lattice code calculations for library preparation. It removes user interaction with the lattice code and reduces his task to defining fuel pin types, enrichments, assembly maps and operational parameters all through a very nice and user-friendly GUI. The second part in reload safety analysis calculations is done by CycleKit, a code which is linked with our core physics code ANDREA. Through CycleKit large sets of calculations with complicated interdependencies can be performed using simple and convenient notation. CycleKit automates the interaction with ANDREA, organizes all the calculations, collects the results, performs limit verification and displays the output in clickable html format. Using this set of tools for reload safety analysis simplifies

  13. Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs

  14. Application of risk-based methodologies to prioritize safety resources

    International Nuclear Information System (INIS)

    Rahn, F.J.; Sursock, J.P.; Hosler, J.

    1993-01-01

    The Electric Power Research Institute (EPRI) started a program entitled risk-based prioritization in 1992. The purpose of this program is to provide generic technical support to the nuclear power industry relative to its recent initiatives in the area of operations and maintenance (O ampersand M) cost control using state-of-the-art risk methods. The approach uses probabilistic risk assessment (PRA), or similar techniques, to allocate resources commensurate with the risk posed by nuclear plant operations. Specifically, those items or events that have high risk significance would receive the most attention, while those with little risk content would command fewer resources. As quantified in a companion paper,close-quote the potential O ampersand M cost reduction inherent in this approach is very large. Furthermore, risk-based methods should also lead to safety improvements. This paper outlines the way that the EPRI technical work complements the technical, policy, and regulatory initiatives taken by others in the industry and provides an example of the approach as used to prioritize motor-operated valve (MOV) testing in response to US Nuclear Regulatory Commission (NRC) Generic Letter 89-10

  15. 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.)

  16. 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

  17. Periodic safety review of the HTR-10 safety analysis

    International Nuclear Information System (INIS)

    Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

    2015-01-01

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

  18. Application of the Integrated Safety Assessment Methodology to Sequences with Loss of Component Cooling Water System

    International Nuclear Information System (INIS)

    Queral, C.; Ibanez, L.; Hortal, J.; Izquierdo, J.M.; Sanchez-Perea, M.; Melendez, E.; Gil, J.; Fernandez, I.; Gomez, J.; Marrao, H.

    2013-01-01

    The Integrated Safety Assessment (ISA) methodology, developed by the Consejo de Seguridad Nuclear (CSN), Spanish Nuclear Regulatory Body, has been applied to a thermo-hydraulic analysis of Zion NPP for sequences with loss of the Component Cooling Water System (CCWS). The ISA methodology allows obtaining the damage domain (the region where the PCT limit is exceeded) for each sequence of the dynamic event tree as a function of the operator actuations times (secondary side cooling and recovery of CCWS) and the time of occurrence of stochastic phenomena (seal LOCA), and computing from it the exceedance frequency by integrating the dynamic reliability equations proposed by ISA. For every sequence, several data are necessary in order to obtain its contribution to the global exceedance damage frequency. These data consist of the results of the simulations performed with MAAP and TRACE codes that are inside of the damage domain and the time-density probability distributions of the manual actions and the time of seal LOCA occurrence. Reported results show an slight increment of the exceedance damage frequency for this kind of sequences in a power up-rate from 100% to 110%. (authors)

  19. Safety methodology implementation in the conceptual design phase of a fusion reactor

    International Nuclear Information System (INIS)

    Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

    2007-01-01

    The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

  20. Safety evaluation methodology of engineering barriers at repository for low and intermediate level radioactive waste

    International Nuclear Information System (INIS)

    Zarnic, R.; Bokan Bosiljkov, V.; Giacomelli, M.

    2007-01-01

    Analyses of the roles of cement-based barriers in radioactive waste isolation show that models used to estimate their characteristics during the lifetime of the repository must consider the alteration of material properties with time due to degradation processes. Reinforced concrete barriers at repositories shall be designed in such manner that they fulfil besides isolative capabilities also the required functions of mechanical resistance and stability. Key elements of safety evaluation are mainly the correct selection of materials for mineral composites with cement binder (cements, aggregates, mineral additives and chemical admixtures) and their design, execution of construction works and production of precast concrete containers (continuous casting of concrete - no cold joints, limited number of construction joints, proper placing and consolidation, finishing and curing), strict control of used materials and inspection of works, as well as investigation after the construction (visual inspection, non-destructive testing, monitoring, ageing assessment on test containers). According to the methodology presented in this paper the lifetime of the repository can be estimated and, if shorter than 300 years or shorter than the period resulting from safety analysis, appropriate corrective measures shall be taken. (author)

  1. Eliciting and communicating expert judgments: methodology and application to nuclear safety

    International Nuclear Information System (INIS)

    Winterfeldt, D. von; Commission of the European Communities, Ispra

    1989-01-01

    Expert judgment has always been used informally in the analysis of complex engineering problems. Increasingly, however, the use of expert judgment has been formalized by eliciting judgments in an explicit, documented and often quantitative way. In nuclear safety studies the need for formal elicitation of expert judgments arises because of the lack of data and experiences, the need to adapt model results to the specific circumstances of a plant, and the large uncertainties surrounding the events and quantities that characterize an accident sequence. The recognition of the need for a formal elicitation of expert judgments has led to one of the most extensive expert elicitation processes to date in the context of the NUREG 1150 study. About 30 safety issues were quantified using expert judgments about probabilities of various uncertain events and quantities, ranging from the failure of a check valve in the cooling system to the pressure built up due to hydrogen production to release fractions of various radionuclides. In total, some 1000 probability distributions were elicited from some 50 experts. This paper first motivates the use of formal expert elicitation in complex engineering studies and describes the methodology of formal expert elicitation. Subsequently, it describes the overall approach of NUREG 1150 and provides an example of the elicitation of the probability of a bypass failure in a pressurized water reactor. The paper ends by discussing some lessons learned, problems encountered and by providing some recommendations

  2. 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)

  3. Airline Safety: A Comparative Analysis.

    Science.gov (United States)

    1987-01-01

    S.TP OFR O T PEIDCV E Airline Safety: A Comparative Analysis TRlES IS1j0’~fJ 6. PERFORMING 01G. REPORT NUMBER AU TNOR( ) Sign . CONTRACT OR GRANT NUMBER...accidents. Perhaps because of an airline’s understandable sensitivity to public knowledge of its accidents, one has little assurance that each airline...62,169 0 Royal Air Maroc 81,451 0 80,861 0 (Morocco) Royal Nepal 11,885 0 19,785 0 SAA (South Africa) 57,226 0 61,618 0 SAHSA (Honduras) 32,658 0 34,894 0

  4. Training and Action for Patient Safety: Embedding Interprofessional Education for Patient Safety within an Improvement Methodology

    Science.gov (United States)

    Slater, Beverley L.; Lawton, Rebecca; Armitage, Gerry; Bibby, John; Wright, John

    2012-01-01

    Introduction: Despite an explosion of interest in improving safety and reducing error in health care, one important aspect of patient safety that has received little attention is a systematic approach to education and training for the whole health care workforce. This article describes an evaluation of an innovative multiprofessional, team-based…

  5. Use of PRA methodology for enhancing operational safety and reliability

    International Nuclear Information System (INIS)

    Chu, B.; Rumble, E.; Najafi, B.; Putney, B.; Young, J.

    1985-01-01

    This paper describes a broad scope, on-going R and D study, sponsored by the Electric Power Research Institute (EPRI) to utilize key features of the state-of-the-art plant information management and system analysis techniques to develop and demonstrate a practical engineering tool for assisting plant engineering and operational staff to perform their activities more effectively. The study is foreseen to consist of two major activities: to develop a user-friendly, integrated software system; and to demonstrate the applications of this software on-site. This integrated software, Reliability Analysis Program with In-Plant Data (RAPID), will consist of three types of interrelated elements: an Executive Controller which will provide engineering and operations staff users with interface and control of the other two software elements, a Data Base Manager which can acquire, store, select, and transfer data, and Applications Modules which will perform the specific reliability-oriented functions. A broad range of these functions has been envisaged. The immediate emphasis will be focused on four application modules: a Plant Status Module, a Technical Specification Optimization Module, a Reliability Assessment Module, and a Utility Module for acquiring plant data

  6. Risk analysis methodology designed for small and medium enterprises

    OpenAIRE

    Ladislav Beránek; Radim Remeš

    2009-01-01

    The aim of this paper is to present risk analysis procedures successfully applied by several Czech small and medium enterprises. The paper presents in detail the individual steps we use in risk analysis of small and medium enterprises in the Czech Republic. Suggested method to risk analysis is based on the modification of the FRAP methodology and the BITS recommendation. Modifications of both methodologies are described in detail. We propose modified risk analysis methodology which is quick a...

  7. Functional Hazard Analysis for Railway Safety

    OpenAIRE

    RAFRAFI, M; EL-KOURSI, EM

    2007-01-01

    The apportionment of railway safety targets is a key issue to develop a common safety management in the European railway system. In this paper, we develop a generic approach based on the Functional Hazard Analysis (FHA), to analyse the safety of railway systems for a unified European network and to comply with the Common Safety Targets (CSTs) required by the European railway safety directive. We suggest to combine the FHA technique with the functional railway architecture, developed by the AE...

  8. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

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

    2005-03-01

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

  9. Probabilistic safety analysis level 2

    International Nuclear Information System (INIS)

    Lantaron, J.A.

    1993-01-01

    In 1989 the Spanish Council of Nuclear Safety selected the Nuclear Power Plant Jose Cabrera to perform the Probabilistic Safety Analysis (PSA) within the National Integrated Program. In this case the level 2 was required which adds to the level 1 all the analysis of processes involved in the accident and their effect in the ''isolation response''. This study was followed in two new Nuclear Power plants (Vandellos ii and Trillo). The objectives of these probabilistic analyses are, from one side, to develop a global assessment of the severe accident behaviour, to understand the most probable severe accident sequences and to quantify, as much as possible, the probability of core global damage and radionuclides release to the environment, and on the other hand, if necessary, to diminish the global probability obtained by modifying procedures, components and systems, to help prevention and mitigation of severe accidents. This study will allow to evaluate operator actions or equipment improvements and will inform our Institution for new risk analyses (a PSA of level 3). (Author)

  10. Preliminary safety analysis of the Gorleben site

    International Nuclear Information System (INIS)

    Bracke, G.; Fischer-Appelt, K.

    2014-01-01

    The safety requirements governing the final disposal of heat-generating radioactive waste in Germany were implemented by the Federal Ministry of Environment, Natural Conservation and Nuclear Safety (BMU) in 2010. The Ministry considers as a fundamental objective the protection of man and the environment against the hazards of radioactive waste. Unreasonable burdens and obligation for future generations shall be avoided. The main safety principles are concentration and inclusion of radioactive and other pollutants in a containment-providing rock zone. Any release of radioactive nuclides may increase the risk for men and the environment only negligibly compared to natural radiation exposure. No intervention or maintenance work shall be necessary in the post-closure phase. Retrieval/recovery of the waste shall be possible up to 500 years after closure. The Gorleben salt dome has been discussed since the 1970's as a possible repository site for heat-generating radioactive waste in Germany. The objective of the project preliminary safety analysis of the Gorleben site (VSG) was to assess if repository concepts at the Gorleben site or other sites with a comparable geology could comply with these requirements based on currently available knowledge (Fischer-Appelt, 2013; Bracke, 2013). In addition to this it was assessed if methodological approaches can be used for a future site selection procedure and which technological and conceptual considerations can be transferred to other geological situations. The objective included the compilation and review of the available exploration data of the Gorleben site and on disposal in salt rock, the development of repository designs, and the identification of the needs for future R and D work and further site investigations. (authors)

  11. Motorcoach and school bus fire safety analysis.

    Science.gov (United States)

    2016-11-01

    This report documents a motorcoach and school bus fire safety analysis performed by the John A. Volpe National Transportation Systems Center (Volpe) for the Federal Motor Carrier Safety Administration. This report aims to: 1) identify the causes, fre...

  12. Radiological safety methodology in radioactive tracer applications for hydrodynamics and environmental studies

    International Nuclear Information System (INIS)

    Suarez, R.; Badano, A.; Dellepere, A.; Artucio, G.; Bertolotti, A.

    1995-01-01

    The use of radioactive tracer techniques as control sewage disposal contamination in Montevideo Estuarine and Carrasco beach has been studied for the Nuclear Technology National Direction. Hydrodynamic models simulation has been introduced as work methodology. As well as radiological safety and radioactive material applications in the environmental studies has been evaluated mainly in the conclusions and recommendations in this report. maps

  13. 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)

  14. Clean Energy Manufacturing Analysis Center Benchmark Report: Framework and Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Sandor, Debra [National Renewable Energy Lab. (NREL), Golden, CO (United States); Chung, Donald [National Renewable Energy Lab. (NREL), Golden, CO (United States); Keyser, David [National Renewable Energy Lab. (NREL), Golden, CO (United States); Mann, Margaret [National Renewable Energy Lab. (NREL), Golden, CO (United States); Engel-Cox, Jill [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-05-23

    This report documents the CEMAC methodologies for developing and reporting annual global clean energy manufacturing benchmarks. The report reviews previously published manufacturing benchmark reports and foundational data, establishes a framework for benchmarking clean energy technologies, describes the CEMAC benchmark analysis methodologies, and describes the application of the methodologies to the manufacturing of four specific clean energy technologies.

  15. A Review of Citation Analysis Methodologies for Collection Management

    Science.gov (United States)

    Hoffmann, Kristin; Doucette, Lise

    2012-01-01

    While there is a considerable body of literature that presents the results of citation analysis studies, most researchers do not provide enough detail in their methodology to reproduce the study, nor do they provide rationale for methodological decisions. In this paper, we review the methodologies used in 34 recent articles that present a…

  16. Safety disconnect: Analysis of the role of labor experience and safety training on work safety perceptions

    Directory of Open Access Journals (Sweden)

    Esteban Lafuente

    2018-02-01

    Originality/value: Work safety constitutes a relevant key performance indicator. The proposed analysis of the role of labor experience and safety training on perceived work safety in different types of employees contributes to better understand how organizations can improve the management of their workforce by triggering specific actions—such as the design of customized training programs—that may help in reducing the safety disconnect between employees, in terms of perceived work safety.

  17. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Gensler, A.; Kühnel, K.; Kuch, S.

    2015-01-01

    Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

  18. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

    Saito, G.H.

    1994-01-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment

  19. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

  20. Methodology to analysis of aging processes of containment spray system

    International Nuclear Information System (INIS)

    Borges, D. da Silva; Lava, D.D.; Moreira, M. de L.; Ferreira Guimarães, A.C.; Fernandes da Silva, L.

    2015-01-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The motivation for write this work emerged from the current perspective nuclear. Numerous nuclear power plants worldwide have an advanced operating time. This problem requires a process to ensure the confiability of the operative systems of these plants, because of this, it is necessary a methodologies capable of estimate the failure probability of the components and systems. In addition to the safety factors involved, such methodologies can to be used to search ways to ensure the extension of the life cycle of nuclear plants, which inevitably will pass by the decommissioning process after the operating time of 40 years. This process negatively affects the power generation, besides demanding an enormous investment for such. Thus, this paper aims to present modeling techniques and sensitivity analysis, which together can generate an estimate of how components, which are more sensitive to the aging process, will behave during the normal operation cycle of a nuclear power plant. (authors)

  1. Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Chung, D.Y.

    1999-01-01

    Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45

  2. Uncertainty and sensitivity analysis in a Probabilistic Safety Analysis level-1

    International Nuclear Information System (INIS)

    Nunez Mc Leod, Jorge E.; Rivera, Selva S.

    1996-01-01

    A methodology for sensitivity and uncertainty analysis, applicable to a Probabilistic Safety Assessment Level I has been presented. 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 systems response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well as different graphical visualization for the control of the study. (author)

  3. A study of software safety analysis system for safety-critical software

    International Nuclear Information System (INIS)

    Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

    2004-01-01

    The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases

  4. Automation for System Safety Analysis

    Science.gov (United States)

    Malin, Jane T.; Fleming, Land; Throop, David; Thronesbery, Carroll; Flores, Joshua; Bennett, Ted; Wennberg, Paul

    2009-01-01

    This presentation describes work to integrate a set of tools to support early model-based analysis of failures and hazards due to system-software interactions. The tools perform and assist analysts in the following tasks: 1) extract model parts from text for architecture and safety/hazard models; 2) combine the parts with library information to develop the models for visualization and analysis; 3) perform graph analysis and simulation to identify and evaluate possible paths from hazard sources to vulnerable entities and functions, in nominal and anomalous system-software configurations and scenarios; and 4) identify resulting candidate scenarios for software integration testing. There has been significant technical progress in model extraction from Orion program text sources, architecture model derivation (components and connections) and documentation of extraction sources. Models have been derived from Internal Interface Requirements Documents (IIRDs) and FMEA documents. Linguistic text processing is used to extract model parts and relationships, and the Aerospace Ontology also aids automated model development from the extracted information. Visualizations of these models assist analysts in requirements overview and in checking consistency and completeness.

  5. Development of Flight Safety Prediction Methodology for U. S. Naval Safety Center. Revision 1

    Science.gov (United States)

    1970-02-01

    POWER DISTRIBUTION K Fbl FAH FBA FAAAAAAAAA K FOG FBB FAAAAAAAAA K FRG F8C FAkAAAAAAA K FOG FBO FAAAAAAAAA K FOG FBE FAAAAAAAAA K FBO FoF FAAAAAAAAA FOG F...safety-prediction math model. *Department of the Navy, Naval Aviation Maintenance anid Material Mangement Manual, Publication 0618-200-0100. iE-15 TABLE

  6. 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)

  7. Manpower analysis in transportation safety. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, C.S.; Bowden, H.M.; Colford, C.A.; DeFilipps, P.J.; Dennis, J.D.; Ehlert, A.K.; Popkin, H.A.; Schrader, G.F.; Smith, Q.N.

    1977-05-01

    The project described provides a manpower review of national, state and local needs for safety skills, and projects future manning levels for transportation safety personnel in both the public and private sectors. Survey information revealed that there are currently approximately 121,000 persons employed directly in transportation safety occupations within the air carrier, highway and traffic safety, motor carrier, pipeline, rail carrier, and marine carrier transportation industry groups. The projected need for 1980 is over 145,000 of which over 80 percent will be in highway safety. An analysis of transportation tasks is included, and shows ten general categories about which the majority of safety activities are focused. A skills analysis shows a generally high level of educational background and several years of experience are required for most transportation safety jobs. An overall review of safety programs in the transportation industry is included, together with chapters on the individual transportation modes.

  8. 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)

  9. 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)

  10. 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

  11. External Events Analysis for LWRS/RISMC Project: Methodology Development and Early Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Parisi, Carlo [Idaho National Laboratory; Prescott, Steven Ralph [Idaho National Laboratory; Yorg, Richard Alan [Idaho National Laboratory; Coleman, Justin Leigh [Idaho National Laboratory; Szilard, Ronaldo Henriques [Idaho National Laboratory

    2016-02-01

    The ultimate scope of Industrial Application #2 (IA) of the LWRS/RISMC project is a realistic simulation of natural external hazards that impose threat to a NPP. This scope requires the development of a methodology and of a qualified set of tools able to perform advanced risk- informed safety analysis. In particular the methodology should be able to combine results from seismic, flooding and thermal-hydraulic (TH) deterministic calculations with dynamic PRA. This summary presents the key points of methodology being developed and the very first sample application of it to a simple problem (spent fuel pool).

  12. A Goal based methodology for HAZOP analysis

    DEFF Research Database (Denmark)

    Rossing, Netta Liin; Lind, Morten; Jensen, Niels

    2010-01-01

    This paper presents a goal based methodology for HAZOP studies in which a functional model of the plant is used to assist in a functional decomposition of the plant starting from the purpose of the plant and continuing down to the function of a single node, e.g. a pipe section. This approach lead...

  13. 14 CFR 35.15 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety analysis. 35.15 Section 35.15... STANDARDS: PROPELLERS Design and Construction § 35.15 Safety analysis. (a)(1) The applicant must analyze the.... This analysis will take into account, if applicable: (i) The propeller system in a typical installation...

  14. Introduction and summary of the 13th meeting of the Scientific Group on Methodologies for the Safety Evaluation of Chemicals (SGOMSEC): alternative testing methodologies.

    OpenAIRE

    Stokes, W S; Marafante, E

    1998-01-01

    A workshop on alternative toxicological testing methodologies was convened by the Scientific Group on Methodologies for the Safety Evaluation of Chemicals (SGOMSEC) 26-31 January 1997 in Ispra, Italy, at the European Centre for the Validation of Alternative Methods. The purpose of the workshop was to assess the current status of alternative testing methodologies available to evaluate adverse human health and environmental effects of chemicals. Another objective of the workshop was to identify...

  15. 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.

  16. Using a Realist Research Methodology in Policy Analysis

    Science.gov (United States)

    Lourie, Megan; Rata, Elizabeth

    2017-01-01

    The article describes the usefulness of a realist methodology in linking sociological theory to empirically obtained data through the development of a methodological device. Three layers of analysis were integrated: 1. the findings from a case study about Maori language education in New Zealand; 2. the identification and analysis of contradictions…

  17. 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

  18. The status of development and practical use of probabilistic safety assessment methodology in CSSR

    International Nuclear Information System (INIS)

    Hrehor, M.

    1987-01-01

    The first part of the paper gives a brief summary of the current status of the nuclear energy programme and its regulatory background in CSSR emphasizing a leading role of the State Nuclear Safety Inspectorate of CsAEC in the development and practical use of probabilistic safety assessment methodology. In the second part a simple practical technique is presented enabling calculation of MTBF and MTTR in the cases which cannot be directly modelled by means of logical operators commonly used in fault tree models. (orig.)

  19. 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.

  20. AST-500 safety analysis experience

    International Nuclear Information System (INIS)

    Falikov, A.A.; Bakhmetiev, A.M.; Kuul, V.S.; Samoilov, O.B.

    1997-01-01

    Characteristic AST-type NHR safety features and requirements are described briefly. The main approaches and results of design and beyond-design accidents analyses for the AST-500 NHR, and the results of probabilistic safety assessments are considered. It is concluded that the AST-500 possesses a high safety level in virtue of the development and realization in the design of self-protection, passivity and defence-in-depth principles. (author). 9 refs, 2 figs

  1. Lunar mission safety and rescue: Hazards analysis and safety requirements

    Science.gov (United States)

    1971-01-01

    The results are presented of the hazards analysis which was concerned only with hazards to personnel and not with loss of equipment or property. Hazards characterization includes the definition of a hazard, the hazard levels, and the hazard groups. The analysis methodology is described in detail. The methodology was used to prepare the top level functional flow diagrams, to perform the first level hazards assessment, and to develop a list of conditions and situations requiring individual hazard studies. The 39 individual hazard study results are presented in total.

  2. 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.

  3. Methodology for the Integration of Safety in the Optimization of the Advanced Reactors Design

    International Nuclear Information System (INIS)

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

    2003-01-01

    In this work a new methodology has been developed and implemented for taking into account the safety levels of the reactor in a design optimization process, by using Design Maps.They represent a new technique for comparing critical variables in case an accidental sequenced happened, with limit values set by the design criteria.So a good balance is achieved, without allowing the economic performance search to cause a too risky reactor, and guaranteeing the competitiveness of it in spite of the safety costs.Up to the moment, there is no design tool able to accomplish this task in an integrated way.A computational tool based on this methodology has been implemented.These tool specially programmed routines allow carrying out the mentioned tasks

  4. Composite Structures Damage Tolerance Analysis Methodologies

    Science.gov (United States)

    Chang, James B.; Goyal, Vinay K.; Klug, John C.; Rome, Jacob I.

    2012-01-01

    This report presents the results of a literature review as part of the development of composite hardware fracture control guidelines funded by NASA Engineering and Safety Center (NESC) under contract NNL04AA09B. The objectives of the overall development tasks are to provide a broad information and database to the designers, analysts, and testing personnel who are engaged in space flight hardware production.

  5. 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)

  6. 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

  7. 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

  8. Application of project management methodology in design management of nuclear safety related structure

    International Nuclear Information System (INIS)

    Chen Mao

    2004-01-01

    This paper focuses on the application of project management methodology in the design management of Nuclear Safety Related Structure (NSRS), considering the design management features of its civil construction. Based on the experiences from the management of several projects, the project management triangle is proposed to be used in the management, to well treat the position of design interface in the project management. Some other management methods are also proposed

  9. 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)

  10. Methodological frontier in operational analysis for roundabouts: a review

    Directory of Open Access Journals (Sweden)

    Orazio Giuffre'

    2016-11-01

    Full Text Available Several studies and researches have shown that modern roundabouts are safe and effective as engineering countermeasures for traffic calming, and they are now widely used worldwide. The increasing use of roundabouts and, more recently, turbo and flower roundabouts, has induced a great variety of experiences in the field of intersection design, traffic safety and capacity modelling. As for unsignalized intersections which represent the starting point to extend knowledge about the operational analysis to roundabouts, the general situation in capacity estimation is still characterized by the discussion between gap acceptance models and empirical regression models. However, capacity modelling must contain both the analytical construction and then solution of the model, and the implementation of driver behavior. Thus, issues on a realistic modelling of driver behavior by the parameters that are included into the models are always of interest for practioners and analysts in transportation and road infrastructure engineering. Based on these considerations, this paper presents a literature review about the key methodological issues in the operational analysis of modern roundabouts. Focus is made on the aspects associated with the gap acceptance behavior, the derivation of the analytical-based models and the calculation of parameters included into the capacity equations, as well as steady state and non-steady state conditions and uncertainty in entry capacity estimation. At last, insights on future developments of the research in this field of investigation will be also outlined.

  11. Methodology for risk analysis of nuclear installations

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Senne Junior, Murillo; Jordao, Elizabete

    2002-01-01

    Both the licensing standards for general uses in nuclear facilities and the specific ones require a risk assessment during their licensing processes. The risk assessment is carried out through the estimation of both probability of the occurrence of the accident, and their magnitudes. This is a complex task because the great deal of potential hazardous events that can occur in nuclear facilities difficult the statement of the accident scenarios. There are also many available techniques to identify the potential accidents, estimate their probabilities, and evaluate their magnitudes. In this paper is presented a new methodology that systematizes the risk assessment process, and orders the accomplishment of their several steps. (author)

  12. Hot Cell Facility (HCF) Safety Analysis Report

    International Nuclear Information System (INIS)

    MITCHELL, GERRY W.; LONGLEY, SUSAN W.; PHILBIN, JEFFREY S.; MAHN, JEFFREY A.; BERRY, DONALD T.; SCHWERS, NORMAN F.; VANDERBEEK, THOMAS E.; NAEGELI, ROBERT E.

    2000-01-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR

  13. Hot Cell Facility (HCF) Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

    2000-11-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

  14. Application of a methodology to determine priorities for nuclear power plant safety issues

    International Nuclear Information System (INIS)

    Daling, P.M.

    1988-01-01

    The Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) is sponsoring a research program to determine priorities of nuclear power plant safety issues. A methodology has been developed at the Pacific Northwest Laboratory (PNL) to provide technical assistance in the development of risk and cost estimates for implementing resolutions to the safety issues. The information development methods are intended to provide the NRC with a consistent level of information for use in ranking the issues. The NRC uses this information, along with judgmental factors, to rank the issues for further consideration by the NRC staff. The primary purpose of the priority rankings are to assist in the allocation of resources to issues that have high potential for reducing public risk as well as to remove issues from further consideration that have little safety significance

  15. Development methodology for the software life cycle process of the safety software

    International Nuclear Information System (INIS)

    Kim, D. H.; Lee, S. S.; Cha, K. H.; Lee, C. S.; Kwon, K. C.; Han, H. B.

    2002-01-01

    A methodology for developing software life cycle processes (SLCP) is proposed to develop the digital safety-critical Engineered Safety Features - Component Control System (ESF-CCS) successfully. A software life cycle model is selected as the hybrid model mixed with waterfall, prototyping, and spiral models and is composed of two stages , development stages of prototype of ESF-CCS and ESF-CCS. To produce the software life cycle (SLC) for the Development of the Digital Reactor Safety System, the Activities referenced in IEEE Std. 1074-1997 are mapped onto the hybrid model. The SLCP is established after the available OPAs (Organizational Process Asset) are applied to the SLC Activities, and the known constraints are reconciled. The established SLCP describes well the software life cycle activities with which the Regulatory Authority provides

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

    International Nuclear Information System (INIS)

    Batandjieva, B.; Metcalf, P.

    2003-01-01

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

  17. Development of economic consequence methodology for process risk analysis.

    Science.gov (United States)

    Zadakbar, Omid; Khan, Faisal; Imtiaz, Syed

    2015-04-01

    A comprehensive methodology for economic consequence analysis with appropriate models for risk analysis of process systems is proposed. This methodology uses loss functions to relate process deviations in a given scenario to economic losses. It consists of four steps: definition of a scenario, identification of losses, quantification of losses, and integration of losses. In this methodology, the process deviations that contribute to a given accident scenario are identified and mapped to assess potential consequences. Losses are assessed with an appropriate loss function (revised Taguchi, modified inverted normal) for each type of loss. The total loss is quantified by integrating different loss functions. The proposed methodology has been examined on two industrial case studies. Implementation of this new economic consequence methodology in quantitative risk assessment will provide better understanding and quantification of risk. This will improve design, decision making, and risk management strategies. © 2014 Society for Risk Analysis.

  18. 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.)

  19. Risk analysis and safety rationale

    International Nuclear Information System (INIS)

    Bengtsson, G.

    1989-01-01

    Decision making with respect to safety is becoming more and more complex. The risk involved must be taken into account together with numerous other factors such as the benefits, the uncertainties and the public perception. Can the decision maker be aided by some kind of system, general rules of thumb, or broader perspective on similar decisions? This question has been addressed in a joint Nordic project relating to nuclear power. Modern techniques for risk assessment and management have been studied, and parallels drawn to such areas as offshore safety and management of toxic chemicals in the environment. The report summarises the finding of 5 major technical reports which have been published in the NORD-series. The topics includes developments, uncertainties and limitations in probabilistic safety assessments, negligible risks, risk-cost trade-offs, optimisation of nuclear safety and radiation protection, and the role of risks in the decision making process. (author) 84 refs

  20. Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oztunali, O.I.; Roles, G.W.

    1986-01-01

    Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of additional alternative disposal technologies, (3) expansion of the methodology used to calculate disposal costs, (4) consideration of an additional exposure pathway involving direct human contact with disposed waste due to a hypothetical drilling scenario, and (5) use of updated health physics analysis procedures (ICRP-30). Volume 1 of this report describes the calculational algorithms of the updated analysis methodology.

  1. Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

    International Nuclear Information System (INIS)

    Oztunali, O.I.; Roles, G.W.

    1986-01-01

    Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of additional alternative disposal technologies, (3) expansion of the methodology used to calculate disposal costs, (4) consideration of an additional exposure pathway involving direct human contact with disposed waste due to a hypothetical drilling scenario, and (5) use of updated health physics analysis procedures (ICRP-30). Volume 1 of this report describes the calculational algorithms of the updated analysis methodology

  2. Using Qualitative Hazard Analysis to Guide Quantitative Safety Analysis

    Science.gov (United States)

    Shortle, J. F.; Allocco, M.

    2005-01-01

    Quantitative methods can be beneficial in many types of safety investigations. However, there are many difficulties in using quantitative m ethods. Far example, there may be little relevant data available. This paper proposes a framework for using quantitative hazard analysis to prioritize hazard scenarios most suitable for quantitative mziysis. The framework first categorizes hazard scenarios by severity and likelihood. We then propose another metric "modeling difficulty" that desc ribes the complexity in modeling a given hazard scenario quantitatively. The combined metrics of severity, likelihood, and modeling difficu lty help to prioritize hazard scenarios for which quantitative analys is should be applied. We have applied this methodology to proposed concepts of operations for reduced wake separation for airplane operatio ns at closely spaced parallel runways.

  3. Shuttle TPS thermal performance and analysis methodology

    Science.gov (United States)

    Neuenschwander, W. E.; Mcbride, D. U.; Armour, G. A.

    1983-01-01

    Thermal performance of the thermal protection system was approximately as predicted. The only extensive anomalies were filler bar scorching and over-predictions in the high Delta p gap heating regions of the orbiter. A technique to predict filler bar scorching has been developed that can aid in defining a solution. Improvement in high Delta p gap heating methodology is still under study. Minor anomalies were also examined for improvements in modeling techniques and prediction capabilities. These include improved definition of low Delta p gap heating, an analytical model for inner mode line convection heat transfer, better modeling of structure, and inclusion of sneak heating. The limited number of problems related to penetration items that presented themselves during orbital flight tests were resolved expeditiously, and designs were changed and proved successful within the time frame of that program.

  4. METHODOLOGICAL ANALYSIS OF TRAINING STUDENT basketball teams

    Directory of Open Access Journals (Sweden)

    Kozina Zh.L.

    2011-06-01

    Full Text Available Considered the leading position of the preparation of basketball teams in high schools. The system includes the following: reliance on top-quality players in the structure of preparedness, widespread use of visual aids, teaching movies and cartoons with a record of technology implementation of various methods by professional basketball players, the application of the methods of autogenic and ideomotor training according to our methodology. The study involved 63 students 1.5 courses from various universities of Kharkov 1.2 digits: 32 experimental group and 31 - control. The developed system of training students, basketball players used within 1 year. The efficiency of the developed system in the training process of students, basketball players.

  5. Notions and methodologies for uncertainty analysis in simulations of transitory events of a nuclear central

    International Nuclear Information System (INIS)

    Alva N, J.; Ortiz V, J.; Amador G, R.; Delfin L, A.

    2007-01-01

    The present work has as objective to gather the basic notions related with the uncertainty analysis and some of the methodologies to be applied in the studies of transitory events analysis of a nuclear power station, in particular of those thermal hydraulics phenomena. The concepts and methodologies mentioned in this work are the result of an exhaustive bibliographical investigation of the topic in the nuclear area. The methodologies of uncertainties analysis have been developed by diverse institutions and they are broadly used at world level for their application in the results of the computer codes of the class of better estimation in the thermal hydraulics analysis and safety of plants and nuclear reactors. The main sources of uncertainty, types of uncertainty and aspects related with the models of better estimation and better estimation method are also presented. (Author)

  6. 10 CFR 70.62 - Safety program and integrated safety analysis.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Safety program and integrated safety analysis. 70.62... Nuclear Material § 70.62 Safety program and integrated safety analysis. (a) Safety program. (1) Each licensee or applicant shall establish and maintain a safety program that demonstrates compliance with the...

  7. An economic analysis methodology for project evaluation and programming.

    Science.gov (United States)

    2013-08-01

    Economic analysis is a critical component of a comprehensive project or program evaluation methodology that considers all key : quantitative and qualitative impacts of highway investments. It allows highway agencies to identify, quantify, and value t...

  8. Vulnerability and Risk Analysis Program: Overview of Assessment Methodology

    National Research Council Canada - National Science Library

    2001-01-01

    .... Over the last three years, a team of national laboratory experts, working in partnership with the energy industry, has successfully applied the methodology as part of OCIP's Vulnerability and Risk Analysis Program (VRAP...

  9. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

  10. [Free will and neurobiology: a methodological analysis].

    Science.gov (United States)

    Brücher, K; Gonther, U

    2006-04-01

    Whether or not the neurobiological basis of mental processes is compatible with the philosophical postulate of free will is a matter of committed debating in our days. What is the meaning of those frequently-quoted experiments concerning voluntary action? Both convictions, being autonomous subjects and exercising a strong influence on the world by applying sciences, have become most important for modern human self-conception. Now these two views are growing apart and appear contradictory because neurobiology tries to reveal the illusionary character of free will. In order to cope with this ostensible dichotomy it is recommended to return to the core of scientific thinking, i. e. to the reflection about truth and methods. The neurobiological standpoint referring to Libet as well as the philosophical approaches to free will must be analysed, considering pre-conceptions and context-conditions. Hence Libet's experiments can be criticised on different levels: methods, methodology and epistemology. Free will is a highly complex system, not a simple fact. Taking these very complicated details into account it is possible to define conditions of compatibility and to use the term free will still in a meaningful way, negotiating the obstacles called pure chance and determinism.

  11. Autoclave nuclear criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

    1991-12-31

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

  12. 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)

  13. Severe accident analysis methodology in support of accident management

    International Nuclear Information System (INIS)

    Boesmans, B.; Auglaire, M.; Snoeck, J.

    1997-01-01

    The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

  14. Social Networks Analysis: Classification, Evaluation, and Methodologies

    Science.gov (United States)

    2011-02-28

    and time performance. We also focus on large-scale network size and dynamic changes in networks and research new capabilities in performing social networks analysis utilizing parallel and distributed processing.

  15. Radiochemical Analysis Methodology for uranium Depletion Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Scatena-Wachel DE

    2007-01-09

    This report provides sufficient material for a test sponsor with little or no radiochemistry background to understand and follow physics irradiation test program execution. Most irradiation test programs employ similar techniques and the general details provided here can be applied to the analysis of other irradiated sample types. Aspects of program management directly affecting analysis quality are also provided. This report is not an in-depth treatise on the vast field of radiochemical analysis techniques and related topics such as quality control. Instrumental technology is a very fast growing field and dramatic improvements are made each year, thus the instrumentation described in this report is no longer cutting edge technology. Much of the background material is still applicable and useful for the analysis of older experiments and also for subcontractors who still retain the older instrumentation.

  16. Diversion Path Analysis Handbook. Volume 1. Methodology

    International Nuclear Information System (INIS)

    Goodwin, K.E.; Schleter, J.C.; Maltese, M.D.K.

    1978-11-01

    Diversion Path Analysis (DPA) is a safeguards evaluation tool which is used to determine the vulnerability of the Material Control and Material Accounting (MC and MA) Subsystems to the threat of theft of Special Nuclear Material (SNM) by a knowledgeable Insider. The DPA team should consist of two individuals who have technical backgrounds. The implementation of DPA is divided into five basic steps: Information and Data Gathering, Process Characterization, Analysis of Diversion Paths, Results and Findings, and Documentation

  17. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  18. Safety analysis in subsurface repositories

    International Nuclear Information System (INIS)

    1985-06-01

    The development of mathematical models to represent the repository-geosphere-biosphere system, and the development of a structure for data acquisition, processing, and use to analyse the safety of subsurface repositories, are presented. To study the behavior of radionuclides in geosphere a laboratory to determine the hydrodynamic dispersion coefficient was constructed. (M.C.K.) [pt

  19. Safety analysis of disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Vieno, T.

    1994-04-01

    The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

  20. Discourse analysis: making complex methodology simple

    NARCIS (Netherlands)

    Bondarouk, Tatiana; Ruel, Hubertus Johannes Maria; Leino, T.; Saarinen, T.; Klein, S.

    2004-01-01

    Discursive-based analysis of organizations is not new in the field of interpretive social studies. Since not long ago have information systems (IS) studies also shown a keen interest in discourse (Wynn et al, 2002). The IS field has grown significantly in its multiplicity that is echoed in the

  1. Scenario aggregation and analysis via Mean-Shift Methodology

    International Nuclear Information System (INIS)

    Mandelli, D.; Yilmaz, A.; Metzroth, K.; Aldemir, T.; Denning, R.

    2010-01-01

    A new generation of dynamic methodologies is being developed for nuclear reactor probabilistic risk assessment (PRA) which explicitly account for the time element in modeling the probabilistic system evolution and use numerical simulation tools to account for possible dependencies between failure events. The dynamic event tree (DET) approach is one of these methodologies. One challenge with dynamic PRA methodologies is the large amount of data they produce which may be difficult to analyze without appropriate software tools. The concept of 'data mining' is well known in the computer science community and several methodologies have been developed in order to extract useful information from a dataset with a large number of records. Using the dataset generated by the DET analysis of the reactor vessel auxiliary cooling system (RVACS) of an ABR-1000 for an aircraft crash recovery scenario and the Mean-Shift Methodology for data mining, it is shown how clusters of transients with common characteristics can be identified and classified. (authors)

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

    International Nuclear Information System (INIS)

    Kim, Yun Goo; Oh, Eung Se

    2016-01-01

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

  3. Software safety analysis practice in installation phase

    International Nuclear Information System (INIS)

    Huang, H. W.; Chen, M. H.; Shyu, S. S.

    2010-10-01

    This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

  4. Software safety analysis practice in installation phase

    Energy Technology Data Exchange (ETDEWEB)

    Huang, H. W.; Chen, M. H.; Shyu, S. S., E-mail: hwhwang@iner.gov.t [Institute of Nuclear Energy Research, No. 1000 Wenhua Road, Chiaan Village, Longtan Township, 32546 Taoyuan County, Taiwan (China)

    2010-10-15

    This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

  5. Methodological proposal for occupational health and safety actions in research laboratories with nanotechnologies activities.

    Science.gov (United States)

    Andrade, Luís Renato Balbão; Amaral, Fernando Gonçalves

    2012-01-01

    Nanotechnologies is a multidisciplinary set of techniques to manipulate matter on nanoscale level, more precisely particles below 100 nm whose characteristic due to small size is essentially different from those found in macro form materials. Regarding to these new properties of the materials there are knowledge gaps about the effects of these particles on human organism and the environment. Although it still being considered emerging technology it is growing increasingly fast as well as the number of products using nanotechnologies in some production level and so the number of researchers involved with the subject. Given this scenario and based on literature related, a comprehensive methodology for health and safety at work for researching laboratories with activities in nanotechnologies was developed, based on ILO structure guidelines for safety and health at work system on which a number of nanospecific recommendations were added to. The work intends to offer food for thought on controlling risks associated to nanotechnologies.

  6. Eliciting and communicating expert judgments: Methodology and application to nuclear safety

    International Nuclear Information System (INIS)

    Winterfeldt, D. von

    1989-01-01

    The most ambitious and certainly the most extensive formal expert judgment process was the elicitation of numerous events and uncertain quantities for safety issues in five nuclear power plants in the U.S. The general methodology for formal expert elicitations are described. An overview of the expert elicitation process of NUREG 1150 is provided and the elicitation of probabilities for the interfacing systems loss of coolant accident LOCA (ISL) in PWRs is given as an example of this elicitation process. Some lessons learned from this study are presented. (DG)

  7. Methodological aspects on drug receptor binding analysis

    International Nuclear Information System (INIS)

    Wahlstroem, A.

    1978-01-01

    Although drug receptors occur in relatively low concentrations, they can be visualized by the use of appropriate radioindicators. In most cases the procedure is rapid and can reach a high degree of accuracy. Specificity of the interaction is studied by competition analysis. The necessity of using several radioindicators to define a receptor population is emphasized. It may be possible to define isoreceptors and drugs with selectivity for one isoreceptor. (Author)

  8. 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)

  9. Survey of systems safety analysis methods and their application to nuclear waste management systems

    Energy Technology Data Exchange (ETDEWEB)

    Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

    1981-11-01

    This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.

  10. Survey of systems safety analysis methods and their application to nuclear waste management systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

    1981-11-01

    This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study

  11. Survey of systems safety analysis methods and their application to nuclear waste management systems

    Science.gov (United States)

    Pelto, P. J.; Winegardner, W. K.; Gallucci, R. H. V.

    1981-11-01

    This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause consequence diagrams, GO methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.

  12. MSSV Modeling for Wolsong-1 Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Bok Ja; Choi, Chul Jin; Kim, Seoung Rae [KEPCO EandC, Daejeon (Korea, Republic of)

    2010-10-15

    The main steam safety valves (MSSVs) are installed on the main steam line to prevent the overpressurization of the system. MSSVs are held in closed position by spring force and the valves pop open by internal force when the main steam pressure increases to open set pressure. If the overpressure condition is relieved, the valves begin to close. For the safety analysis of anticipated accident condition, the safety systems are modeled conservatively to simulate the accident condition more severe. MSSVs are also modeled conservatively for the analysis of over-pressurization accidents. In this paper, the pressure transient is analyzed at over-pressurization condition to evaluate the conservatism for MSSV models

  13. Advanced Methodology for Containment M/E Release Analysis

    International Nuclear Information System (INIS)

    Kim, C. W.; Park, S. J.; Song, J. H.; Choi, H. R.; Seo, J. T.

    2006-01-01

    Recently, a new mass and energy (M/E) release analysis methodology for the equipment environmental qualification (EEQ) on loss-of-coolant accident (LOCA) has been developed and adopted on small break LOCA (SBLOCA). This new M/E release analysis methodology for EEQ is extended to the M/E release analysis for the containment design for large break LOCA (LBLOCA) and main steam line break (MSLB) accident. The advanced methodology of the M/E release analysis for the containment design includes the same engine as the M/E methodology for EEQ, however, conservative approaches for the M/E release such as break spillage model and multiplier on heat transfer coefficient (HTC) etc. are added. The computer code systems used in this methodology are RELAP5K/CONTEMPT4 (or RELAP5- ME) like KREM (KEPRI Realistic Evaluation Model) which couples RELAP5/MOD3.1/K and CONTEMPT4/ MOD5. RELAP5K is based on RELAP5/MOD3.1/K and includes conservatisms for the M/E release and long-term analysis model. The advanced methodology adopting the recent analysis technology is able to calculate the various transient stages of a LOCA in a single code system and also can calculate the M/E release analysis during the long term cooling period with the containment response. This advanced methodology for the M/E release is developed based on the LOCA and applied to the MSLB. The results are compared with the Ulchin Nuclear Unit (UCN) 3 and 4 FSAR

  14. Analysis of Alternatives for Risk Assessment Methodologies and Tools

    Energy Technology Data Exchange (ETDEWEB)

    Nachtigal, Noel M. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). System Analytics; Fruetel, Julia A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Gleason, Nathaniel J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Helms, Jovana [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Imbro, Dennis Raymond [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Sumner, Matthew C. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis

    2013-10-01

    The purpose of this document is to provide a basic overview and understanding of risk assessment methodologies and tools from the literature and to assess the suitability of these methodologies and tools for cyber risk assessment. Sandia National Laboratories (SNL) performed this review in support of risk modeling activities performed for the Stakeholder Engagement and Cyber Infrastructure Resilience (SECIR) division of the Department of Homeland Security (DHS) Office of Cybersecurity and Communications (CS&C). The set of methodologies and tools covered in this document is not intended to be exhaustive; instead, it focuses on those that are commonly used in the risk assessment community. The classification of methodologies and tools was performed by a group of analysts with experience in risk analysis and cybersecurity, and the resulting analysis of alternatives has been tailored to address the needs of a cyber risk assessment.

  15. Electromagnetic safety analysis during major disruption

    International Nuclear Information System (INIS)

    Gao Chunming; Wang Yafei; Chen Zhi; Feng Kaiming

    2006-01-01

    The electromagnetic safety analysis during major disruption is important for safety analysis of the CH HCSB TBM. In this paper, using finite element method, the electromagnetic safety analysis of the CH HCSB TBM is carried out in consideration of major disruption. First, the finite element models of the CH HCSB TBM and its sub-module are established; second, the distributions of the induced eddy currents and electromagnetic forces on the whole CH HCSB TBM module and its sub-module are calculated; third, the torquemoment on whole CH HCSB TBM module and its sub-module are calculated from the distributions of the electromagnetic forces. Comparing the maximum allowable values of the parameters of the materials with the calculated data, the electromagnetic safety of the CH HCSB TBM is investigated. (authors)

  16. Safety analysis of autonomous excavator functionality

    International Nuclear Information System (INIS)

    Seward, D.; Pace, C.; Morrey, R.; Sommerville, I.

    2000-01-01

    This paper presents an account of carrying out a hazard analysis to define the safety requirements for an autonomous robotic excavator. The work is also relevant to the growing generic class of heavy automated mobile machinery. An overview of the excavator design is provided and the concept of a safety manager is introduced. The safety manager is an autonomous module responsible for all aspects of system operational safety, and is central to the control system's architecture. Each stage of the hazard analysis is described, i.e. system model creation, hazard definition and hazard analysis. Analysis at an early stage of the design process, and on a system that interfaces directly to an unstructured environment, exposes certain issues relevant to the application of current hazard analysis methods. The approach taken in the analysis is described. Finally, it is explained how the results of the hazard analysis have influenced system design, in particular, safety manager specifications. Conclusions are then drawn about the applicability of hazard analysis of requirements in general, and suggestions are made as to how the approach can be taken further

  17. Advanced Power Plant Development and Analysis Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    A.D. Rao; G.S. Samuelsen; F.L. Robson; B. Washom; S.G. Berenyi

    2006-06-30

    Under the sponsorship of the U.S. Department of Energy/National Energy Technology Laboratory, a multi-disciplinary team led by the Advanced Power and Energy Program of the University of California at Irvine is defining the system engineering issues associated with the integration of key components and subsystems into advanced power plant systems with goals of achieving high efficiency and minimized environmental impact while using fossil fuels. These power plant concepts include 'Zero Emission' power plants and the 'FutureGen' H2 co-production facilities. The study is broken down into three phases. Phase 1 of this study consisted of utilizing advanced technologies that are expected to be available in the 'Vision 21' time frame such as mega scale fuel cell based hybrids. Phase 2 includes current state-of-the-art technologies and those expected to be deployed in the nearer term such as advanced gas turbines and high temperature membranes for separating gas species and advanced gasifier concepts. Phase 3 includes identification of gas turbine based cycles and engine configurations suitable to coal-based gasification applications and the conceptualization of the balance of plant technology, heat integration, and the bottoming cycle for analysis in a future study. Also included in Phase 3 is the task of acquiring/providing turbo-machinery in order to gather turbo-charger performance data that may be used to verify simulation models as well as establishing system design constraints. The results of these various investigations will serve as a guide for the U. S. Department of Energy in identifying the research areas and technologies that warrant further support.

  18. Applications of a methodology for the analysis of learning trends in nuclear power plants

    International Nuclear Information System (INIS)

    Cho, Hang Youn; Choi, Sung Nam; Yun, Won Yong

    1995-01-01

    A methodology is applied to identify the learning trend related to the safety and availability of U.S. commercial nuclear power plants. The application is intended to aid in reducing likelihood of human errors. To assure that the methodology can be easily adapted to various types of classification schemes of operation data, a data bank classified by the Transient Analysis Classification and Evaluation(TRACE) scheme is selected for the methodology. The significance criteria for human-initiated events affecting the systems and for events caused by human deficiencies were used. Clustering analysis was used to identify the learning trend in multi-dimensional histograms. A computer code is developed based on the K-Means algorithm and applied to find the learning period in which error rates are monotonously decreasing with plant age

  19. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

    2001-02-01

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  20. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

    Kim, Hee Cheol; Kim, K. K.; Kim, S. H.

    2002-04-01

    The state-of-the-arts for the integral reactor was performed to investigate the safety features. The safety and performance of SMART were assessed using the technologies developed during the study. For this purpose, the computer code system and the analysis methodology were developed and the safety and performance analyses on SMART basic design were carried out for the design basis event and accident. The experimental facilities were designed for the core flow distribution test and the self-pressurizing pressurizer performance test. The tests on the 2-phase critical flow with non-condensable gas were completed and the results were used to assess the critical flow model. Probabilistic Safety Assessment(PSA) was carried out to evaluate the safety level and to optimize the design by identifying and remedying any weakness in the design. A joint study with KINS was carried out to promote licensing environment. The generic safety issues of integral reactors were identified and the solutions were formulated. The economic evaluation of the SMART desalination plant and the activities related to the process control were carried out in the scope of the study

  1. Relationships between accident investigations, risk analysis, and safety management

    International Nuclear Information System (INIS)

    Harms-Ringdahl, Lars

    2004-01-01

    Several different approaches to achieve safety are in common use, and examples are accident investigations (AI), risk analysis (RA), and safety management systems (SMS). The meaning of these concepts and their practical applications vary quite a lot, which might cause confusion. A summary of definitions is presented. A general comparison is made of application areas and methodology. A proposal is made how to indicate parameters of variation. At one end of the scale there are organisations, which are highly organised in respect to safety. At the other end are small companies with informal safety routines. Although the three concepts differ in a number of respects, there are many links between them which is illustrated in a model. A number of relations have been described mainly concerned with more advanced organisations. Behind the practical safety work, there are varying sets of more or less explicit explanations and theories on safety and accident causation. Depending on the theory applied, the relations between approaches can be more or less clear and essential

  2. CONTENT ANALYSIS, DISCOURSE ANALYSIS, AND CONVERSATION ANALYSIS: PRELIMINARY STUDY ON CONCEPTUAL AND THEORETICAL METHODOLOGICAL DIFFERENCES

    Directory of Open Access Journals (Sweden)

    Anderson Tiago Peixoto Gonçalves

    2016-08-01

    Full Text Available This theoretical essay aims to reflect on three models of text interpretation used in qualitative research, which is often confused in its concepts and methodologies (Content Analysis, Discourse Analysis, and Conversation Analysis. After the presentation of the concepts, the essay proposes a preliminary discussion on conceptual and theoretical methodological differences perceived between them. A review of the literature was performed to support the conceptual and theoretical methodological discussion. It could be verified that the models have differences related to the type of strategy used in the treatment of texts, the type of approach, and the appropriate theoretical position.

  3. 50 Years of coastal erosion analysis: A new methodological approach.

    Science.gov (United States)

    Prieto Campos, Antonio; Diaz Cuevas, Pilar; Ojeda zujar, Jose; Guisado-Pintado, Emilia

    2017-04-01

    Coasts over the world have been subjected to increased anthropogenic pressures which combined with natural hazards impacts (storm events, rising sea-levels) have led to strong erosion problems with negative impacts on the economy and the safety of coastal communities. The Andalusian coast (South Spain) is a renowned global tourist destination. In the past decades a deep transformation in the economic model led to significant land use changes: strong regulation of rivers, urbanisation and occupation of dunes, among others. As a result irreversible transformations on the coastline, from the aggressive urbanisation undertaken, are now to be faced by local authorities and suffered by locals and visitors. Moreover, the expected impacts derived from the climate change aggravated by anthropic activities emphasises the need for tools that facilitates decision making for a sustainable coastal management. In this contribution a homogeneous (only a proxy and one photointerpreter) methodology is proposed for the calculation of coastal erosion rates of exposed beaches in Andalusia (640 km) through the use of detailed series (1:2500) of open source orthophotographies for the period (1956-1977-2001-2011). The outstanding combination of the traditional software DSAS (Digital Shoreline Analysis System) with a spatial database (PostgreSQL) which integrates the resulting erosion rates with related coastal thematic information (geomorphology, presence of engineering infrastructures, dunes and ecosystems) enhances the capacity of analysis and exploitation. Further, the homogeneity of the method used allows the comparison of the results among years in a highly diverse coast, with both Mediterranean and Atlantic façades. The novelty development and integration of a PostgreSQL/Postgis database facilitates the exploitation of the results by the user (for instance by relating calculated rates with other thematic information as geomorphology of the coast or the presence of a dune field on

  4. Development of a systematic methodology to select hazard analysis techniques for nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Vanderley de; Reis, Sergio Carneiro dos; Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mails: vasconv@cdtn.br; reissc@cdtn.br; aclc@cdtn.br; Jordao, Elizabete [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Faculdade de Engenharia Quimica]. E-mail: bete@feq.unicamp.br

    2008-07-01

    In order to comply with licensing requirements of regulatory bodies risk assessments of nuclear facilities should be carried out. In Brazil, such assessments are part of the Safety Analysis Reports, required by CNEN (Brazilian Nuclear Energy Commission), and of the Risk Analysis Studies, required by the competent environmental bodies. A risk assessment generally includes the identification of the hazards and accident sequences that can occur, as well as the estimation of the frequencies and effects of these unwanted events on the plant, people, and environment. The hazard identification and analysis are also particularly important when implementing an Integrated Safety, Health, and Environment Management System following ISO 14001, BS 8800 and OHSAS 18001 standards. Among the myriad of tools that help the process of hazard analysis can be highlighted: CCA (Cause- Consequence Analysis); CL (Checklist Analysis); ETA (Event Tree Analysis); FMEA (Failure Mode and Effects Analysis); FMECA (Failure Mode, Effects and Criticality Analysis); FTA (Fault Tree Analysis); HAZOP (Hazard and Operability Study); HRA (Human Reliability Analysis); Pareto Analysis; PHA (Preliminary Hazard Analysis); RR (Relative Ranking); SR (Safety Review); WI (What-If); and WI/CL (What-If/Checklist Analysis). The choice of a particular technique or a combination of techniques depends on many factors like motivation of the analysis, available data, complexity of the process being analyzed, expertise available on hazard analysis, and initial perception of the involved risks. This paper presents a systematic methodology to select the most suitable set of tools to conduct the hazard analysis, taking into account the mentioned involved factors. Considering that non-reactor nuclear facilities are, to a large extent, chemical processing plants, the developed approach can also be applied to analysis of chemical and petrochemical plants. The selected hazard analysis techniques can support cost

  5. Development of a systematic methodology to select hazard analysis techniques for nuclear facilities

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Reis, Sergio Carneiro dos; Costa, Antonio Carlos Lopes da; Jordao, Elizabete

    2008-01-01

    In order to comply with licensing requirements of regulatory bodies risk assessments of nuclear facilities should be carried out. In Brazil, such assessments are part of the Safety Analysis Reports, required by CNEN (Brazilian Nuclear Energy Commission), and of the Risk Analysis Studies, required by the competent environmental bodies. A risk assessment generally includes the identification of the hazards and accident sequences that can occur, as well as the estimation of the frequencies and effects of these unwanted events on the plant, people, and environment. The hazard identification and analysis are also particularly important when implementing an Integrated Safety, Health, and Environment Management System following ISO 14001, BS 8800 and OHSAS 18001 standards. Among the myriad of tools that help the process of hazard analysis can be highlighted: CCA (Cause- Consequence Analysis); CL (Checklist Analysis); ETA (Event Tree Analysis); FMEA (Failure Mode and Effects Analysis); FMECA (Failure Mode, Effects and Criticality Analysis); FTA (Fault Tree Analysis); HAZOP (Hazard and Operability Study); HRA (Human Reliability Analysis); Pareto Analysis; PHA (Preliminary Hazard Analysis); RR (Relative Ranking); SR (Safety Review); WI (What-If); and WI/CL (What-If/Checklist Analysis). The choice of a particular technique or a combination of techniques depends on many factors like motivation of the analysis, available data, complexity of the process being analyzed, expertise available on hazard analysis, and initial perception of the involved risks. This paper presents a systematic methodology to select the most suitable set of tools to conduct the hazard analysis, taking into account the mentioned involved factors. Considering that non-reactor nuclear facilities are, to a large extent, chemical processing plants, the developed approach can also be applied to analysis of chemical and petrochemical plants. The selected hazard analysis techniques can support cost

  6. Model Based Safety Analysis with smartIflow †

    Directory of Open Access Journals (Sweden)

    Philipp Hönig

    2017-01-01

    Full Text Available Verification of safety requirements is one important task during the development of safety critical systems. The increasing complexity of systems makes manual analysis almost impossible. This paper introduces a new methodology for formal verification of technical systems with smartIflow (State Machines for Automation of Reliability-related Tasks using Information FLOWs. smartIflow is a new modeling language that has been especially designed for the purpose of automating the safety analysis process in early product life cycle stages. It builds up on experience with existing approaches. As is common practice in current approaches, components are modeled as finite state machines. However, new concepts are introduced to describe component interactions. Events play a major role for internal interactions between components as well as for external (user interactions. Our approach to the verification of formally specified safety requirements is a two-step method. First, an exhaustive simulation creates knowledge about a great variety of possible behaviors of the system, especially including reactions on suddenly occurring (possibly intermittent faults. In the second step, safety requirements specified in CTL (Computation Tree Logic are verified using model checking techniques, and counterexamples are generated if these are not satisfied. The practical applicability of this approach is demonstrated based on a Java implementation using a simple Two-Tank-Pump-Consumer system.

  7. Comprehensive default methodology for the analysis of exposures to mixtures of chemicals accidentally released to the atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Craig, D.K. [Westinghouse Savannah River Company, Aiken, SC (United States); Baskett, R.L.; Powell, T.J. [Lawrence Livermore National Lab., CA (United States); Davis, J.S. [Duke Engineering Services, Hanford, Inc., Charlotte, NC (United States); Dukes, L.L. [Automated Solutions of Albuquerque, NM (United States); Hansen, D.J. [Brookhaven National Lab., Upton, NY (United States); Petrocchi, A.J. [AlphaTRAC, Inc. (United States); Sutherland, P.J. [Battelle Memorial Institute, Columbus, OH (United States)

    1997-07-01

    Safety analysis of Department of Energy (DOE) facilities requires consideration of potential exposures to mixtures of chemicals released to the atmosphere. Exposure to chemical mixtures may lead to additive, synergistic, or antagonistic health effects. In the past, the consequences of each chemical have been analyzed separately. This approach may not adequately protect the health of persons exposed to mixtures. However, considerable time would be required to evaluate all possible mixtures. The objective of this paper is to present reasonable default methodology developed by the EFCOG Safety Analysis Working Group Nonradiological Hazardous Material Subgroup (NHMS) for use in safety analysis within the DOE Complex.

  8. Comprehensive default methodology for the analysis of exposures to mixtures of chemicals accidentally released to the atmosphere

    International Nuclear Information System (INIS)

    Craig, D.K.; Baskett, R.L.; Powell, T.J.; Davis, J.S.; Dukes, L.L.; Hansen, D.J.; Petrocchi, A.J.; Sutherland, P.J.

    1997-01-01

    Safety analysis of Department of Energy (DOE) facilities requires consideration of potential exposures to mixtures of chemicals released to the atmosphere. Exposure to chemical mixtures may lead to additive, synergistic, or antagonistic health effects. In the past, the consequences of each chemical have been analyzed separately. This approach may not adequately protect the health of persons exposed to mixtures. However, considerable time would be required to evaluate all possible mixtures. The objective of this paper is to present reasonable default methodology developed by the EFCOG Safety Analysis Working Group Nonradiological Hazardous Material Subgroup (NHMS) for use in safety analysis within the DOE Complex

  9. 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

  10. Methodology of safety evaluation about land disposal of low level radioactive wastes

    International Nuclear Information System (INIS)

    Suzuki, Atsuyuki

    1986-01-01

    Accompanying the progress of the construction project of low level radioactive waste storage facilities in Aomori Prefecture, the full scale land disposal of low level radioactive wastes shows its symptom also in Japan. In this report, the scientific methodology to explain the safety about the land disposal of low level radioactive wastes is discussed. The land disposal of general wastes by shallow burying has already had sufficient results. In the case of low level radioactive wastes, also the land disposal by shallow burying is considered. Low level radioactive wastes can be regarded as one form of industrial wastes, as there are many common parts in the scientific and theoretical base of the safety. Attention is paid most to the contamination of ground water. Low level radioactive wastes are solid wastes, accordingly the degree of contamination should be less. The space in which ground water existes, the phenomena of ground water movement, the phenomena of ground water dispersion and Fick's law, the adsorption effect of strata, and the evaluation of source term are explained. These are the method to analyze the degree of contamination from safety evaluation viewpoint. (Kako, I.)

  11. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

  12. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    International Nuclear Information System (INIS)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

  13. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  14. Application of Software Safety Analysis Methods

    International Nuclear Information System (INIS)

    Park, G. Y.; Hur, S.; Cheon, S. W.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.; Lee, S. J.; Koo, Y. H.

    2009-01-01

    A fully digitalized reactor protection system, which is called the IDiPS-RPS, was developed through the KNICS project. The IDiPS-RPS has four redundant and separated channels. Each channel is mainly composed of a group of bistable processors which redundantly compare process variables with their corresponding setpoints and a group of coincidence processors that generate a final trip signal when a trip condition is satisfied. Each channel also contains a test processor called the ATIP and a display and command processor called the COM. All the functions were implemented in software. During the development of the safety software, various software safety analysis methods were applied, in parallel to the verification and validation (V and V) activities, along the software development life cycle. The software safety analysis methods employed were the software hazard and operability (Software HAZOP) study, the software fault tree analysis (Software FTA), and the software failure modes and effects analysis (Software FMEA)

  15. From Safety Analysis to Formal Specification

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark; Ravn, Anders P.; Stavridou, Victoria

    1998-01-01

    Software for safety critical systems must deal with the hazards identified bysafety analysis. This paper investigates, how the results of onesafety analysis technique, fault trees, are interpreted as software safetyrequirements to be used in the program design process. We propose thatfault tree...... analysis and program development use the samesystem model. This model is formalized in areal-time, interval logic, based on a conventional dynamic systems modelwith state evolving over time. Fault trees are interpreted astemporal formulas, and it is shown how such formulas can be usedfor deriving safety...

  16. Safety characteristics analysis of nuclear power plants with PHWR PT

    International Nuclear Information System (INIS)

    Stosic, Z.

    1983-01-01

    The paper deals with analysis of basic safety characteristics of heavy water Candu reactor. Inherent safety characteristics, r/a material inventory, systematization of normal abnormal and transient conditions, safety systems and availability analysis are considered. (author)

  17. Safety assessment for deep underground disposal vault-pathways analysis

    International Nuclear Information System (INIS)

    Lyon, R.B.; Rosinger, E.L.J.

    1980-01-01

    The concept verification phase of the Canadian programme for the disposal of nuclear fuel waste encompasses a period of about three years before the start of site selection. During this time, the methodology for Environmental and Safety Assessment studies is being developed by focusing on a model site. Pathways analysis is an important component of these studies. It involves the prediction of the rate at which radionuclides might be released from a disposal vault and travel through the geosphere and biosphere to reach man. The pathways analysis studies cover three major topics: geosphere pathways analysis, biosphere pathways analysis and potentially-disruptive-phenomena analysis. Geosphere pathways analysis includes a total systems analysis, using the computer program GARD2, vault analysis, which considers container failure and waste leaching, hydrogeological modelling and geochemical modelling. Biosphere pathways analysis incorporates a compartmental modelling approach using the computer program RAMM, and a food chain analysis using the computer program FOOD II. Potentially-disruptive-phenomena analysis involves the estimation of the probability and consequences of events such as earthquakes which might reduce the effectiveness of the barriers preventing the release of radionuclides. The current stage of development of the required methodology and data is discussed in each of the three areas and preliminary results are presented. (author)

  18. Systemic design methodologies for electrical energy systems analysis, synthesis and management

    CERN Document Server

    Roboam, Xavier

    2012-01-01

    This book proposes systemic design methodologies applied to electrical energy systems, in particular analysis and system management, modeling and sizing tools. It includes 8 chapters: after an introduction to the systemic approach (history, basics & fundamental issues, index terms) for designing energy systems, this book presents two different graphical formalisms especially dedicated to multidisciplinary devices modeling, synthesis and analysis: Bond Graph and COG/EMR. Other systemic analysis approaches for quality and stability of systems, as well as for safety and robustness analysis tools are also proposed. One chapter is dedicated to energy management and another is focused on Monte Carlo algorithms for electrical systems and networks sizing. The aim of this book is to summarize design methodologies based in particular on a systemic viewpoint, by considering the system as a whole. These methods and tools are proposed by the most important French research laboratories, which have many scientific partn...

  19. PIXE methodology of rare earth element analysis and its applications

    International Nuclear Information System (INIS)

    Ma Xinpei

    1992-01-01

    The Proton Induced X-ray Emission (PIXE) methodology of rare earth element (REEs) analysis is discussed, including the significance of REE analysis, the principle of PIXE applied to REE, selection of characteristic X-ray for Lanthanide series elements, deconvolution of highly over lapped PIXE spectrum and minimum detection limit (MDL) of REEs. Some practical applications are presented. And the specialities of PIXE analysis to the high pure REE chemicals are discussed. (author)

  20. A Global Sensitivity Analysis Methodology for Multi-physics Applications

    Energy Technology Data Exchange (ETDEWEB)

    Tong, C H; Graziani, F R

    2007-02-02

    Experiments are conducted to draw inferences about an entire ensemble based on a selected number of observations. This applies to both physical experiments as well as computer experiments, the latter of which are performed by running the simulation models at different input configurations and analyzing the output responses. Computer experiments are instrumental in enabling model analyses such as uncertainty quantification and sensitivity analysis. This report focuses on a global sensitivity analysis methodology that relies on a divide-and-conquer strategy and uses intelligent computer experiments. The objective is to assess qualitatively and/or quantitatively how the variabilities of simulation output responses can be accounted for by input variabilities. We address global sensitivity analysis in three aspects: methodology, sampling/analysis strategies, and an implementation framework. The methodology consists of three major steps: (1) construct credible input ranges; (2) perform a parameter screening study; and (3) perform a quantitative sensitivity analysis on a reduced set of parameters. Once identified, research effort should be directed to the most sensitive parameters to reduce their uncertainty bounds. This process is repeated with tightened uncertainty bounds for the sensitive parameters until the output uncertainties become acceptable. To accommodate the needs of multi-physics application, this methodology should be recursively applied to individual physics modules. The methodology is also distinguished by an efficient technique for computing parameter interactions. Details for each step will be given using simple examples. Numerical results on large scale multi-physics applications will be available in another report. Computational techniques targeted for this methodology have been implemented in a software package called PSUADE.

  1. Methodological Tool or Methodology? Beyond Instrumentality and Efficiency with Qualitative Data Analysis Software

    Directory of Open Access Journals (Sweden)

    Pengfei Zhao

    2016-04-01

    Full Text Available Qualitative data analysis software (QDAS has become increasingly popular among researchers. However, very few discussions have developed regarding the effect of QDAS on the validity of qualitative data analysis. It is a pressing issue, especially because the recent proliferation of conceptualizations of validity has challenged, and to some degree undermined, the taken-for-granted connection between the methodologically neutral understanding of validity and QDAS. This article suggests an alternative framework for examining the relationship between validity and the use of QDAS. Shifting the analytic focus from instrumentality and efficiency of QDAS to the research practice itself, we propose that qualitative researchers should formulate a "reflective space" at the intersection of their methodological approach, the built-in validity structure of QDAS and the specific research context, in order to make deliberative and reflective methodological decisions. We illustrate this new framework through discussion of a collaborative action research project. URN: http://nbn-resolving.de/urn:nbn:de:0114-fqs1602160

  2. Projective goals - concepts and pragmatic aspects based on the terminology and methodology of safety science

    International Nuclear Information System (INIS)

    Compes, P.C.

    1991-01-01

    Protective goals set the line of orientation of tasks and activities in the field of accident prevention. They have to be based on safety-science methods in order to develop from the conceptual idea to the practically feasible solution, while using the scientific methods to take into account the facts and the capabilities of a situation and, proceeding from them, finding an efficient and rational, optimal pragmatic approach by way of various strategies or tactics. In this process, the activities of defining, informing, thinking and developing need the proper terminology. Safety is absence of danger, protection is limitation of danger and prevention of damage. So it is protection what is needed with danger being given, and risks have to be minimized. Riskology is a novel method of safety science, combining risk analysis and risk control into a systematic concept which is practice-oriented. Applying this to the field of nuclear engineering, the hitherto achieved should receive new impulses. (orig.) [de

  3. The practical implementation of integrated safety management for nuclear safety analysis and fire hazards analysis documentation

    International Nuclear Information System (INIS)

    COLLOPY, M.T.

    1999-01-01

    In 1995 Mr. Joseph DiNunno of the Defense Nuclear Facilities Safety Board issued an approach to describe the concept of an integrated safety management program which incorporates hazard and safety analysis to address a multitude of hazards affecting the public, worker, property, and the environment. Since then the U S . Department of Energy (DOE) has adopted a policy to systematically integrate safety into management and work practices at all levels so that missions can be completed while protecting the public, worker, and the environment. While the DOE and its contractors possessed a variety of processes for analyzing fire hazards at a facility, activity, and job; the outcome and assumptions of these processes have not always been consistent for similar types of hazards within the safety analysis and the fire hazard analysis. Although the safety analysis and the fire hazard analysis are driven by different DOE Orders and requirements, these analyses should not be entirely independent and their preparation should be integrated to ensure consistency of assumptions, consequences, design considerations, and other controls. Under the DOE policy to implement an integrated safety management system, identification of hazards must be evaluated and agreed upon to ensure that the public. the workers. and the environment are protected from adverse consequences. The DOE program and contractor management need a uniform, up-to-date reference with which to plan. budget, and manage nuclear programs. It is crucial that DOE understand the hazards and risks necessarily to authorize the work needed to be performed. If integrated safety management is not incorporated into the preparation of the safety analysis and the fire hazard analysis, inconsistencies between assumptions, consequences, design considerations, and controls may occur that affect safety. Furthermore, confusion created by inconsistencies may occur in the DOE process to grant authorization of the work. In accordance with

  4. A methodology to incorporate organizational factors into human reliability analysis

    International Nuclear Information System (INIS)

    Li Pengcheng; Chen Guohua; Zhang Li; Xiao Dongsheng

    2010-01-01

    A new holistic methodology for Human Reliability Analysis (HRA) is proposed to model the effects of the organizational factors on the human reliability. Firstly, a conceptual framework is built, which is used to analyze the causal relationships between the organizational factors and human reliability. Then, the inference model for Human Reliability Analysis is built by combining the conceptual framework with Bayesian networks, which is used to execute the causal inference and diagnostic inference of human reliability. Finally, a case example is presented to demonstrate the specific application of the proposed methodology. The results show that the proposed methodology of combining the conceptual model with Bayesian Networks can not only easily model the causal relationship between organizational factors and human reliability, but in a given context, people can quantitatively measure the human operational reliability, and identify the most likely root causes or the prioritization of root causes caused human error. (authors)

  5. Core melt progression and consequence analysis methodology development in support of the Savannah River Reactor PSA

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Sharp, D.A.; Amos, C.N.; Wagner, K.C.; Bradley, D.R.

    1992-01-01

    A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained

  6. IAEA Review for Gap Analysis of Safety Analysis Capability

    International Nuclear Information System (INIS)

    Basic, Ivica; Kim, Manwoong; Huges, Peter; Lim, B-K; D'Auria, Francesco; Louis, Vidard Michael

    2014-01-01

    The IAEA Asian Nuclear Safety Network (ANSN) was launched in 2002 in the framework of the Extra Budgetary Programme (EBP) on the Safety of Nuclear Installations in the South East Asia, Pacific and Far East Countries. The main objective is to strengthen and expand human and advanced Information Technology (IT) network to pool, analyse and share nuclear safety knowledge and practical experience for peaceful uses in this region. Under the ANSN framework, a technical group on Safety Analysis (SATG) was established in 2004 aimed to providing a forum for the exchange of experience in the following areas of safety analysis: · To provide a forum for an exchange of experience in the area of safety analysis, · To maintain and improve the knowledge on safety analysis method, · To enhance the utilization of computer codes, · To pool and analyse the issues related with safety analysis of research reactor, and · To facilitate mutual interested on safety analysis among member countries. A sustainable and successful nuclear energy programme requires a strong technical infrastructure, including a workforce made up of highly specialized and well-educated professionals. A significant portion of this technical capacity must be dedicated to safety- especially to safety analysis- as only then can it serve as the basis for making the right decisions during the planning, licensing, construction and operation of new nuclear facilities. In this regard, the IAEA has provided ANSN member countries with comprehensive training opportunities for capacity building in safety analysis. Nevertheless, the SATG recognizes that it is difficult to achieve harmonization in this area among all member countries because of their different competency levels. Therefore, it is necessary to quickly identify the most obvious gaps in safety analysis capability and then to use existing resources to begin to fill those gaps. The goal of this Expert Mission (EM) for gap finding service is to facilitate

  7. Waste Isolation Pilot Plant Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

  8. Waste Isolation Pilot Plant Safety Analysis Report

    International Nuclear Information System (INIS)

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

  9. Interpersonal Dynamics in a Simulated Prison: A Methodological Analysis

    Science.gov (United States)

    Banuazizi, Ali; Movahedi, Siamak

    1975-01-01

    A critical overview is presented of the Stanford Prison Experiment, conducted by Zimbardo and his coinvestigators in which they attempted a structural analysis of the problems of imprisonment. Key assumptions are questioned, primarily on methodological grounds, which casts doubts on the plausibility of the experimenters' final causal inferences.…

  10. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

    1992-12-01

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

  11. A Local Approach Methodology for the Analysis of Ultimate Strength ...

    African Journals Online (AJOL)

    The local approach methodology in contrast to classical fracture mechanics can be used to predict the onset of tearing fracture, and the effects of geometry in tubular joints. Finite element analysis of T-joints plate geometries, and tubular joints has been done. The parameters of constraint, equivalent stress, plastic strain and ...

  12. METHODOLOGY FOR INSTITUTIONAL ANALYSIS OF STABILITY OF REGIONAL FINANCIAL SYSTEM

    Directory of Open Access Journals (Sweden)

    A. V. Milenkov

    2016-01-01

    Full Text Available The relevance of the article due to the urgent need to develop a methodological framework in the field of regional finance research dictated by the substantial increase in the volume and composition of the socio-economic problems, the solution of which, including financial support, the responsibility of the public authorities of the Russian Federation. The article presents the results of the author's research in the field of institutional analysis of the stability of the regional financial system as a set of institutions and organizations interacting with the regional real sector of economy.Methodology. The methodological basis of this article are the economic and statistical methods of analysis, legal documents in the field of the sustainability of the regional financial system, publications in the field of economic and financial security.Conclusions / relevance. The practical significance of the work lies in the provisions of orientation, conclusions and recommendations aimed at the widespread use of search and adaptation of the institutional analysis of the sources of the regional stability of the financial system, which can be used by the legislative and executive authorities of the Russian Federation, the Ministry of Defence in the current activity.Methodological approaches to the structuring objectives of institutional analysis on the basis of the hierarchical representation of the institutional environment of functioning of federal subject the financial system.

  13. Safety analysis for complex systems

    Science.gov (United States)

    Onesty, J. P.; Peercy, R. L., Jr.

    1981-01-01

    Operational risk assessment considers hardware, environment, and human factors. Technique starts with division of postulated mission into segments which are further subdivided into separate operational steps. Consequences of steps, nonoccurrence, premature operation, out-of-sequence operation, and inadvertent execution are examined at subevent, event, and phase levels. Hazards are identified and treated individually. Analysis is well suited to application in energy and transportation fields.

  14. A Statistical Methodology for Determination of Safety Systems Actuation Setpoints Based on Extreme Value Statistics

    Directory of Open Access Journals (Sweden)

    D. R. Novog

    2008-01-01

    Full Text Available This paper provides a novel and robust methodology for determination of nuclear reactor trip setpoints which accounts for uncertainties in input parameters and models, as well as accounting for the variations in operating states that periodically occur. Further it demonstrates that in performing best estimate and uncertainty calculations, it is critical to consider the impact of all fuel channels and instrumentation in the integration of these uncertainties in setpoint determination. This methodology is based on the concept of a true trip setpoint, which is the reactor setpoint that would be required in an ideal situation where all key inputs and plant responses were known, such that during the accident sequence a reactor shutdown will occur which just prevents the acceptance criteria from being exceeded. Since this true value cannot be established, the uncertainties in plant simulations and plant measurements as well as operational variations which lead to time changes in the true value of initial conditions must be considered. This paper presents the general concept used to determine the actuation setpoints considering the uncertainties and changes in initial conditions, and allowing for safety systems instrumentation redundancy. The results demonstrate unique statistical behavior with respect to both fuel and instrumentation uncertainties which has not previously been investigated.

  15. Accident Analysis and Highway Safety

    Directory of Open Access Journals (Sweden)

    Omar Noorliyana

    2017-01-01

    Full Text Available Since 2010, Federal Route FT050 (Jalan Batu Pahat-Kluang has undergone many changes, including the improvement of geometric features (i.e., construction of median, dedicated U-turns and additional lanes and upgrading the quality of the road surface. Unfortunately, even with these enhancements, accidents continue to occur along this route. This study covered both accident analysis and blackspot study. Accident point weightage was used to identify blackspot locations. The results reveal hazardous road locations and blackspot ranking along the route.

  16. 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)

  17. Active disturbance rejection control: methodology and theoretical analysis.

    Science.gov (United States)

    Huang, Yi; Xue, Wenchao

    2014-07-01

    The methodology of ADRC and the progress of its theoretical analysis are reviewed in the paper. Several breakthroughs for control of nonlinear uncertain systems, made possible by ADRC, are discussed. The key in employing ADRC, which is to accurately determine the "total disturbance" that affects the output of the system, is illuminated. The latest results in theoretical analysis of the ADRC-based control systems are introduced. Copyright © 2014 ISA. Published by Elsevier Ltd. All rights reserved.

  18. Methodological proposals to elevate the quality of the training in the field of Health and Safety Policy in the Work.

    Directory of Open Access Journals (Sweden)

    Hortensia Hernández Fernández

    2014-09-01

    Full Text Available The training and the development of the Human Capital as regards Health and Safety Policy are important challenges that at the present time have the organizations, fundamentally because they are directed to the prevention of work accidents and professional illnesses. This work is a contribution to the methodological proposals to elevate the quality of the training process in the field of Health and Safety Policy for persons attending the activity and that don't have pedagogic formation. For its development theoretical and empiric investigation methods were used, obtaining the elaboration of methodological proposals which is about the main theoretical aspects that facilitate the elaboration of the objectives from the methodological and didactic point of view, the selection of the methods, teaching means and types of evaluations that can be applied during the training process according to the focuses of the formation, professional development and organizational forms to impart the educational actions in the field of Health and Safety Policy.

  19. K West integrated water treatment system subproject safety analysis document

    International Nuclear Information System (INIS)

    SEMMENS, L.S.

    1999-01-01

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System

  20. K West integrated water treatment system subproject safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    SEMMENS, L.S.

    1999-02-24

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System.

  1. Disposal criticality analysis methodology for fissile waste forms

    International Nuclear Information System (INIS)

    Davis, J.W.; Gottlieb, P.

    1998-03-01

    A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository

  2. Methodological Variability Using Electronic Nose Technology For Headspace Analysis

    Science.gov (United States)

    Knobloch, Henri; Turner, Claire; Spooner, Andrew; Chambers, Mark

    2009-05-01

    Since the idea of electronic noses was published, numerous electronic nose (e-nose) developments and applications have been used in analyzing solid, liquid and gaseous samples in the food and automotive industry or for medical purposes. However, little is known about methodological pitfalls that might be associated with e-nose technology. Some of the methodological variation caused by changes in ambient temperature, using different filters and changes in mass flow rates are described. Reasons for a lack of stability and reproducibility are given, explaining why methodological variation influences sensor responses and why e-nose technology may not always be sufficiently robust for headspace analysis. However, the potential of e-nose technology is also discussed.

  3. A Tool for the Concise Analysis of Patient Safety Incidents.

    Science.gov (United States)

    Pham, Julius Cuong; Hoffman, Carolyn; Popescu, Ioana; Ijagbemi, O Mayowa; Carson, Kathryn A

    2016-01-01

    Patient safety incidents, sometimes referred to as adverse events, incidents, or patient safety events, are too common an occurrence in health care. Most methods for incident analysis are time and labor intensive. Given the significant resource requirements of a root cause analysis, for example, there is a need for a more targeted and efficient method of analyzing a larger number of incidents. Although several concise incident analysis tools are in existence, there are no published studies regarding their usability or effectiveness. Building on previous efforts, a Concise Incident Analysis (CIA) methodology and tool were developed to facilitate analysis of no- or low-harm incidents. Staff from 11 hospitals in five countries-Australia, Canada, Hong Kong, India, and the United States-pilot tested the tool in two phases. The tool was evaluated and refined after each phase on the basis of user perceptions of usability and effectiveness. From September 2013 through January 2014, 52 patient safety incidents were analyzed. A broad variety of incident types were investigated, the most frequent being patient falls (25%). Incidents came from a variety of hospital work areas, the most frequent being from the medical ward (37%). Most incidents investigated resulted in temporary harm or no harm (94%). All or most sites found the tool "understandable" (100%), "easy to use" (89%), and "effective" (89%). Some 95% of participants planned to continue to use all or some parts of the tool after the pilot. Qualitative feedback suggested that the tool allowed analysis of incidents that were not currently being analyzed because of insufficient resources. The tool was described as simple to use, easy to document, and aligned with the flow of the incident analysis. A concise tool for the investigation of patient safety incidents with low or no harm was well accepted across a select group of hospitals from five countries.

  4. Two methodologies for optical analysis of contaminated engine lubricants

    International Nuclear Information System (INIS)

    Aghayan, Hamid; Yang, Jun; Bordatchev, Evgueni

    2012-01-01

    The performance, efficiency and lifetime of modern combustion engines significantly depend on the quality of the engine lubricants. However, contaminants, such as gasoline, moisture, coolant and wear particles, reduce the life of engine mechanical components and lubricant quality. Therefore, direct and indirect measurements of engine lubricant properties, such as physical-mechanical, electro-magnetic, chemical and optical properties, are intensively utilized in engine condition monitoring systems and sensors developed within the last decade. Such sensors for the measurement of engine lubricant properties can be used to detect a functional limit of the in-use lubricant, increase drain interval and reduce the environmental impact. This paper proposes two new methodologies for the quantitative and qualitative analysis of the presence of contaminants in the engine lubricants. The methodologies are based on optical analysis of the distortion effect when an object image is obtained through a thin random optical medium (e.g. engine lubricant). The novelty of the proposed methodologies is in the introduction of an object with a known periodic shape behind a thin film of the contaminated lubricant. In this case, an acquired image represents a combined lubricant–object optical appearance, where an a priori known periodic structure of the object is distorted by a contaminated lubricant. In the object shape-based optical analysis, several parameters of an acquired optical image, such as the gray scale intensity of lubricant and object, shape width at object and lubricant levels, object relative intensity and width non-uniformity coefficient are newly proposed. Variations in the contaminant concentration and use of different contaminants lead to the changes of these parameters measured on-line. In the statistical optical analysis methodology, statistical auto- and cross-characteristics (e.g. auto- and cross-correlation functions, auto- and cross-spectrums, transfer function

  5. Safety analysis of accident localization system

    International Nuclear Information System (INIS)

    1999-01-01

    A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

  6. 14 CFR 33.75 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.75 Safety analysis. (a... applicable: (i) Aircraft-level devices and procedures assumed to be associated with a typical installation...

  7. Uncertainty analysis for Ulysses safety evaluation report

    International Nuclear Information System (INIS)

    Frank, M.V.

    1991-01-01

    As part of the effort to review the Ulysses Final Safety Analysis Report and to understand the risk of plutonium release from the Ulysses spacecraft General Purpose Heat Source---Radioisotope Thermal Generator (GPHS-RTG), the Interagency Nuclear Safety Review Panel (INSRP) and the author performed an integrated, quantitative analysis of the uncertainties of the calculated risk of plutonium release from Ulysses. Using state-of-art probabilistic risk assessment technology, the uncertainty analysis accounted for both variability and uncertainty of the key parameters of the risk analysis. The results show that INSRP had high confidence that risk of fatal cancers from potential plutonium release associated with calculated launch and deployment accident scenarios is low

  8. Hazard classification methodology

    International Nuclear Information System (INIS)

    Brereton, S.J.

    1996-01-01

    This document outlines the hazard classification methodology used to determine the hazard classification of the NIF LTAB, OAB, and the support facilities on the basis of radionuclides and chemicals. The hazard classification determines the safety analysis requirements for a facility

  9. Deterministic and probabilistic approach to safety analysis

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1980-01-01

    The examples discussed in this paper show that reliability analysis methods fairly well can be applied in order to interpret deterministic safety criteria in quantitative terms. For further improved extension of applied reliability analysis it has turned out that the influence of operational and control systems and of component protection devices should be considered with the aid of reliability analysis methods in detail. Of course, an extension of probabilistic analysis must be accompanied by further development of the methods and a broadening of the data base. (orig.)

  10. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    International Nuclear Information System (INIS)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S.

    2015-01-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  11. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  12. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Suk, S. D.

    2002-05-01

    In the present study, the KALIMER safety analysis has been made for the transients considered in the design concept, hypothetical core disruptive accident (HCDA), and containment performance with the establishment of the design basis. Such analyses have not been possible without the computer code improvement, and the experience attained during this research period must have greatly contributed to the achievement of the self reliance in the domestic technology establishment on the safety analysis areas of the conceptual design. The safety analysis codes have been improved to extend their applicable ranges for detailed conceptual design, and a basic computer code system has been established for HCDA analysis. A code-to-code comparison analysis has been performed as a part of code verification attempt, and the leading edge technology of JNC also has been brought for the technology upgrade. In addition, the research and development on the area of the database establishment has been made for the efficient and systematic project implementation of the conceptual design, through performances on the development of a project scheduling management, integration of the individually developed technology, establishment of the product database, and so on, taking into account coupling of the activities conducted in each specific area

  13. The PEC reactor. Safety analysis: Detailed reports

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

  14. 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

  15. RISK ANALYSIS IN CONSTRUCTION PROJECTS: A PRACTICAL SELECTION METHODOLOGY

    OpenAIRE

    Alberto De Marco; Muhammad Jamaluddin Thaheem

    2014-01-01

    Project Risk Management (PRM) is gaining attention from researchers and practitioners in the form of sophisticated tools and techniques to help construction managers perform risk management. However, the large variety of techniques has made selecting an appropriate solution a complex and risky task in itself. Accordingly, this study proposes a practical framework methodology to assist construction project managers and practitioners in choosing a suitable risk analysis technique based on selec...

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

    DEFF Research Database (Denmark)

    Duijm, Nijs Jan

    2017-01-01

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

  17. Safety and efficacy of subretinal visual implants in humans: methodological aspects.

    Science.gov (United States)

    Stingl, Katarina; Bach, Michael; Bartz-Schmidt, Karl-Ulrich; Braun, Angelika; Bruckmann, Anna; Gekeler, Florian; Greppmaier, Udo; Hörtdörfer, Gernot; Kusnyerik, Akos; Peters, Tobias; Wilhelm, Barbara; Wilke, Robert; Zrenner, Eberhart

    2013-01-01

    Replacing the function of visual pathway neurons by electronic implants is a novel approach presently explored by various groups in basic research and clinical trials. The novelty raises unexplored methodological aspects of clinical trial design that may require adaptation and validation. We present procedures of efficacy and safety testing for subretinal visual implants in humans, as developed during our pilot trial 2005 to 2009 and multi-centre clinical trial since 2010. Planning such a trial requires appropriate inclusion and exclusion criteria. For subretinal electronic visual implants, patients with photoreceptor degeneration are the target patient group, whereas presence of additional diseases affecting clear optic media or the visual pathway must be excluded. Because sham surgery is not possible, a masked study design with implant power ON versus OFF is necessary. Prior to the efficacy testing by psychophysical tests, the implant's technical characteristics have to be controlled via electroretinography (ERG). Moreover the testing methods require adaptation to the particular technology. We recommend standardised tasks first to determine the light perception thresholds, light localisation and movement detection, followed by grating acuity and vision acuity test via Landolt C rings. A laboratory setup for assessing essential activities of daily living is presented. Subjective visual experiences with the implant in a natural environment, as well as questionnaires and psychological counselling are further important aspects. A clinical trial protocol for artificial vision in humans, which leads a patient from blindness to the state of very low vision is a challenge and cannot be defined completely prior to the study. Available tests of visual function may not be sufficiently suited for efficacy testing of artificial vision devices. A protocol based on experience with subretinal visual implants in 22 patients is presented that has been found adequate to monitor

  18. Crossing trend analysis methodology and application for Turkish rainfall records

    Science.gov (United States)

    Şen, Zekâi

    2018-01-01

    Trend analyses are the necessary tools for depicting possible general increase or decrease in a given time series. There are many versions of trend identification methodologies such as the Mann-Kendall trend test, Spearman's tau, Sen's slope, regression line, and Şen's innovative trend analysis. The literature has many papers about the use, cons and pros, and comparisons of these methodologies. In this paper, a completely new approach is proposed based on the crossing properties of a time series. It is suggested that the suitable trend from the centroid of the given time series should have the maximum number of crossings (total number of up-crossings or down-crossings). This approach is applicable whether the time series has dependent or independent structure and also without any dependence on the type of the probability distribution function. The validity of this method is presented through extensive Monte Carlo simulation technique and its comparison with other existing trend identification methodologies. The application of the methodology is presented for a set of annual daily extreme rainfall time series from different parts of Turkey and they have physically independent structure.

  19. Safety analysis of an ancient iron structure

    International Nuclear Information System (INIS)

    Kweon, Young Gak; Yoon, Byeng Hyun; Lim, Jae Kyun; Lee, Sung Bum

    2002-01-01

    Safety analysis of an ancient iron structure, Danggan, constructed over than a thousand years ago was performed. The structure is consisted of 24 iron cylinders of which the total height is about 15.4 m. The analysis was done by the ultrasonic test to measure thickness of each cylinder, the radiographic test to investigate the inside of cylinders, the measurement of inclination of the structure and the structural analysis to estimate the stress level applied by the wind. Results showed that Danggan structure was on state being well safe at present, but it could be dangerous when the inclination of the structure becomes severely progressive.

  20. Development and application of an uncertainty methodology for the Sizewell B large LOCA safety case

    International Nuclear Information System (INIS)

    Lightfoot, P.; Trow, M.

    1994-01-01

    This paper presents an uncertainty methodology which has been successfully applied to the licensing of Sizewell B for large break LOCA. The emphasis of this approach has been on gaining a detailed understanding of the physical process and of the sensitivity to individual phenomena. The major contributors to uncertainty have been identified, and have subsequently been included in a combined uncertainty analysis. The combined uncertainty analysis demonstrated that uncertainties did not combine in a highly non-linear manner phenomena such as the random reflood effect and clad ballooning have been treated a bounding biases in the assessment of the overall bounding peak clad temperature. The plant initial and boundary conditions have been conservatively defined for the uncertainty analysis. A better estimate calculation, which uses more realistic assumptions, shows a large benefit in the predicted peak clad temperature, thereby demonstrating the conservatism of the uncertainty analysis. The UK licensing regime is not prescriptive in terms of the approach to large LOCA analysis, and no attempt has been made to apply a formal probability or confidence limit to the final bounding peak clad temperature is conservative. The Sizewell B uncertainty analysis was completed within the timescale and resources limitations. It has been shown to be practical in its application and reductions in the required analysis scope have been identified for any future plants of similar design

  1. Live kidney donation: are concerns about long-term safety justified?—A methodological review

    International Nuclear Information System (INIS)

    Janki, Shiromani; Steyerberg, Ewout W.; Hofman, Albert; IJzermans, Jan N. M.

    2017-01-01

    Live kidney donors are exhaustively screened pre-donation, creating a cohort inherently healthier at baseline than the general population. In recent years, three renowned research groups reported unfavourable outcomes for live kidney donors post-donation that contradicted their previous studies. Here, we compared the study design and analysis of the most recent and previous studies to determine whether the different outcomes were due to methodological design or reflect a real potential disadvantage for living kidney donors. All six studies on long-term risk after live kidney donation were thoroughly screened for the selection of study population, controls, data quality, and statistical analysis. Our detailed review of the methodology revealed key differences with respect to selection of donors and compared non-donors, data quality, follow-up duration, and statistical analysis. In all studies, the comparison group of non-donors was healthier than the donors due to more extensive exclusion criteria for non-donors. Five of the studies used both restriction and matching to address potential confounding. Different matching strategies and statistical analyses were used in the more recent studies compared to previous studies and follow-up was longer. Recently published papers still face bias. Strong points compared to initial analyses are the extended follow-up time, large sample sizes and better analysis, hence increasing the reliability to estimate potential risks for living kidney donors on the long-term. Future studies should focus on equal selection criteria for donors and non-donors, and in the analysis, follow-up duration, matched sets, and low absolute risks among donors should be accounted for when choosing the statistical technique.

  2. Applications of the TSUNAMI sensitivity and uncertainty analysis methodology

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Hopper, Calvin M.; Elam, Karla R.; Goluoglu, Sedat; Parks, Cecil V.

    2003-01-01

    The TSUNAMI sensitivity and uncertainty analysis tools under development for the SCALE code system have recently been applied in four criticality safety studies. TSUNAMI is used to identify applicable benchmark experiments for criticality code validation, assist in the design of new critical experiments for a particular need, reevaluate previously computed computational biases, and assess the validation coverage and propose a penalty for noncoverage for a specific application. (author)

  3. Safety of GM crops: compositional analysis.

    Science.gov (United States)

    Brune, Philip D; Culler, Angela Hendrickson; Ridley, William P; Walker, Kate

    2013-09-04

    The compositional analysis of genetically modified (GM) crops has continued to be an important part of the overall evaluation in the safety assessment program for these materials. The variety and complexity of genetically engineered traits and modes of action that will be used in GM crops in the near future, as well as our expanded knowledge of compositional variability and factors that can affect composition, raise questions about compositional analysis and how it should be applied to evaluate the safety of traits. The International Life Sciences Institute (ILSI), a nonprofit foundation whose mission is to provide science that improves public health and well-being by fostering collaboration among experts from academia, government, and industry, convened a workshop in September 2012 to examine these and related questions, and a series of papers has been assembled to describe the outcomes of that meeting.

  4. Computational methods for nuclear criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.

    1992-01-01

    Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)

  5. Summary of the Supplemental Model Reports Supporting the Disposal Criticality Analysis Methodology Topical Report

    International Nuclear Information System (INIS)

    Brownson, D. A.

    2002-01-01

    The Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) has committed to a series of model reports documenting the methodology to be utilized in the Disposal Criticality Analysis Methodology Topical Report (YMP 2000). These model reports detail and provide validation of the methodology to be utilized for criticality analyses related to: (1) Waste form/waste package degradation; (2) Waste package isotopic inventory; (3) Criticality potential of degraded waste form/waste package configurations (effective neutron multiplication factor); (4) Probability of criticality (for each potential critical configuration as well as total event); and (5) Criticality consequences. This purpose of this summary report is to provide a status of the model reports and a schedule for their completion. This report also provides information relative to the model report content and validation. The model reports and their revisions are being generated as a result of: (1) Commitments made in the Disposal Criticality Analysis Methodology Topical Report (YMP 2000); (2) Open Items from the Safety Evaluation Report (Reamer 2000); (3) Key Technical Issue agreements made during DOE/U.S. Nuclear Regulatory Commission (NRC) Technical Exchange Meeting (Reamer and Williams 2000); and (4) NRC requests for additional information (Schlueter 2002)

  6. Reliability Analysis for Safety Grade PLC

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Kyung Chul; Hwang, Sung Jae; Jung, Tae Hyok; Kim, Tae Hee; Song, Seung Whan [POSCO ICT Co., Seoul (Korea, Republic of)

    2010-10-15

    In this paper, describe reliability analysis for digital safety grade PLC which developed with the aim to use the operating nuclear power plants and new plants by POSCO ICT co., POSAFE-Q consist of the Sub Rack, power modules, processor modules, communication modules, digital input / output module (DI / DO), analog input / output modules (AI / AO), pulse counter module, TC (Thermocouple), RTD (Resistance Temperature Detector), Local Repeater

  7. Computer graphics in reactor safety analysis

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.

    1989-01-01

    This paper describes a family of three computer graphics codes designed to assist the analyst in three areas: the modelling of complex three-dimensional finite element models of reactor structures; the interpretation of computational results; and the reporting of the results of numerical simulations. The purpose and key features of each code are presented. The graphics output used in actual safety analysis are used to illustrate the capabilities of each code. 5 refs., 10 figs

  8. Multi-resolution Convolution Methodology for ICP Waveform Morphology Analysis.

    Science.gov (United States)

    Shaw, Martin; Piper, Ian; Hawthorne, Christopher

    2016-01-01

    Intracranial pressure (ICP) monitoring is a key clinical tool in the assessment and treatment of patients in neurointensive care. ICP morphology analysis can be useful in the classification of waveform features.A methodology for the decomposition of an ICP signal into clinically relevant dimensions has been devised that allows the identification of important ICP waveform types. It has three main components. First, multi-resolution convolution analysis is used for the main signal decomposition. Then, an impulse function is created, with multiple parameters, that can represent any form in the signal under analysis. Finally, a simple, localised optimisation technique is used to find morphologies of interest in the decomposed data.A pilot application of this methodology using a simple signal has been performed. This has shown that the technique works with performance receiver operator characteristic area under the curve values for each of the waveform types: plateau wave, B wave and high and low compliance states of 0.936, 0.694, 0.676 and 0.698, respectively.This is a novel technique that showed some promise during the pilot analysis. However, it requires further optimisation to become a usable clinical tool for the automated analysis of ICP signals.

  9. Development of test methodology for dynamic mechanical analysis instrumentation

    Science.gov (United States)

    Allen, V. R.

    1982-01-01

    Dynamic mechanical analysis instrumentation was used for the development of specific test methodology in the determination of engineering parameters of selected materials, esp. plastics and elastomers, over a broad range of temperature with selected environment. The methodology for routine procedures was established with specific attention given to sample geometry, sample size, and mounting techniques. The basic software of the duPont 1090 thermal analyzer was used for data reduction which simplify the theoretical interpretation. Clamps were developed which allowed 'relative' damping during the cure cycle to be measured for the fiber-glass supported resin. The correlation of fracture energy 'toughness' (or impact strength) with the low temperature (glassy) relaxation responses for a 'rubber-modified' epoxy system was negative in result because the low-temperature dispersion mode (-80 C) of the modifier coincided with that of the epoxy matrix, making quantitative comparison unrealistic.

  10. Comparative analysis of safety related site characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Johan (ed.)

    2010-12-15

    This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

  11. 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

  12. Multi-criteria analysis for evaluating the radiological and ecological safety measures in radioactive waste management

    International Nuclear Information System (INIS)

    Sazykina, T.G.; Kryshev, I.I.

    2006-01-01

    A methodological approach is presented for multicriterial evaluating the effectiveness of radiation ecological safety measures during radioactive waste management. The approach is based on multicriterial analysis with consideration of radiological, ecological, social, economical consequences of various safety measures. The application of the multicriterial approach is demonstrated taking as an example of decision-making on the most effective actions for rehabilitation of a water subject, contaminated with radionuclides [ru

  13. Qualitative analysis in reliability and safety studies

    International Nuclear Information System (INIS)

    Worrell, R.B.; Burdick, G.R.

    1976-01-01

    The qualitative evaluation of system logic models is described as it pertains to assessing the reliability and safety characteristics of nuclear systems. Qualitative analysis of system logic models, i.e., models couched in an event (Boolean) algebra, is defined, and the advantages inherent in qualitative analysis are explained. Certain qualitative procedures that were developed as a part of fault-tree analysis are presented for illustration. Five fault-tree analysis computer-programs that contain a qualitative procedure for determining minimal cut sets are surveyed. For each program the minimal cut-set algorithm and limitations on its use are described. The recently developed common-cause analysis for studying the effect of common-causes of failure on system behavior is explained. This qualitative procedure does not require altering the fault tree, but does use minimal cut sets from the fault tree as part of its input. The method is applied using two different computer programs. 25 refs

  14. Methodology, outcome, safety and in vivo accuracy in traditional frame-based stereoelectroencephalography.

    Science.gov (United States)

    van der Loo, Lars E; Schijns, Olaf E M G; Hoogland, Govert; Colon, Albert J; Wagner, G Louis; Dings, Jim T A; Kubben, Pieter L

    2017-07-05

    Stereoelectroencephalography (SEEG) is an established diagnostic technique for the localization of the epileptogenic zone in drug-resistant epilepsy. In vivo accuracy of SEEG electrode positioning is of paramount importance since higher accuracy may lead to more precise resective surgery, better seizure outcome and reduction of complications. To describe experiences with the SEEG technique in our comprehensive epilepsy center, to illustrate surgical methodology, to evaluate in vivo application accuracy and to consider the diagnostic yield of SEEG implantations. All patients who underwent SEEG implantations between September 2008 and April 2016 were analyzed. Planned electrode trajectories were compared with post-implantation trajectories after fusion of pre- and postoperative imaging. Quantitative analysis of deviation using Euclidean distance and directional errors was performed. Explanatory variables for electrode accuracy were analyzed using linear regression modeling. The surgical methodology, procedure-related complications and diagnostic yield were reported. Seventy-six implantations were performed in 71 patients, and a total of 902 electrodes were implanted. Median entry and target point deviations were 1.54 mm and 2.93 mm. Several factors that predicted entry and target point accuracy were identified. The rate of major complications was 2.6%. SEEG led to surgical therapy of various modalities in 53 patients (69.7%). This study demonstrated that entry and target point localization errors can be predicted by linear regression models, which can aid in identification of high-risk electrode trajectories and further enhancement of accuracy. SEEG is a reliable technique, as demonstrated by the high accuracy of conventional frame-based implantation methodology and the good diagnostic yield.

  15. Considerations on a methodological framework for the analysis of texts

    Directory of Open Access Journals (Sweden)

    David Andrés Camargo Mayorga

    2017-03-01

    Full Text Available This article presents a review of relevant literature for the construction of a methodological framework for the analysis of texts in applied social sciences, such as economics, which we have supported in the main hermeneutical approaches from philosophy, linguistics and social sciences. In essence, they assume that every discourse carries meaning - be it truthful or not - and that they express complex social relations. Thus, any analysis of content happens finally to be a certain type of hermeneutics (interpretation, while trying to account for multiple phenomena immersed in the production, application, use and reproduction of knowledge within the text. When applying discourse analysis in teaching texts in economic sciences, we find traces of legalistic, political, ethnocentric tendencies, among other discourses hidden from the text. For this reason, the analysis of the internal discourse of the text allows us to delve inside the state ideology and its underlying or latent discourses.

  16. Methodological challenges in qualitative content analysis: A discussion paper.

    Science.gov (United States)

    Graneheim, Ulla H; Lindgren, Britt-Marie; Lundman, Berit

    2017-09-01

    This discussion paper is aimed to map content analysis in the qualitative paradigm and explore common methodological challenges. We discuss phenomenological descriptions of manifest content and hermeneutical interpretations of latent content. We demonstrate inductive, deductive, and abductive approaches to qualitative content analysis, and elaborate on the level of abstraction and degree of interpretation used in constructing categories, descriptive themes, and themes of meaning. With increased abstraction and interpretation comes an increased challenge to demonstrate the credibility and authenticity of the analysis. A key issue is to show the logic in how categories and themes are abstracted, interpreted, and connected to the aim and to each other. Qualitative content analysis is an autonomous method and can be used at varying levels of abstraction and interpretation. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. New Methodology of Block Cipher Analysis using Chaos Game

    Directory of Open Access Journals (Sweden)

    Budi Sulistyo

    2011-08-01

    Full Text Available Block cipher analysis covers randomness analysis and cryptanalysis. This paper proposes a new method potentially used for randomness analysis and cryptanalysis. The method uses true random sequence concept as a reference for measuring randomness level of a random sequence. By using this concept, this paper defines bias which represents violation of a random sequence from true random sequence. In this paper, block cipher is treated as a mapping function of a discrete time dynamical system. The dynamical system framework is used to make the application of various analysis techniques developed in dynamical system field becomes possible. There are three main parts of the methodology presented in this paper: the dynamical system framework for block cipher analysis, a new chaos game scheme and an extended measure concept related to chaos game and fractal analysis. This paper also presents the general procedures of the proposed method, which includes: symbolic dynamic analysis of discrete dynamical system whose block cipher as its mapping function, random sequence construction, the random sequence usage as input of a chaos game scheme, output measurement of chaos game scheme using extended measure concept, analysis the result of the measurement. The analysis process and of a specific real or sample block cipher and the analysis result are beyond the scope of this paper.

  18. Uncertainty propagation in probabilistic safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Fleming, P.V.

    1981-09-01

    The uncertainty propagation in probabilistic safety analysis of nuclear power plants, is done. The methodology of the minimal cut is implemented in the computer code SVALON and the results for several cases are compared with corresponding results obtained with the SAMPLE code, which employs the Monte Carlo method to propagate the uncertanties. The results have show that, for a relatively small number of dominant minimal cut sets (n approximately 25) and error factors (r approximately 5) the SVALON code yields results which are comparable to those obtained with SAMPLE. An analysis of the unavailability of the low pressure recirculation system of Angra 1 for both the short and long term recirculation phases, are presented. The results for the short term phase are in good agreement with the corresponding one given in WASH-1400. (E.G.) [pt

  19. 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.

  20. Methodology of assessment of the impact on the risk of changes of monitoring requirements of safety equipment integrating treatment and analysis of uncertainty of model and parameter; Metodologia de evaluacion del impacto en el riego de cambios de requisitos de vigilancia de equipos de seguridad integrando tratamiento y analisis de incertidumbres de modelo y parametros

    Energy Technology Data Exchange (ETDEWEB)

    Martorell, S.; Villamizar, M.; Marton, I.; Villanueva, J. f.; Carlos, S.; Sanchez, A. I.

    2013-07-01

    The feasibility of the proposed methodology has been demonstrated with the results obtained in a case of application for the analysis of changes in the requirements of maximum time of unavailability of system batteries using an APS level 1. The proposed methodology is consistent with the proposal by the American regulatory guide RG 1,174, applicable to LWR designs in Spanish, although large plants their level of technical development.

  1. Hazard Analysis and Safety Requirements for Small Drone Operations: To What Extent Do Popular Drones Embed Safety?

    Science.gov (United States)

    Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimihailidou, Maria Mikela

    2018-03-01

    Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this article presents: (1) a set of safety requirements generated from the application of the systems theoretic process analysis (STPA) technique on a generic small drone system; (2) a gap analysis between the set of safety requirements and the ones met by 19 popular drone models; (3) the extent of the differences between those models, their manufacturers, and the countries of origin; and (4) the association of drone prices with the extent they meet the requirements derived by STPA. The application of STPA resulted in 70 safety requirements distributed across the authority, manufacturer, end user, or drone automation levels. A gap analysis showed high dissimilarities regarding the extent to which the 19 drones meet the same safety requirements. Statistical results suggested a positive correlation between drone prices and the extent that the 19 drones studied herein met the safety requirements generated by STPA, and significant differences were identified among the manufacturers. This work complements the existing risk assessment frameworks for small drones, and contributes to the establishment of a commonly endorsed international risk analysis framework. Such a framework will support the development of a holistic and methodologically justified standardization scheme for small drone flights. © 2017 Society for Risk Analysis.

  2. Conception and methodology of a prospective safety report for uranium mills tailings ponds

    International Nuclear Information System (INIS)

    Weiss, D.; Larue, J.; Fischer- Appelt, K.

    2006-01-01

    Uranium mill tailings ponds (MTP) stand for the highest potential risk of all legacies from uranium milling. The overall objective of this work was to develop a standardised application document and a working tool for responsible authorities to conduct a continuous safety status description and assessment as well as a prognosis of the respective tailings pond site. - Safety Assessment Principles are based on guide and control values specified in or derived from laws, guidances, norms or standards for the main components who are: technical installations and geo-mechanics, radioactivity, chemical-toxic pollutants. - Data Base: Monitoring data (ground- and seepage water), Radon in near bottom air, Data and information from expertises, reports and technical documents, Site specific data and information (data base A.LAS.KA. / FbU). - Exposure Pathways Analysis. Determination of the radiation exposure to members of the public of different age caused by radioactivity discharges from the tailings pond by means of an authorized calculation procedure. 0.1 to 0.3 mSv/yr with a predominant contribution of ingestion of drinking water (about 80 %). Determination of the hazard resulting from chemical-toxic pollutants according to German regulations. Only relevance of Arsenic via seepage water path. - Forecast of Contaminant Propagation via Groundwater. A three dimensional site specific model was generated to forecast the contaminant distribution in the downstream groundwater flow by means of the available monitoring data. - Safety Assessment. The present status of the tailings pond 'Lengenfeld' was evaluated as to be safe for all three risk components and therefore it is no need for short term measures to minimize hazards or to reduce the contaminants spreading via groundwater. The geochemical environment of the tailings can be regarded as steady, but they are hydraulically tensed by overlaying spar cover. The low permeability of the tailings prevents a significant vertical

  3. System analysis of vehicle active safety problem

    Science.gov (United States)

    Buznikov, S. E.

    2018-02-01

    The problem of the road transport safety affects the vital interests of the most of the population and is characterized by a global level of significance. The system analysis of problem of creation of competitive active vehicle safety systems is presented as an interrelated complex of tasks of multi-criterion optimization and dynamic stabilization of the state variables of a controlled object. Solving them requires generation of all possible variants of technical solutions within the software and hardware domains and synthesis of the control, which is close to optimum. For implementing the task of the system analysis the Zwicky “morphological box” method is used. Creation of comprehensive active safety systems involves solution of the problem of preventing typical collisions. For solving it, a structured set of collisions is introduced with its elements being generated also using the Zwicky “morphological box” method. The obstacle speed, the longitudinal acceleration of the controlled object and the unpredictable changes in its movement direction due to certain faults, the road surface condition and the control errors are taken as structure variables that characterize the conditions of collisions. The conditions for preventing typical collisions are presented as inequalities for physical variables that define the state vector of the object and its dynamic limits.

  4. 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)

  5. Safety analysis of urban arterials at the meso level.

    Science.gov (United States)

    Li, Jia; Wang, Xuesong

    2017-11-01

    Urban arterials form the main structure of street networks. They typically have multiple lanes, high traffic volume, and high crash frequency. Classical crash prediction models investigate the relationship between arterial characteristics and traffic safety by treating road segments and intersections as isolated units. This micro-level analysis does not work when examining urban arterial crashes because signal spacing is typically short for urban arterials, and there are interactions between intersections and road segments that classical models do not accommodate. Signal spacing also has safety effects on both intersections and road segments that classical models cannot fully account for because they allocate crashes separately to intersections and road segments. In addition, classical models do not consider the impact on arterial safety of the immediately surrounding street network pattern. This study proposes a new modeling methodology that will offer an integrated treatment of intersections and road segments by combining signalized intersections and their adjacent road segments into a single unit based on road geometric design characteristics and operational conditions. These are called meso-level units because they offer an analytical approach between micro and macro. The safety effects of signal spacing and street network pattern were estimated for this study based on 118 meso-level units obtained from 21 urban arterials in Shanghai, and were examined using CAR (conditional auto regressive) models that corrected for spatial correlation among the units within individual arterials. Results showed shorter arterial signal spacing was associated with higher total and PDO (property damage only) crashes, while arterials with a greater number of parallel roads were associated with lower total, PDO, and injury crashes. The findings from this study can be used in the traffic safety planning, design, and management of urban arterials. Copyright © 2017 Elsevier Ltd. All

  6. Methodology applied in Cuba for siting, designing, and building a radioactive waste repository under safety conditions

    International Nuclear Information System (INIS)

    Orbera, L.; Peralta, J.L.; Franklin, R.; Gil, R.; Chales, G.; Rodriguez, A.

    1993-01-01

    The work presents the methodology used in Cuba for siting, designing, and building a radioactive waste repository safely. This methodology covers both the technical and socio-economic factors, as well as those of design and construction so as to have a safe siting for this kind of repository under Cuba especial condition. Applying this methodology will results in a safe repository

  7. 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.

  8. Proficiency examination in English language: Needs analysis and methodological proposals

    Directory of Open Access Journals (Sweden)

    Maria Elizete Luz Saes

    2014-04-01

    Full Text Available The purpose of this work is to provide tools for reflections on some learning difficulties presented by students when they are submitted to English proficiency examinations, as well as to suggest some methodological proposals that can be implemented among didactic support groups, monitoring or in classrooms, by means of face-to-face or distance learning activities. The observations resulting from the performance presented by the students have motivated the preparation of this paper, whose theoretical assumptions are based on the needs analysis of the target audience, the exploration of oral and written discursive genres and the possibilities of interaction provided by technological mediation.

  9. METHODOLOGY FOR ANALYSIS OF DECISION MAKING IN AIR NAVIGATION SYSTEM

    Directory of Open Access Journals (Sweden)

    Volodymyr Kharchenko

    2011-03-01

    Full Text Available Abstract. In the research of Air Navigation System as a complex socio-technical system the methodologyof analysis of human-operator's decision-making has been developed. The significance of individualpsychologicalfactors as well as the impact of socio-psychological factors on the professional activities of ahuman-operator during the flight situation development from normal to catastrophic were analyzed. On thebasis of the reflexive theory of bipolar choice the expected risks of decision-making by the Air NavigationSystem's operator influenced by external environment, previous experience and intentions were identified.The methods for analysis of decision-making by the human-operator of Air Navigation System usingstochastic networks have been developed.Keywords: Air Navigation System, bipolar choice, human operator, decision-making, expected risk, individualpsychologicalfactors, methodology of analysis, reflexive model, socio-psychological factors, stochastic network.

  10. A development of containment performance analysis methodology using GOTHIC code

    International Nuclear Information System (INIS)

    Lee, B. C.; Yoon, J. I.; Byun, C. S.; Lee, J. Y.; Lee, J. Y.

    2003-01-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code

  11. A development of containment performance analysis methodology using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. C.; Yoon, J. I. [Future and Challenge Company, Seoul (Korea, Republic of); Byun, C. S.; Lee, J. Y. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, J. Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-10-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code.

  12. Effectiveness of Occupational Health and Safety Training: A Systematic Review with Meta-Analysis

    Science.gov (United States)

    Ricci, Federico; Chiesi, Andrea; Bisio, Carlo; Panari, Chiara; Pelosi, Annalisa

    2016-01-01

    Purpose: This meta-analysis aims to verify the efficacy of occupational health and safety (OHS) training in terms of knowledge, attitude and beliefs, behavior and health. Design/methodology/approach: The authors included studies published in English (2007-2014) selected from ten databases. Eligibility criteria were studies concerned with the…

  13. Hydrogen safety : consequence analysis and hydrogen refuelling stations

    International Nuclear Information System (INIS)

    Benard, P.; Hourri, A.; Angers, B.; Mustafa, V.; Hay, D.R.; Tchouvelev, A.V.

    2006-01-01

    In addition to production and storage issues, the lack of specific standards for hydrogen as vehicle fuel is widely regarded as an obstacle to the introduction of hydrogen on the energy market. The tolerance of accidents involving emerging technologies, however rare, can be lower than for existing, familiar technologies, particularly when misperceptions of risk exist and little comparative risk analysis with more familiar fuels are available. This will result in a more conservative approach in establishing clearance distances and other safety issues for hydrogen refuelling stations by authorities having jurisdiction, which directly impact the cost of introducing a hydrogen refuelling infrastructure particularly in urban areas where real estate costs are high. Thus the lack of specific knowledge, experience and a clear and consistent methodology to assess the risk of hydrogen service stations and vehicles is widely regarded as a serious impediment to its use. In this presentation we present the R and D activities carried out under the Canadian Hydrogen Safety Programme and the Auto 21 Network of Centres of Excellence to develop a scientific and engineering basis for determining safety standards and safe industry practices specific to hydrogen, focusing on refuelling stations and hydrogen-fuelled vehicles. (author)

  14. Methodological Development of the Probabilistic Model of the Safety Assessment of Hontomin P.D.T

    International Nuclear Information System (INIS)

    Hurtado, A.; Eguilior, S.; Recreo, F.

    2011-01-01

    In the framework of CO 2 Capture and Geological Storage, Risk Analysis plays an important role, because it is an essential requirement of knowledge to make up local, national and supranational definition and planning of carbon injection strategies. This is because each project is at risk of failure. Even from the early stages, it should take into account the possible causes of this risk and propose corrective methods along the process, i.e., managing risk. Proper risk management reduces the negative consequences arising from the project. The main method of reduction or neutralizing of risk is mainly the identification, measurement and evaluation of it, together with the development of decision rules. This report presents the developed methodology for risk analysis and the results of its application. The risk assessment requires determination of the random variables that will influence the functioning of the system. It is very difficult to set up probability distribution of a random variable in the classical sense (objective probability) when a particular event rarely occurred or even it has a incomplete development. In this situation, we have to determine the subjective probability, especially at an early stage of projects, when we have not enough information about the system. This subjective probability is constructed from assessment of experts judgement to estimate the possibility of certain random events could happen depending on geological features of the area of application. The proposed methodology is based on the application of Bayesian Probabilistic Networks for estimating the probability of risk of leakage. These probabilistic networks can define graphically relations of dependence between the variables and joint probability function through a local factorization of probability functions. (Author) 98 refs.

  15. Development of methodology for evaluation of long-term safety aspects of organic cement paste components

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, M.; Holgersson, S. (Chalmers University of Technology, Goeteborg (Sweden)); Ervanne, H. (Helsinki Univ. (Finland)) (and others)

    2008-12-15

    Long-term safety aspects of superplasticizers (SP) and other cement paste components were studied in this joint Nagra - NUMO - SKB - Posiva project with aim to develop a methodology for the evaluation of the long-term safety aspects of superplasticizers (SP) and other organic components of cement pastes. The study also evaluated the effects of SPs and other cement paste components that have already been used or that are most likely to be used in the construction of the high-level nuclear waste repositories in Sweden, Switzerland, Finland and Japan. The main long-term safety issue of concern is whether the superplasticizers and/or other organic components of cement pastes might affect the transport properties of radionuclides. A full evaluation of whether the superplasticizers can be used in a high-level nuclear waste repository cannot be answered based on the studies but a classification of the superplasticizers based on their impact on sorption of radionuclides has been done. The basic methodology for testing, leaching and analyzing of leachants and solid samples of different types was developed at CRIEPI. Two different methodologies for studying the impact of SPs on the sorption of Eu on crushed rock were tested and developed by Helsinki University (HU) and Chalmers University of Technology (CTH). Methods for analyzing organics leaching from grouts were successfully tested by CRIEPI and CTH (Chalmers University of Technology). At CRIEPI the total organic content (TOC) of the leachants was analyzed by Infrared absorption spectrometry (IR) followed by Gel Permeation Chromatography (GPC) for the identification of the organic compounds. At CTH several different analytical methods were tested (e.g. IR, UV spectroscopy, NMR, MALDI-TOF), but these methods still require improvement. In addition to SPs, organics are present in several components of cement pastes, for example in cement grinding aid (CGA) and micro silica slurry. The results suggests that the main high

  16. Development of methodology for evaluation of long-term safety aspects of organic cement paste components

    International Nuclear Information System (INIS)

    Andersson, M.; Holgersson, S.; Ervanne, H.

    2008-12-01

    Long-term safety aspects of superplasticizers (SP) and other cement paste components were studied in this joint Nagra - NUMO - SKB - Posiva project with aim to develop a methodology for the evaluation of the long-term safety aspects of superplasticizers (SP) and other organic components of cement pastes. The study also evaluated the effects of SPs and other cement paste components that have already been used or that are most likely to be used in the construction of the high-level nuclear waste repositories in Sweden, Switzerland, Finland and Japan. The main long-term safety issue of concern is whether the superplasticizers and/or other organic components of cement pastes might affect the transport properties of radionuclides. A full evaluation of whether the superplasticizers can be used in a high-level nuclear waste repository cannot be answered based on the studies but a classification of the superplasticizers based on their impact on sorption of radionuclides has been done. The basic methodology for testing, leaching and analyzing of leachants and solid samples of different types was developed at CRIEPI. Two different methodologies for studying the impact of SPs on the sorption of Eu on crushed rock were tested and developed by Helsinki University (HU) and Chalmers University of Technology (CTH). Methods for analyzing organics leaching from grouts were successfully tested by CRIEPI and CTH (Chalmers University of Technology). At CRIEPI the total organic content (TOC) of the leachants was analyzed by Infrared absorption spectrometry (IR) followed by Gel Permeation Chromatography (GPC) for the identification of the organic compounds. At CTH several different analytical methods were tested (e.g. IR, UV spectroscopy, NMR, MALDI-TOF), but these methods still require improvement. In addition to SPs, organics are present in several components of cement pastes, for example in cement grinding aid (CGA) and micro silica slurry. The results suggests that the main high

  17. 3D analysis methods - Study and seminar[BWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Daaviittila, A. [Valtion Teknillinen Tutkimuskeskus (Finland)

    2003-10-01

    The first part of the report results from a study that was performed as a Nordic co-operation activity with active participation from Studsvik Scandpower and Westinghouse Atom in Sweden, and VTT in Finland. The purpose of the study was to identify and investigate the effects rising from using the 3D transient com-puter codes in BWR safety analysis, and their influence on the transient analysis methodology. One of the main questions involves the critical power ratio (CPR) calculation methodology. The present way, where the CPR calculation is per-formed with a separate hot channel calculation, can be artificially conservative. In the investigated cases, no dramatic minimum CPR effect coming from the 3D calculation is apparent. Some cases show some decrease in the transient change of minimum CPR with the 3D calculation, which confirms the general thinking that the 1D calculation is conservative. On the other hand, the observed effect on neutron flux behaviour is quite large. In a slower transient the 3D effect might be stronger. The second part of the report is a summary of a related seminar that was held on the 3D analysis methods. The seminar was sponsored by the Reactor Safety part (NKS-R) of the Nordic Nuclear Safety Research Programme (NKS). (au)

  18. Reanalysis of traffic enforcement data from Victoria : a methodological study into the evaluation of safety measures.

    NARCIS (Netherlands)

    Oppe, S. & Bijleveld, F.D.

    2003-01-01

    There is an increased interest in the Netherlands in the safety effects of traffic enforcement measures. This regards the intermediate effect of enforcement on behaviour as well as the final effect on safety itself. In addition to these general safety effects, regional differences are also of

  19. Ignalina Safety Analysis Group's report for the year 1998

    International Nuclear Information System (INIS)

    Uspuras, E.; Augutis, J.; Bubelis, E.; Cesna, B.; Kaliatka, A.

    1999-02-01

    Results of Ignalina NPP Safety Analysis Group's research are presented. The main fields of group's activities in 1998 were following: safety analysis of reactor's cooling system, safety analysis of accident localization system, investigation of the problem graphite - fuel channel, reactor core modelling, assistance to the regulatory body VATESI in drafting regulations and reviewing safety reports presented by Ignalina NPP during the process of licensing of unit 1

  20. Safety analysis reports. Current status (third key report)

    International Nuclear Information System (INIS)

    1999-01-01

    A review of Ukrainian regulations and laws concerned with Nuclear power and radiation safety is presented with an overview of the requirements for the Safety Analysis Report Contents. Status of Safety Analysis Reports (SAR) is listed for each particular Ukrainian NPP including SAR development schedules. Organisational scheme of SAR development works includes: general technical co-ordination on Safety Analysis Report development; list of leading organisations and utilization of technical support within international projects

  1. ACRR fuel storage racks criticality safety analysis

    International Nuclear Information System (INIS)

    Bodette, D.E.; Naegeli, R.E.

    1997-10-01

    This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

  2. ESSAA: Embedded system safety analysis assistant

    Science.gov (United States)

    Wallace, Peter; Holzer, Joseph; Guarro, Sergio; Hyatt, Larry

    1987-01-01

    The Embedded System Safety Analysis Assistant (ESSAA) is a knowledge-based tool that can assist in identifying disaster scenarios. Imbedded software issues hazardous control commands to the surrounding hardware. ESSAA is intended to work from outputs to inputs, as a complement to simulation and verification methods. Rather than treating the software in isolation, it examines the context in which the software is to be deployed. Given a specified disasterous outcome, ESSAA works from a qualitative, abstract model of the complete system to infer sets of environmental conditions and/or failures that could cause a disasterous outcome. The scenarios can then be examined in depth for plausibility using existing techniques.

  3. Development of a heat exchanger root-cause analysis methodology

    International Nuclear Information System (INIS)

    Jarrel, D.B.

    1989-01-01

    The objective of this work is to determine a generic methodology for approaching the accurate identification of the root cause of component failure. Root-cause determinations are an everyday challenge to plant personnel, but they are handled with widely differing degrees of success due to the approaches, levels of diagnostic expertise, and documentation. The criterion for success is simple: If the root cause of the failure has truly been determined and corrected, the same causal failure relationship will not be demonstrated again in the future. The approach to root-cause analysis (RCA) element definition was to first selectively choose and constrain a functionally significant component (in this case a component cooling water to service water heat exchanger) that has demonstrated prevalent failures. Then a root cause of failure analysis was performed by a systems engineer on a large number of actual failure scenarios. The analytical process used by the engineer was documented and evaluated to abstract the logic model used to arrive at the root cause. For the case of the heat exchanger, the actual root-cause diagnostic approach is described. A generic methodology for the solution of the root cause of component failure is demonstrable for this general heat exchanger sample

  4. A methodology for reliability analysis in health networks.

    Science.gov (United States)

    Spyrou, Stergiani; Bamidis, Panagiotis D; Maglaveras, Nicos; Pangalos, George; Pappas, Costas

    2008-05-01

    A reliability model for a health care domain based on requirement analysis at the early stage of design of regional health network (RHN) is introduced. RHNs are considered as systems supporting the services provided by health units, hospitals, and the regional authority. Reliability assessment in health care domain constitutes a field-of-quality assessment for RHN. A novel approach for predicting system reliability in the early stage of designing RHN systems is presented in this paper. The uppermost scope is to identify the critical processes of an RHN system prior to its implementation. In the methodology, Unified Modeling Language activity diagrams are used to identify megaprocesses at regional level and the customer behavior model graph (CBMG) to describe the states transitions of the processes. CBMG is annotated with: 1) the reliability of each component state and 2) the transition probabilities between states within the scope of the life cycle of the process. A stochastic reliability model (Markov model) is applied to predict the reliability of the business process as well as to identify the critical states and compare them with other processes to reveal the most critical ones. The ultimate benefit of the applied methodology is the design of more reliable components in an RHN system. The innovation of the approach of reliability modeling lies with the analysis of severity classes of failures and the application of stochastic modeling using discrete-time Markov chain in RHNs.

  5. APPROPRIATE ALLOCATION OF CONTINGENCY USING RISK ANALYSIS METHODOLOGY

    Directory of Open Access Journals (Sweden)

    Andi Andi

    2004-01-01

    Full Text Available Many cost overruns in the world of construction are attributable to either unforeseen events or foreseen events for which uncertainty was not appropriately accommodated. It is argued that a significant improvement to project management performance may result from greater attention to the process of analyzing project risks. The objective of this paper is to propose a risk analysis methodology for appropriate allocation of contingency in project cost estimation. In the first step, project risks will be identified. Influence diagramming technique is employed to identify and to show how the risks affect the project cost elements and also the relationships among the risks themselves. The second step is to assess the project costs with regards to the risks under consideration. Using a linguistic approach, the degree of uncertainty of identified project risks is assessed and quantified. The problem of dependency between risks is taken into consideration during this analysis. For the final step, as the main purpose of this paper, a method for allocating appropriate contingency is presented. Two types of contingencies, i.e. project contingency and management reserve are proposed to accommodate the risks. An illustrative example is presented at the end to show the application of the methodology.

  6. Contemporary Impact Analysis Methodology for Planetary Sample Return Missions

    Science.gov (United States)

    Perino, Scott V.; Bayandor, Javid; Samareh, Jamshid A.; Armand, Sasan C.

    2015-01-01

    Development of an Earth entry vehicle and the methodology created to evaluate the vehicle's impact landing response when returning to Earth is reported. NASA's future Mars Sample Return Mission requires a robust vehicle to return Martian samples back to Earth for analysis. The Earth entry vehicle is a proposed solution to this Mars mission requirement. During Earth reentry, the vehicle slows within the atmosphere and then impacts the ground at its terminal velocity. To protect the Martian samples, a spherical energy absorber called an impact sphere is under development. The impact sphere is composed of hybrid composite and crushable foam elements that endure large plastic deformations during impact and cause a highly nonlinear vehicle response. The developed analysis methodology captures a range of complex structural interactions and much of the failure physics that occurs during impact. Numerical models were created and benchmarked against experimental tests conducted at NASA Langley Research Center. The postimpact structural damage assessment showed close correlation between simulation predictions and experimental results. Acceleration, velocity, displacement, damage modes, and failure mechanisms were all effectively captured. These investigations demonstrate that the Earth entry vehicle has great potential in facilitating future sample return missions.

  7. Simplifying documentation while approaching site closure: integrated health and safety plans as documented safety analysis

    International Nuclear Information System (INIS)

    Brown, Tulanda

    2003-01-01

    At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). By isolating any remediation activities that deal with Enriched Restricted Materials, the SBRs and PRs assure that the hazard categories of former nuclear facilities undergoing remediation remain less than Nuclear. These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 150 structures, including six major nuclear production plants. This paper presents the FCP method for maintaining safety basis documentation, using the D and D I-HASP as an example

  8. RISMC Toolkit and Methodology Research and Development Plan for External Hazards Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report includes the description and development plan for a Risk Informed Safety Margins Characterization (RISMC) toolkit and methodology that will evaluate multihazard risk in an integrated manner to support the operating nuclear fleet.

  9. RISMC Toolkit and Methodology Research and Development Plan for External Hazards Analysis

    International Nuclear Information System (INIS)

    Coleman, Justin Leigh

    2016-01-01

    This report includes the description and development plan for a Risk Informed Safety Margins Characterization (RISMC) toolkit and methodology that will evaluate multihazard risk in an integrated manner to support the operating nuclear fleet.

  10. CONTENT ANALYSIS IN PROJECT MANAGEMENT: PROPOSALOF A METHODOLOGICAL FRAMEWORK

    Directory of Open Access Journals (Sweden)

    Alessandro Prudêncio Lukosevicius

    2016-12-01

    Full Text Available Content analysis (CA is a popular approach among researchers from different areas, but incipient in project management (PM. However, the volume of usage apparently does not translate into application quality. The method receives constant criticism about the scientific rigor adopted, especially when led by junior researchers. This article proposes a methodological framework for CA and investigate the use of CA in PM research. To accomplish this goal, literature systematic review is conjugated with CA related to 23 articles from EBSCO base in the last 20 years (1996 – 2016. The findings showed that the proposed framework can help researchers better apply the CA and suggests that the use of the method in terms of quantity and quality in PM research should be expanded. In addition to the framework, another contribution of this research is an analysis of the use of CA in PM in the last 20 years.

  11. Overview of the Preliminary Safety Analysis of the National Ignition Facility

    Science.gov (United States)

    Brereton, S.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.; Yatabe, J.

    1997-06-01

    The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF.

  12. Concept analysis of safety climate in healthcare providers.

    Science.gov (United States)

    Lin, Ying-Siou; Lin, Yen-Chun; Lou, Meei-Fang

    2017-06-01

    To report an analysis of the concept of safety climate in healthcare providers. Compliance with safe work practices is essential to patient safety and care outcomes. Analysing the concept of safety climate from the perspective of healthcare providers could improve understanding of the correlations between safety climate and healthcare provider compliance with safe work practices, thus enhancing quality of patient care. Concept analysis. The electronic databases of CINAHL, MEDLINE, PubMed and Web of Science were searched for literature published between 1995-2015. Searches used the keywords 'safety climate' or 'safety culture' with 'hospital' or 'healthcare'. The concept analysis method of Walker and Avant analysed safety climate from the perspective of healthcare providers. Three attributes defined how healthcare providers define safety climate: (1) creation of safe working environment by senior management in healthcare organisations; (2) shared perception of healthcare providers about safety of their work environment; and (3) the effective dissemination of safety information. Antecedents included the characteristics of healthcare providers and healthcare organisations as a whole, and the types of work in which they are engaged. Consequences consisted of safety performance and safety outcomes. Most studies developed and assessed the survey tools of safety climate or safety culture, with a minority consisting of interventional measures for improving safety climate. More prospective studies are needed to create interventional measures for improving safety climate of healthcare providers. This study is provided as a reference for use in developing multidimensional safety climate assessment tools and interventional measures. The values healthcare teams emphasise with regard to safety can serve to improve safety performance. Having an understanding of the concept of and interventional measures for safety climate allows healthcare providers to ensure the safety of their

  13. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  14. Application of best estimate plus uncertainty in review of research reactor safety analysis

    Directory of Open Access Journals (Sweden)

    Adu Simon

    2015-01-01

    Full Text Available To construct and operate a nuclear research reactor, the licensee is required to obtain the authorization from the regulatory body. One of the tasks of the regulatory authority is to verify that the safety analysis fulfils safety requirements. Historically, the compliance with safety requirements was assessed using a deterministic approach and conservative assumptions. This provides sufficient safety margins with respect to the licensing limits on boundary and operational conditions. Conservative assumptions were introduced into safety analysis to account for the uncertainty associated with lack of knowledge. With the introduction of best estimate computational tools, safety analyses are usually carried out using the best estimate approach. Results of such analyses can be accepted by the regulatory authority only if appropriate uncertainty evaluation is carried out. Best estimate computer codes are capable of providing more realistic information on the status of the plant, allowing the prediction of real safety margins. The best estimate plus uncertainty approach has proven to be reliable and viable of supplying realistic results if all conditions are carefully followed. This paper, therefore, presents this concept and its possible application to research reactor safety analysis. The aim of the paper is to investigate the unprotected loss-of-flow transients "core blockage" of a miniature neutron source research reactor by applying best estimate plus uncertainty methodology. The results of our calculations show that the temperatures in the core are within the safety limits and do not pose any significant threat to the reactor, as far as the melting of the cladding is concerned. The work also discusses the methodology of the best estimate plus uncertainty approach when applied to the safety analysis of research reactors for licensing purposes.

  15. Safety and safety analysis. From CP1 to Fukushima

    Energy Technology Data Exchange (ETDEWEB)

    Yadigaroglu, George [ASCOMP GmbH, Zurich (Switzerland)

    2012-02-15

    The safety of nuclear installations has been a serious concern starting from the days of infancy of this technology. When Fermi and co-workers built the first nuclear reactor in 1941, the Chicago Pile-1 or CP1 at the University of Chicago, some basic safety principles still in use today were already part of this very simple experiment. During the fast-growth period in the 1960ies, a number of NPP systems were conceived, tested and some of them built, mainly in the US and in the Soviet Union, but also in the UK, in France and in Canada, before just a handful of nuclear systems dominated: the LWRs conquered some 3 quarters of the world market and their dominance continues till today. The fission process has been amazingly well ''designed'' by nature: a remarkably simple to produce, self-sustained reaction that can be easily controlled, modulated and adjusted by a variety of available materials. Fission leads to large release of energy that can be easily collected and transformed into useful work. The process has only a major drawback, the inexorable production and accumulation in the core of the radioactive fission products that also produce decay heat. Criticality considerations put apart, the major goal of reactor safety is the confinement and cooling of these fission products. Although safety has been a major concern from the very first nuclear developments, feedback and actions following incidents and accidents have contributed to continuous enhancements. In particular, the three major nuclear accidents, TMI, Chernobyl and Fukushima had or will hopefully have in the future major impacts on safety improvements. Lessons learned from TMI have greatly enhanced the safety of LWRs, while Chernobyl triggered a number of radio-ecology studies and improved the readiness for radiological crisis management. It is hoped that Fukushima will be the trigger for much stronger international oversight and harmonization of safety practices, something that has

  16. Short course on system safety analysis

    International Nuclear Information System (INIS)

    Sudmann, R.H.

    1992-01-01

    This course provides and introduction to methods generally used in safety analysis and accident investigation. It is a non-mathematical approach, directed toward a casual user. The participant will learn techniques allowing them to dissect a system or incident in order identify real or potential safety problems. These techniques will be applied to analyze events which have occurred within DOE facilities. As a manager or staff person with general oversight responsibilities, the participant should gain an awareness of the big picture and not just ''dig for facts.'' This can be accomplished by being alert and responsive to the atmosphere and condition of the plant; mood and impression of the worker and the behavioral climate. The techniques taught in the course can be used to identify critical areas or indicators. These indicators will signal problems before the ''facts'' will. Analysis techniques taught are used to gauge the breadth of the ''forest'' and not necessarily to identify the trees. For this course includes a technical background with experience in a chemical processing operations and a knowledge of basic chemistry and engineering is desirable. The course should help in a present or future assignment in an oversight role

  17. 242-A evaporator safety analysis report

    International Nuclear Information System (INIS)

    CAMPBELL, T.A.

    1999-01-01

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR

  18. 242-A evaporator safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    CAMPBELL, T.A.

    1999-05-17

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR.

  19. Use of safety analysis to site comfirmation procedure in case of hard rock repository

    International Nuclear Information System (INIS)

    Peltonen, E.K.

    1984-02-01

    The role of safety analysis in a confirmation procedure of a candidate disposal site of radioactive wastes is discussed. Items dealt with include principle reasons and practical goals of the use of safety analysis, methodology of safety analysis and assessment, as well as usefulness and adequacy of the present safety analysis. Safety analysis is a tool, which enables one to estimate quantitatively the possible radiological impacts from the disposal. The results can be compared with the criteria and the suitability conclusions drawn. Because of its systems analytical nature safety analysis is an effective method to reveal, what are the most important factors of the disposal system and the most critical site characteristics inside the lumped parameters often provided by the experimental site investigation methods. Furthermore it gives information on the accuracy needs of different site properties. This can be utilized to judge whether the quality and quantity of the measurements for the characterization are sufficient as well as to guide the further site investigations. A more practical discussion regarding the applicability of the use of safety analysis is presented by an example concerning the assessment of a Finnish candidate site for low- and intermediate-level radioactive waste repository. (author)

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

    International Nuclear Information System (INIS)

    1994-01-01

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

  1. Motorcoach and school bus fire safety analysis : technology brief.

    Science.gov (United States)

    2016-11-01

    In 2009, the Federal Motor Carrier Safety Administration (FMCSA) published findings from a study entitled Motorcoach Fire Safety Analysis. The objective of this study was to gather and analyze information regarding the causes, frequency, and se...

  2. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  3. The verification methodologies for a software modeling of Engineered Safety Features- Component Control System (ESF-CCS)

    International Nuclear Information System (INIS)

    Lee, Young-Jun; Cheon, Se-Woo; Cha, Kyung-Ho; Park, Gee-Yong; Kwon, Kee-Choon

    2007-01-01

    The safety of a software is not guaranteed through a simple testing of the software. The testing reviews only the static functions of a software. The behavior, dynamic state of a software is not reviewed by a software testing. The Ariane5 rocket accident and the failure of the Virtual Case File Project are determined by a software fault. Although this software was tested thoroughly, the potential errors existed internally. There are a lot of methods to solve these problems. One of the methods is a formal methodology. It describes the software requirements as a formal specification during a software life cycle and verifies a specified design. This paper suggests the methods which verify the design to be described as a formal specification. We adapt these methods to the software of a ESF-CCS (Engineered Safety Features-Component Control System) and use the SCADE (Safety Critical Application Development Environment) tool for adopting the suggested verification methods

  4. Analysis of the low-altitude proton flux asymmetry: methodology

    CERN Document Server

    Kruglanski, M

    1999-01-01

    Existing East-West asymmetry models of the trapped proton fluxes at low altitudes depend on the local magnetic dip angle and a density scale height derived from atmospheric models. We propose an alternative approach which maps the directional flux over a drift shell (B sub m , L) in terms of the local pitch and azimuthal angles alpha and beta, where beta is defined in the local mirror plane as the angle between the proton arrival direction and the surface normal to the drift shell. This approach has the advantage that it only depends on drift shell parameters and does not involve an atmosphere model. A semi-empirical model based on the new methodology is able to reproduce the angular distribution of a set of SAMPEX/PET proton flux measurements. Guidelines are proposed for spacecraft missions and data analysis procedures that are intended to be used for the building of new trapped radiation environment models.

  5. Computational methodology for ChIP-seq analysis

    Science.gov (United States)

    Shin, Hyunjin; Liu, Tao; Duan, Xikun; Zhang, Yong; Liu, X. Shirley

    2015-01-01

    Chromatin immunoprecipitation coupled with massive parallel sequencing (ChIP-seq) is a powerful technology to identify the genome-wide locations of DNA binding proteins such as transcription factors or modified histones. As more and more experimental laboratories are adopting ChIP-seq to unravel the transcriptional and epigenetic regulatory mechanisms, computational analyses of ChIP-seq also become increasingly comprehensive and sophisticated. In this article, we review current computational methodology for ChIP-seq analysis, recommend useful algorithms and workflows, and introduce quality control measures at different analytical steps. We also discuss how ChIP-seq could be integrated with other types of genomic assays, such as gene expression profiling and genome-wide association studies, to provide a more comprehensive view of gene regulatory mechanisms in important physiological and pathological processes. PMID:25741452

  6. Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

    International Nuclear Information System (INIS)

    Hayakawa, Masahiko

    1981-01-01

    The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

  7. Methodology of analysis sustainable development of Ukraine by using the theory fuzzy logic

    Directory of Open Access Journals (Sweden)

    Methodology of analysis sustainable development of Ukraine by using the theory fuzzy logic

    2016-02-01

    Full Text Available Article objective is analysis of the theoretical and methodological aspects for the assessment of sustainable development in times of crisis. The methodical approach to the analysis of sustainable development territory taking into account the assessment of the level of economic security has been proposed. A necessity of development of the complex methodical approach to the accounting of the indeterminacy properties and multicriterial in the tasks to provide economic safety on the basis of using the fuzzy logic theory (or the fuzzy sets theory was proved. The results of using the method of fuzzy sets of during the 2002-2012 years the dynamics of changes dynamics of sustainable development in Ukraine were presented.

  8. Safety analysis of the VLJ repository

    International Nuclear Information System (INIS)

    Vieno, T.; Nordman, H.

    1991-05-01

    The VLJ repository is an underground disposal facility for the low and medium level waste generated at the Olkiluoto nuclear power plant. The repository is located within 1 km from TVO I and TVO II (2 x 710 MWe) BWR's on the Olkiluoto island at the west coast of Finland. It contains two rock silos excavated at the depth of 60...100 meters in the bedrock. Low level waste will be disposed of in a shotcreted rock silo. For bituminized medium level waste, a separate silo of reinforced concrete has been built inside the shotcreted rock silo. The post-closure safety analysis has been done for the Final Safety Analysis Report (FSAR) of the VLJ repository. In addition to the normal evolution scenario, several disturbed evolution and accident scenarios have been analysed. In the reference scenario, radio-nuclides are assumed to be released from the bituminized waste within 500 years, the concrete silo is assumed to gradually disintegrate and finally to collapse at 5 000 years, all concrete in the silo is assumed to be also chemically depleted within 6 000 years, and all the seals of the repository are assumed to deteriorate within 12 000 years. The ability of alone natural barriers to restrict the release of radionuclides into the biosphere has been evaluated by means of scenarios where the degradation of engineered barriers has been assumed to take place at a still faster rate. In one of the disturbed evolution scenarios it has been assumed that the concrete silo for medium level waste is severely impaired immediately after sealing of the repository. Effects of gas generation and consequences of human intrusion have been evaluated, too. The results of the safety analysis show that radiation doses of any significance are caused only if a well is bored in the vicinity of the repository or if the groundwater discharge spot is inhabited and used for cultivation. In the reference scenario the maximum expectation value of the individual dose rate is 0.3 mSv/a

  9. A 'Toolbox' Equivalent Process for Safety Analysis Software

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Eng, Tony

    2004-01-01

    Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (Quality Assurance for Safety-Related Software) identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls that prevent or mitigate potential accidents. The development and maintenance of a collection, or 'toolbox', of multiple-site use, standard solution, Software Quality Assurance (SQA)-compliant safety software is one of the major improvements identified in the associated DOE Implementation Plan (IP). The DOE safety analysis toolbox will contain a set of appropriately quality-assured, configuration-controlled, safety analysis codes, recognized for DOE-broad, safety basis applications. Currently, six widely applied safety analysis computer codes have been designated for toolbox consideration. While the toolbox concept considerably reduces SQA burdens among DOE users of these codes, many users of unique, single-purpose, or single-site software may still have sufficient technical justification to continue use of their computer code of choice, but are thwarted by the multiple-site condition on toolbox candidate software. The process discussed here provides a roadmap for an equivalency argument, i.e., establishing satisfactory SQA credentials for single-site software that can be deemed ''toolbox-equivalent''. The process is based on the model established to meet IP Commitment 4.2.1.2: Establish SQA criteria for the safety analysis ''toolbox'' codes. Implementing criteria that establish the set of prescriptive SQA requirements are based on implementation plan/procedures from the Savannah River Site, also incorporating aspects of those from the Waste Isolation Pilot Plant (SNL component) and the Yucca Mountain Project. The major requirements are met with evidence of a software quality assurance plan, software requirements and design documentation, user's instructions, test report, a

  10. Current status of safety analysis report for ANPP

    International Nuclear Information System (INIS)

    Amirjanyan, A.

    1999-01-01

    Current situation concerning Armenian NPP safety analysis report is considered within the frame of accepted safety practice. Licensing procedure is being developed. Technical support group was established in the Armenian Nuclear Regulatory Authority (ANRA). The task of the group is to study modern methods of NPP in depth safety analysis for technical assistance for the ANRA, and perform independent safety assessments. ANRA will be obliged to demand assistance from various foreign organisations for preparation of different parts of the Safety Analysis Report like determination though certain parts can be prepared in Armenia

  11. Systematic safety analysis of old nuclear power plants

    International Nuclear Information System (INIS)

    Dredemis, G.

    1985-11-01

    A program of systematic safety analysis of old nuclear power plants has been engaged by French safety authorities. Beyond the reshaping of safety documents (safety reports, general rules of operation, incidental and accidental procedures, internal emergency plan and manual of quality organization), this examination consisted of an analysis of the operation experience of circuits frequently actuated and a systematic analysis of safety circuits. This paper is based on the presentation of the exercise carried out at the Ardennes nuclear power plant operating for 15 years. This paper reviews also the main studies and modifications engaged on this power plant [fr

  12. 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

  13. The safety case for a HLW repository in Opalinus clay: aims, methodology, first results

    International Nuclear Information System (INIS)

    Zuidema, Piet

    2002-01-01

    Piet Zuidema (Nagra, Switzerland) described the development of the safety case for a high level waste repository in Opalinus clay in which canisters would be placed in large vaults. The current phase of work was concerned with demonstrating the feasibility of the disposal concept. The Safety Case is taken to mean a set of arguments to support a statement that the proposed facility will meet relevant safety criteria and will include arguments giving the basis for confidence that those arguments are correct and properly taking account of uncertainties. The safety strategy was concerned both with the inherent robustness of the disposal concept and the adequacy of the assessment capability. As regards the former, the arguments being advanced were primarily qualitative. Key issues in terms of the documentation of the Safety Case were traceability and transparency of information, including how to ensure that key arguments did not become obscured because of the need to make available very large quantities of information

  14. Methodology for Modeling and Analysis of Business Processes (MMABP

    Directory of Open Access Journals (Sweden)

    Vaclav Repa

    2015-10-01

    Full Text Available This paper introduces the methodology for modeling business processes. Creation of the methodology is described in terms of the Design Science Method. Firstly, the gap in contemporary Business Process Modeling approaches is identified and general modeling principles which can fill the gap are discussed. The way which these principles have been implemented in the main features of created methodology is described. Most critical identified points of the business process modeling are process states, process hierarchy and the granularity of process description. The methodology has been evaluated by use in the real project. Using the examples from this project the main methodology features are explained together with the significant problems which have been met during the project. Concluding from these problems together with the results of the methodology evaluation the needed future development of the methodology is outlined.

  15. Methodology for the analysis and retirement of assets: Power transformers

    Directory of Open Access Journals (Sweden)

    Gustavo Adolfo Gómez-Ramírez

    2015-09-01

    Full Text Available The following article consists of the development of a methodology of the area of the engineering of high voltage for the analysis and retirement of repaired power transformers, based on engineering criteria, in order to establish a correlation between the conditions of the transformer from several points of view: electrical, mechanical, dielectrics and thermal. Realizing an analysis of the condition of the question, there are two situations of great transcendency, first one, the international procedure are a “guide” for the acceptance of new transformers, by what they cannot be applied to the letter for repaired transformers, due to the process itself of degradation that the transformer has suffered with to happen of the years and all the factors that they were carrying to a possible repair. Second one, with base in the most recent technical literature, there have been analyzed articles, which analyze the oil dielectrics and the paper, which correlations are established between the quality of the insulating paper and the furan concentrations in the oils. To finish, great part of the investigation till now realized, it has focused in the analysis of the transformer, from the condition of the dielectric oil, so in most cases, there is not had the possibility of realizing a forensic engineering inside the transformer in operation and of being able this way, of analyzing the components of design who can compromise the integrity and operability of the same one.

  16. Transuranium analysis methodologies for biological and environmental samples

    International Nuclear Information System (INIS)

    Wessman, R.A.; Lee, K.D.; Curry, B.; Leventhal, L.

    1978-01-01

    Analytical procedures for the most abundant transuranium nuclides in the environment (i.e., plutonium and, to a lesser extent, americium) are available. There is a lack of procedures for doing sequential analysis for Np, Pu, Am, and Cm in environmental samples, primarily because of current emphasis on Pu and Am. Reprocessing requirements and waste disposal connected with the fuel cycle indicate that neptunium and curium must be considered in environmental radioactive assessments. Therefore it was necessary to develop procedures that determine all four of these radionuclides in the environment. The state of the art of transuranium analysis methodology as applied to environmental samples is discussed relative to different sample sources, such as soil, vegetation, air, water, and animals. Isotope-dilution analysis with 243 Am ( 239 Np) and 236 Pu or 242 Pu radionuclide tracers is used. Americium and curium are analyzed as a group, with 243 Am as the tracer. Sequential extraction procedures employing bis(2-ethyl-hexyl)orthophosphoric acid (HDEHP) were found to result in lower yields and higher Am--Cm fractionation than ion-exchange methods

  17. New Methodology of Block Cipher Analysis Using Chaos Game

    Directory of Open Access Journals (Sweden)

    Budi Sulistyo

    2014-11-01

    Full Text Available Block cipher analysis  covers randomness analysis and cryptanalysis. This paper proposes a new method potentially used for randomness analysis and cryptanalysis. The method uses true random sequence  concept as a reference for measuring randomness level of a random sequence. By using this concept, this paper  defines  bias  which represents  violation  of  a  random  sequence  from  true random sequence. In this paper, block cipher   is treated as a mapping function of a discrete time dynamical system. The dynamical system framework is used to make  the  application  of  various  analysis  techniques  developed  in  dynamical system  field  becomes  possible.  There  are three  main parts of  the methodology presented  in  this  paper:  the  dynamical  system  framework  for  block  cipher analysis, a  new chaos game  scheme and an extended measure  concept related to chaos game and fractal analysis. This paper also presents the general procedures of the proposed method, which includes: symbolic dynamic analysis of discr ete dynamical system whose block cipher as its mapping function, random sequence construction,  the  random  sequence  usage  as  input  of  a  chaos  game  scheme, output  measurement  of  chaos  game  scheme  using  extended  measure  concept, analysis  the  result  of  the  measurement.  The  analysis  process  and  of  a  specific real or sample block cipher and the analysis result are beyond the scope of this paper.

  18. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    Eisawy, E.A.; Sallam, H.

    2012-01-01

    The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

  19. Methodology and boundary conditions applied to the analysis on internal flooding for Kozloduy NPP units 5 and 6

    International Nuclear Information System (INIS)

    Demireva, E.; Goranov, S.; Horstmann, R.

    2004-01-01

    Within the Modernization Program of Units 5 and 6 of Kozloduy NPP a comprehensive analysis of internal flooding has been carried out for the reactor building outside the containment and for the turbine hall by FRAMATOME ANP and ENPRO Consult. The objective of this presentation is to provide information on the applied methodology and boundary conditions. A separate report called 'Methodology and boundary conditions' has been elaborated in order to provide the fundament for the study. The methodology report provides definitions and advice for the following topics: scope of the study; safety objectives; basic assumptions and postulates (plant conditions, grace periods for manual actions, single failure postulate, etc.); sources of flooding (postulated piping leaks and ruptures, malfunctions and personnel error); main activities of the flooding analysis; study conclusions and suggestions of remedial measures. (authors)

  20. ARIES-RS safety design and analysis

    International Nuclear Information System (INIS)

    Steiner, D.; El-Guebaly, L.; Herring, S.; Khater, H.; Mogahed, E.; Thayer, R.; Tillack, M.S.

    1997-01-01

    The ARIES-RS safety design and analysis focused on achieving two objectives: (1) The avoidance of sheltering or evacuation in the event of an accident; and (2) the generation of only low-level waste, no greater than Class C. The ARIES-RS baseline design employs V-4Cr-4Ti as the blanket structural material and a low activation ferritic steel in the reflector and shield. In the event of a LOCA, the baseline design first wall maximum temperature falls in the range of 1100-1200 C. For this temperature range, the hazard assessment indicates that the dose at the site boundary will be less than 1 rem per year. Thus, no sheltering or evacuation would be required in the event of a LOCA. Although the baseline design satisfies the first safety objective noted above, a first wall maximum temperature of ∝1100-1200 C would likely compromise the integrity of the vanadium blanket structure and would require blanket replacement following such a temperature excursion. To avoid this situation, a modified blanket design incorporating supplemental heat removal is also proposed. Preliminary analysis of this modified design suggests that the first wall maximum temperature can be kept below the temperature range of concern, ∝1000-1100 C, in the event of a LOCA. When the ferritic steel used in the reflector and shield is one reduced in Ir and Ag impurities, all in-vessel components qualify for near-surface shallow land burial as Class C low-level waste. (orig.)

  1. A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cowherd, Wilson M.; Nielsen, Joseph W.; Choe, Dong O.

    2016-11-01

    In support of experiments in the ATR, a new methodology was devised for loading the ATR Driver Core. This methodology will replace the existing methodology used by the INL Neutronic Analysis group to analyze experiments. Studied in this paper was the as-run analysis for ATR Cycle 152B, specifically comparing measured lobe powers and eigenvalue calculations.

  2. Documented Safety Analysis for the Waste Storage Facilities March 2010

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D T

    2010-03-05

    This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  3. Documented Safety Analysis for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-06-16

    This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  4. Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Cai Jianping; Shen Sen; Barkman, N.

    1999-01-01

    The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied

  5. Economic consideration of nuclear safety and cost benefit analysis in nuclear safety regulation

    International Nuclear Information System (INIS)

    Choi, Y. S.; Choi, K. S.; Choi, K. W.; Song, I. J.; Park, D. K.

    2001-01-01

    For the optimization of nuclear safety regulation, understanding of economic aspects of it becomes increasingly important together with the technical approach used so far to secure nuclear safety. Relevant economic theories on private and public goods were reviewed to re-illuminate nuclear safety from the economic perspective. The characteristics of nuclear safety as a public good was reviewed and discussed in comparison with the car safety as a private safety good. It was shown that the change of social welfare resulted from the policy change induced can be calculated by the summation of compensating variation(CV) of individuals. It was shown that the value of nuclear safety could be determined in monetary term by this approach. The theoretical background and history of cost benefit analysis of nuclear safety regulation were presented and topics for future study were suggested

  6. Supporting Space Systems Design via Systems Dependency Analysis Methodology

    Science.gov (United States)

    Guariniello, Cesare

    assess the behavior of each system based on its internal status and on the topology of its dependencies on systems connected to it. Designers and decision makers can therefore quickly analyze and explore the behavior of complex systems and evaluate different architectures under various working conditions. The methods support educated decision making both in the design and in the update process of systems architecture, reducing the need to execute extensive simulations. In particular, in the phase of concept generation and selection, the information given by the methods can be used to identify promising architectures to be further tested and improved, while discarding architectures that do not show the required level of global features. The methods, when used in conjunction with appropriate metrics, also allow for improved reliability and risk analysis, as well as for automatic scheduling and re-scheduling based on the features of the dependencies and on the accepted level of risk. This dissertation illustrates the use of the two methods in sample aerospace applications, both in the operational and in the developmental domain. The applications show how to use the developed methodology to evaluate the impact of failures, assess the criticality of systems, quantify metrics of interest, quantify the impact of delays, support informed decision making when scheduling the development of systems and evaluate the achievement of partial capabilities. A larger, well-framed case study illustrates how the Systems Operational Dependency Analysis method and the Systems Developmental Dependency Analysis method can support analysis and decision making, at the mid and high level, in the design process of architectures for the exploration of Mars. The case study also shows how the methods do not replace the classical systems engineering methodologies, but support and improve them.

  7. Failure mode effect analysis and fault tree analysis as a combined methodology in risk management

    Science.gov (United States)

    Wessiani, N. A.; Yoshio, F.

    2018-04-01

    There have been many studies reported the implementation of Failure Mode Effect Analysis (FMEA) and Fault Tree Analysis (FTA) as a method in risk management. However, most of the studies usually only choose one of these two methods in their risk management methodology. On the other side, combining these two methods will reduce the drawbacks of each methods when implemented separately. This paper aims to combine the methodology of FMEA and FTA in assessing risk. A case study in the metal company will illustrate how this methodology can be implemented. In the case study, this combined methodology will assess the internal risks that occur in the production process. Further, those internal risks should be mitigated based on their level of risks.

  8. 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

  9. A Predictive Safety Management System Software Package Based on the Continuous Hazard Tracking and Failure Prediction Methodology

    Science.gov (United States)

    Quintana, Rolando

    2003-01-01

    The goal of this research was to integrate a previously validated and reliable safety model, called Continuous Hazard Tracking and Failure Prediction Methodology (CHTFPM), into a software application. This led to the development of a safety management information system (PSMIS). This means that the theory or principles of the CHTFPM were incorporated in a software package; hence, the PSMIS is referred to as CHTFPM management information system (CHTFPM MIS). The purpose of the PSMIS is to reduce the time and manpower required to perform predictive studies as well as to facilitate the handling of enormous quantities of information in this type of studies. The CHTFPM theory encompasses the philosophy of looking at the concept of safety engineering from a new perspective: from a proactive, than a reactive, viewpoint. That is, corrective measures are taken before a problem instead of after it happened. That is why the CHTFPM is a predictive safety because it foresees or anticipates accidents, system failures and unacceptable risks; therefore, corrective action can be taken in order to prevent all these unwanted issues. Consequently, safety and reliability of systems or processes can be further improved by taking proactive and timely corrective actions.

  10. The analysis of RWAP(Rod Withdrawal at Power) using the KEPRI methodology

    International Nuclear Information System (INIS)

    Yang, C. K.; Kim, Y. H.

    2001-01-01

    KEPRI developed new methodology which was based on RASP(Reactor Analysis Support Package). In this paper, The analysis of RWAP(Rod Withdrawal at Power) accident which can result in reactivity and power distribution anomaly was performed using the KEPRI methodology. The calculation describes RWAP transient and documents the analysis, including the computer code modeling assumptions and input parameters used in the analysis. To validity for the new methodology, the result of calculation was compared with FSAR. As compared with FSAR, result of the calculation using the KEPRI Methodology is similar to FSAR's. And result of the sensitivity of postulated parameters were similar to the existing methodology

  11. Reliability analysis for power supply system in a reprocessing facility based on GO methodology

    International Nuclear Information System (INIS)

    Wang Renze

    2014-01-01

    GO methodology was applied to analyze the reliability of power supply system in a typical reprocessing facility. Based on the fact that tie breakers are set in the system, tie breaker operator was defined. Then GO methodology modeling and quantitative analysis were performed sequently, minimal cut sets and average unavailability of the system were obtained. Parallel analysis between GO methodology and fault tree methodology was also performed. The results showed that setup of tie breakers was rational and necessary and that the modeling was much easier and the chart was much more succinct for GO methodology parallel with fault tree methodology to analyze the reliability of the power supply system. (author)

  12. Analysis of road safety management systems in Europe.

    NARCIS (Netherlands)

    Muhlrad, N. Vallet, G. Butler, I. Gitelman, V. Doveh, E. Dupont, E. Thomas, P. Talbot, R. Papadimitriou, E. Yannis, G. Persia, L. Giustiniani, G. Machata, K. & Bax, C.A.

    2014-01-01

    The objective of this paper is the analysis of road safety management in European countries and the identification of “good practice”. A road safety management investigation model was created, based on several “good practice” criteria. Road safety management systems have been thoroughly investigated

  13. Cost analysis in health centers using 'Step Down' methodology

    Directory of Open Access Journals (Sweden)

    Matejić S.

    2015-01-01

    Full Text Available Health care reform aims to improve Health System performance by achievement of one of the four objectives: reducing costs by increasing the efficiency of health care provision. Performance improvement implies acceptance of innovations in all Health care activities including health care financial management. Successful implementation of health care financing reform requires previous costs and activities analysis in health institutions. In the work we performed comparative analysis of the costs of 27 health institutions by applying innovative system for health care services costs analysis and control. Initialy spreadsheet system was made, by using internationaly recognised 'Step Down' methodology, for cost control and analisys in the hospitals and was adapted for Primary health care institutions. Results achieved: The dominant cost for employees salaries, on average around 80%, does not depend on the size of Primary health institution (Dom zdravlja; Significant differences in the percentage values of the cost of medicines, medical supplies, diagnostic services; There is an obvious difference percentage values of technical maintenance costs as a result of uneven percentage of the number of non-medical employees, differences in infrastructure organization, the difference in the condition and type of equipment, the difference in the type of space heating and type of fuel for heating, patients transportation obligations especialy of home treatment services and polyvalent patronage. There is a big difference in average cost per outpatient examination, as a consequence of uneven number of services performed, especialy in the dentistry services. There is a significant difference in the number of preventive health examinations performed which has a direct impact on the cost of these inspections. The main conclusion of the analysis done indicates that in the actual situation of disparities, in terms of costs, can joperdize implementation of Primary health

  14. FRACTAL ANALYSIS OF TRABECULAR BONE: A STANDARDISED METHODOLOGY

    Directory of Open Access Journals (Sweden)

    Ian Parkinson

    2011-05-01

    Full Text Available A standardised methodology for the fractal analysis of histological sections of trabecular bone has been established. A modified box counting method has been developed for use on a PC based image analyser (Quantimet 500MC, Leica Cambridge. The effect of image analyser settings, magnification, image orientation and threshold levels, was determined. Also, the range of scale over which trabecular bone is effectively fractal was determined and a method formulated to objectively calculate more than one fractal dimension from the modified Richardson plot. The results show that magnification, image orientation and threshold settings have little effect on the estimate of fractal dimension. Trabecular bone has a lower limit below which it is not fractal (λ<25 μm and the upper limit is 4250 μm. There are three distinct fractal dimensions for trabecular bone (sectional fractals, with magnitudes greater than 1.0 and less than 2.0. It has been shown that trabecular bone is effectively fractal over a defined range of scale. Also, within this range, there is more than 1 fractal dimension, describing spatial structural entities. Fractal analysis is a model independent method for describing a complex multifaceted structure, which can be adapted for the study of other biological systems. This may be at the cell, tissue or organ level and compliments conventional histomorphometric and stereological techniques.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  16. RiskSOAP: Introducing and applying a methodology of risk self-awareness in road tunnel safety.

    Science.gov (United States)

    Chatzimichailidou, Maria Mikela; Dokas, Ioannis M

    2016-05-01

    Complex socio-technical systems, such as road tunnels, can be designed and developed with more or less elements that can either positively or negatively affect the capability of their agents to recognise imminent threats or vulnerabilities that possibly lead to accidents. This capability is called risk Situation Awareness (SA) provision. Having as a motive the introduction of better tools for designing and developing systems that are self-aware of their vulnerabilities and react to prevent accidents and losses, this paper introduces the Risk Situation Awareness Provision (RiskSOAP) methodology to the field of road tunnel safety, as a means to measure this capability in this kind of systems. The main objective is to test the soundness and the applicability of RiskSOAP to infrastructure, which is advanced in terms of technology, human integration, and minimum number of safety requirements imposed by international bodies. RiskSOAP is applied to a specific road tunnel in Greece and the accompanying indicator is calculated twice, once for the tunnel design as defined by updated European safety standards and once for the 'as-is' tunnel composition, which complies with the necessary safety requirements, but calls for enhancing safety according to what EU and PIARC further suggest. The derived values indicate the extent to which each tunnel version is capable of comprehending its threats and vulnerabilities based on its elements. The former tunnel version seems to be more enhanced both in terms of it risk awareness capability and safety as well. Another interesting finding is that despite the advanced tunnel safety specifications, there is still room for enriching the safe design and maintenance of the road tunnel. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. Preparing a Safety Analysis Report using the building block approach

    International Nuclear Information System (INIS)

    Herrington, C.C.

    1990-01-01

    The credibility of the applicant in a licensing proceeding is severely impacted by the quality of the license application, particularly the Safety Analysis Report. To ensure the highest possible credibility, the building block approach was devised to support the development of a quality Safety Analysis Report. The approach incorporates a comprehensive planning scheme that logically ties together all levels of the investigation and provides the direction necessary to prepare a superior Safety Analysis Report

  18. Design of a Construction Safety Training System using Contextual Design Methodology

    OpenAIRE

    Baldev, Darshan H.

    2006-01-01

    In the U.S., the majority of construction companies are small companies with 10 or fewer employees (BLS, 2004). The fatality rate in the construction industry is high, indicating a need for implementing safety training to a greater extent. This research addresses two main goals: to make recommendations and design a safety training system for small construction companies, and to use Contextual Design to design the training system. Contextual Design was developed by Holtzblatt (Beyer and Holtzb...

  19. Reliability Analysis for Safety Grade PLC(POSAFE-Q)

    International Nuclear Information System (INIS)

    Choi, Kyung Chul; Song, Seung Whan; Park, Gang Min; Hwang, Sung Jae

    2012-01-01

    Safety Grade PLC(Programmable Logic Controller), POSAFE-Q, was developed recently in accordance with nuclear regulatory and requirements. In this paper, describe reliability analysis for digital safety grade PLC (especially POSAFE-Q). Reliability analysis scope is Prediction, Calculation of MTBF (Mean Time Between Failure), FMEA (Failure Mode Effect Analysis), PFD (Probability of Failure on Demand). (author)

  20. SAFETY-BASED CAPACITY ANALYSIS FOR CHINESE HIGHWAYS

    Directory of Open Access Journals (Sweden)

    Ping YI, Ph.D.

    2004-01-01

    Full Text Available Many years of research have led to the development of theories and methodologies in roadway capacity analysis in the developed countries. However, those resources coexist with roadway design and traffic control practices in the local country, and cannot be simply transferred to China for applications. For example, the Highway Capacity Manual in the United State describes roadway capacity under ideal conditions and estimates practical capacities under prevailing conditions in the field. This capacity and the conditions for change are expected to be different on Chinese roadways as the local roadway design (lane width, curves and grades, vehicle size, and traffic mix are different. This research looks into an approach to the capacity issue different from the Highway Capacity Manual. According to the car-following principle, this paper first describes the safety criteria that affect traffic operations. Several speed schemes are subsequently discussed as they are affected by the maximum speed achievable under the local conditions. The study has shown that the effect of geometric and traffic conditions can be effectually reflected in the maximum speed adopted by the drivers. For most Chinese highways without a posted speed limit, the choice of speed by the drivers from the safety prospective is believed to have incorporated considerations of the practical driving conditions. Based on this, a condition for capacity calculation is obtained by comparing the desired vs. safety-based distance headways. The formulations of the model are mathematically sound and physically meaningful, and preliminary testing of the model is encouraging. Future research includes field data acquisition for calibration and adjustment, and model testing on Chinese highways.

  1. Health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California. Volume 9. Methodologies for review of the health and safety aspects of proposed nuclear, geothermal, and fossil-fuel sites and facilities

    International Nuclear Information System (INIS)

    Nero, A.V.; Quinby-Hunt, M.S.

    1977-01-01

    This report sets forth methodologies for review of the health and safety aspects of proposed nuclear, geothermal, and fossil-fuel sites and facilities for electric power generation. The review is divided into a Notice of Intention process and an Application for Certification process, in accordance with the structure to be used by the California Energy Resources Conservation and Development Commission, the first emphasizing site-specific considerations, the second examining the detailed facility design as well. The Notice of Intention review is divided into three possible stages: an examination of emissions and site characteristics, a basic impact analysis, and an assessment of public impacts. The Application for Certification review is divided into five possible stages: a review of the Notice of Intention treatment, review of the emission control equipment, review of the safety design, review of the general facility design, and an overall assessment of site and facility acceptability

  2. Safety Analysis for a Radioisotope Stirling Generator

    International Nuclear Information System (INIS)

    William D. Richins; Jeffrey M. Lacy; Stephen R. Novascone; Barbara H. Dolphin

    2007-01-01

    The Idaho National Laboratory INL is conducting safety analyses of Radioisotope Stirling Generators for the Department of Energy (NE-50) to support the use of these devices as terrestrial power sources. These systems are electrical power generators converting thermal energy from plutonium (238Pu) decay to electrical energy via a Stirling cycle generator. The design and function are similar to the RTG (Radioisotope Thermoelectric Generator) used in space missions since the early 1960's, with a more efficient Stirling cycle generator replacing the proven thermoelectric converter. The subject generator is the product of a collaborative effort by Lockheed Martin, Infinia, and the Glenn Research Center. This paper discusses the methods the INL is employing in the safety analysis effort, along with the software tools, lessons learned, and results. The overall goal of our safety analyses is to determine the probability of an accidental plutonium release over the life of the generator. Historical accident rates for various storage and transportation modes were investigated using event tree methods. Source terms were developed for these accidents including primarily impact, fire, and creep rupture. A negative result was defined as rupture of the tantalum alloy containment vessel surrounding the encapsulated plutonia pellet. Damage due to identified impact accidents was evaluated using non-linear finite element software tools. Material models, gathered from a wide variety of sources, included strain-rate and temperature dependencies on yield strength, strain hardening, and rupture. The overall simulation results predicted by our software tools will be validated by impact testing. Results from deterministic impact, fire, and creep rupture analyses were integrated into the probabilistic (Monte Carlo) risk assessment by correlation functions relating accident parameters to component damage. This approach presented challenges, which are addressed. Other significant issues

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  4. Advanced Crash Avoidance Technologies (ACAT) Program - Final Report of the Volvo-Ford-UMTRI Project: Safety Impact Methodology for Lane Departure Warning - Method Development and Estimation of Benefits

    Science.gov (United States)

    2010-10-01

    The Volvo-Ford-UMTRI project: Safety Impact Methodology (SIM) for Lane Departure Warning is part of the U.S. Department of Transportation's Advanced Crash Avoidance Technologies (ACAT) program. The project developed a basic analytical framework for e...

  5. Biosensors for functional food safety and analysis.

    Science.gov (United States)

    Lavecchia, Teresa; Tibuzzi, Arianna; Giardi, Maria Teresa

    2010-01-01

    The importance of safety and functionality analysis of foodstuffs and raw materials is supported by national legislations and European Union (EU) directives concerning not only the amount of residues of pollutants and pathogens but also the activity and content of food additives and the health claims stated on their labels. In addition, consumers' awareness of the impact of functional foods' on their well-being and their desire for daily healthcare without the intake pharmaceuticals has immensely in recent years. Within this picture, the availability of fast, reliable, low cost control systems to measure the content and the quality of food additives and nutrients with health claims becomes mandatory, to be used by producers, consumers and the governmental bodies in charge of the legal supervision of such matters. This review aims at describing the most important methods and tools used for food analysis, starting with the classical methods (e.g., gas-chromatography GC, high performance liquid chromatography HPLC) and moving to the use of biosensors-novel biological material-based equipments. Four types of bio-sensors, among others, the novel photosynthetic proteins-based devices which are more promising and common in food analysis applications, are reviewed. A particular highlight on biosensors for the emerging market of functional foods is given and the most widely applied functional components are reviewed with a comprehensive analysis of papers published in the last three years; this report discusses recent trends for sensitive, fast, repeatable and cheap measurements, focused on the detection of vitamins, folate (folic acid), zinc (Zn), iron (Fe), calcium (Ca), fatty acids (in particular Omega 3), phytosterols and phytochemicals. A final market overview emphasizes some practical aspects ofbiosensor applications.

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

    International Nuclear Information System (INIS)

    2011-01-01

    The IAEA's Statute authorizes the Agency to 'establish or adopt' standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

  7. An analysis of the traffic safety phenomenon.

    OpenAIRE

    Asmussen, E. & Kranenburg, A.

    1982-01-01

    The lack of traffic safety is a combination of the critical coincidence of circumstances in the traffic of incidents (near-accidents) and accidents with unwanted (permanent) consequences, such as fatalities, injured and disabled persons and material damage. This definition covers the whole of the critical coincidence of circumstances in traffic. In order to elucidate the phenomenon of traffic safety or the lack of traffic safety, accidents and incidents can be analysed.

  8. Qualitative Research for Patient Safety Using ICTs: Methodological Considerations in the Technological Age.

    Science.gov (United States)

    Yee, Kwang Chien; Wong, Ming Chao; Turner, Paul

    2017-01-01

    Considerable effort and resources have been dedicated to improving the quality and safety of patient care through health information systems, but there is still significant scope for improvement. One contributing factor to the lack of progress in patient safety improvement especially where technology has been deployed relates to an over-reliance on purely objective, quantitative, positivist research paradigms as the basis for generating and validating evidence of improvement. This paper argues the need for greater recognition and accommodation of evidence of improvement generated through more subjective, qualitative and pragmatic research paradigms to aid patient safety especially where technology is deployed. This paper discusses how acknowledging the role and value of more subjective ontologies and pragmatist epistemologies can support improvement science research. This paper illustrates some challenges and benefits from adopting qualitative research methods in patient safety improvement projects, particularly focusing challenges in the technological era. While adopting methods that can more readily capture, analyse and interpret direct user experiences, attitudes, insights and behaviours in their contextual settings, patient safety can be enhanced 'on the ground' and errors reduced and/or mitigated, challenges of using these methods with the younger "technologically-centred" healthcare professionals and patients needs to recognised.

  9. Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

    Directory of Open Access Journals (Sweden)

    Zongzhi Li

    2017-04-01

    Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.

  10. Modeling methodology for supply chain synthesis and disruption analysis

    Science.gov (United States)

    Wu, Teresa; Blackhurst, Jennifer

    2004-11-01

    The concept of an integrated or synthesized supply chain is a strategy for managing today's globalized and customer driven supply chains in order to better meet customer demands. Synthesizing individual entities into an integrated supply chain can be a challenging task due to a variety of factors including conflicting objectives, mismatched incentives and constraints of the individual entities. Furthermore, understanding the effects of disruptions occurring at any point in the system is difficult when working toward synthesizing supply chain operations. Therefore, the goal of this research is to present a modeling methodology to manage the synthesis of a supply chain by linking hierarchical levels of the system and to model and analyze disruptions in the integrated supply chain. The contribution of this research is threefold: (1) supply chain systems can be modeled hierarchically (2) the performance of synthesized supply chain system can be evaluated quantitatively (3) reachability analysis is used to evaluate the system performance and verify whether a specific state is reachable, allowing the user to understand the extent of effects of a disruption.

  11. General reliability and safety methodology and its application to wind energy conversion systems

    Energy Technology Data Exchange (ETDEWEB)

    Edesess, M.; McConnell, R. D.

    1979-09-01

    In conventional system reliability calculations, each component may be in the Operable state or the Under Repair state. These calculations derive system unavailability, or the probability of the system's being down for repairs. By introducing a third component state between Operable and Under Repair - namely, Defective, But Defect Undetected - the methods developed in this report enable system safety projections to be made in addition to availability projections. Also provided is a mechanism for computing the effect of inspection schedules on both safety and availability. A Reliability and Safety Program (RASP) is detailed which performs these computations and also calculates costs for system inspections and repairs. RASP is applied to a simplified wind energy conversion system example.

  12. 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

  13. Application of human reliability analysis methodology of second generation; Aplicacion de metodologia de analisis de confiabilidad humana de segunda generacion

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz S, T. de J.; Nelson E, P. F. [Facultad de Ingenieria, Departamento de Sistemas Energeticos, UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)], e-mail: trs@cie.unam.mx

    2009-10-15

    The human reliability analysis (HRA) is a very important part of probabilistic safety analysis. The main contribution of HRA in nuclear power plants is the identification and characterization of the issues that are brought together for an error occurring in the human tasks that occur under normal operation conditions and those made after abnormal event. Additionally, the analysis of various accidents in history, it was found that the human component has been a contributing factor in the cause. Because of need to understand the forms and probability of human error in the 60 decade begins with the collection of generic data that result in the development of the first generation of HRA methodologies. Subsequently develop methods to include in their models additional performance shaping factors and the interaction between them. So by the 90 mid, comes what is considered the second generation methodologies. Among these is the methodology A Technique for Human Event Analysis (ATHEANA). The application of this method in a generic human failure event, it is interesting because it includes in its modeling commission error, the additional deviations quantification to nominal scenario considered in the accident sequence of probabilistic safety analysis and, for this event the dependency actions evaluation. That is, the generic human failure event was required first independent evaluation of the two related human failure events . So the gathering of the new human error probabilities involves the nominal scenario quantification and cases of significant deviations considered by the potential impact on analyzed human failure events. Like probabilistic safety analysis, with the analysis of the sequences were extracted factors more specific with the highest contribution in the human error probabilities. (Author)

  14. A METHODOLOGICAL APPROACH TO THE STRATEGIC ANALYSIS OF FOOD SECURITY

    Directory of Open Access Journals (Sweden)

    Anastasiia Mostova

    2017-12-01

    Full Text Available The objective of present work is to substantiate the use of tools for strategic analysis in order to develop a strategy for the country’s food security under current conditions and to devise the author’s original technique to perform strategic analysis of food security using a SWOT-analysis. The methodology of the study. The article substantiates the need for strategic planning of food security. The author considers stages in strategic planning and explains the importance of the stage of strategic analysis of the country’s food security. It is proposed to apply a SWOT-analysis when running a strategic analysis of food security. The study is based on the system of indicators and characteristics of the country’s economy, agricultural sector, market trends, material-technical, financial, human resources, which are essential to obtain an objective assessment of the impact of trends and factors on food security, and in order to further develop the procedure for conducting a strategic analysis of the country’s food security. Results of the study. The procedure for strategic analysis of food security is developed based on the tool of a SWOT-analysis, which implies three stages: a strategic analysis of weaknesses and strengths, opportunities and threats; construction of the matrix of weaknesses and strengths, opportunities, and threats (SWOT-analysis matrix; formation of the food security strategy based on the SWOT-analysis matrix. A list of characteristics was compiled in order to conduct a strategic analysis of food security and to categorize them as strengths or weaknesses, threats, and opportunities. The characteristics are systemized into strategic groups: production, market; resources; consumption: this is necessary for the objective establishing of strategic directions, responsible performers, allocation of resources, and effective control, for the purpose of further development and implementation of the strategy. A strategic analysis

  15. Safety Analysis versus Type Inference with Partial Types

    DEFF Research Database (Denmark)

    Schwartzbach, Michael Ignatieff; Palsberg, Jens

    1992-01-01

    perspectives, however. Safety analysis is global in that it can only analyze a complete program. In contrast, type inference is local in that it can analyze pieces of a program in isolation. In this paper we prove that safety analysis is sound, relative to both a strict and a lazy operational semantics. We......Safety analysis is an algorithm for determining if a term in an untyped lambda calculus with constants is safe, i.e., if it does not cause an error during evaluation. This ambition is also shared by algorithms for type inference. Safety analysis and type inference are based on rather different...... also prove that safety analysis accepts strictly more safe lambda terms than does type inference for simple types. The latter result demonstrates that global program analysis can be more precise than local ones....

  16. Iterative Transport-Diffusion Methodology For LWR Core Analysis

    Science.gov (United States)

    Colameco, David; Ivanov, Boyan D.; Beacon, Daniel; Ivanov, Kostadin N.

    2014-06-01

    This paper presents an update on the development of an advanced methodology for core calculations that uses local heterogeneous solutions for on-the-fly nodal cross-section generation. The Iterative Transport-Diffusion Method is an embedded transport approach that is expected to provide results with near 3D transport accuracy for a fraction of the time required by a full 3D transport method. In this methodology, the infinite environment used for homogenized nodal cross-section generation is replaced with a simulated 3D environment of the diffusion calculation. This update focuses on burnup methodology, axial leakage and 3D modeling.

  17. Applying Qualitative Hazard Analysis to Support Quantitative Safety Analysis for Proposed Reduced Wake Separation Conops

    Science.gov (United States)

    Shortle, John F.; Allocco, Michael

    2005-01-01

    This paper describes a scenario-driven hazard analysis process to identify, eliminate, and control safety-related risks. Within this process, we develop selective criteria to determine the applicability of applying engineering modeling to hypothesized hazard scenarios. This provides a basis for evaluating and prioritizing the scenarios as candidates for further quantitative analysis. We have applied this methodology to proposed concepts of operations for reduced wake separation for closely spaced parallel runways. For arrivals, the process identified 43 core hazard scenarios. Of these, we classified 12 as appropriate for further quantitative modeling, 24 that should be mitigated through controls, recommendations, and / or procedures (that is, scenarios not appropriate for quantitative modeling), and 7 that have the lowest priority for further analysis.

  18. Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

    International Nuclear Information System (INIS)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-01-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  19. Safety Injection Tank Performance Analysis Using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Oan; Lee, Jeong Ik; Nietiadi Yohanes Setiawan [KAIST, Daejeon (Korea, Republic of); Addad Yacine [KUSTAR, Abu Dhabi (United Arab Emirates); Bang, Young Seok; Yoo, Seung Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    This may affect the core cooling capability and threaten the fuel integrity during LOCA situations. However, information on the nitrogen flow rate during discharge is very limited due to the associated experimental measurement difficulties, and these phenomena are hardly reflected in current 1D system codes. In the current study, a CFD analysis is presented which hopefully should allow obtaining a more realistic prediction of the SIT performance which can then be reflected on 1D system codes to simulate various accident scenarios. Current Computational Fluid Dynamics (CFD) calculations have had limited success in predicting the fluid flow accurately. This study aims to find a better CFD prediction and more accurate modeling to predict the system performance during accident scenarios. The safety injection tank with fluidic device was analyzed using commercial CFD. A fine resolution grid was used to capture the vortex of the fluidic device. The calculation so far has shown good consistency with the experiment. Calculation should complete by the conference date and will be thoroughly analyzed to be discussed. Once a detailed CFD computation is finished, a small-scale experiment will be conducted for the given conditions. Using the experimental results and the CFD model, physical models can be validated to give more reliable results. The data from CFD and experiments will provide a more accurate K-factor of the fluidic device which can later be applied in system code inputs.

  20. Considering new methodologies in strategies for safety assessment of foods and food ingredients

    NARCIS (Netherlands)

    Brett Jeffery; Gareth Edwards; Alan R. Boobis; Anne Constable; Bobbie Bradford; Jossie A. Garthoff; Cyrille Krul; Bas J. Blaauboer; Jeroen Schuermansk; Mardas Daneshian; Andrew Cockburn

    2016-01-01

    from the article Abstract Toxicology and safety assessment are changing and require new strategies for evaluating risk that are less depending on apical toxicity endpoints in animal models and relying more on knowledge of the mechanism of toxicity. This manuscript describes a number of

  1. A Monitoring System for the Safety of Building Structure Based on W2T Methodology

    NARCIS (Netherlands)

    Wang, Haiyuan; Huang, Zhisheng; Zhong, Ning; Huang, Jiajin; Han, Yuzhong; Zhang, Feng; Zhong, Ning; Ma, Jianhua; Liu, Jiming; Huang, Runhe; Tao, Xiaohui

    2016-01-01

    With the development of the Internet of things, monitoring systems for the safety of building structure (SBS) provide people with the important data about the main supporting points in the buildings. More and more data give the engineers an overload work problem, which can be solved by a systematic

  2. Drug research methodology. Volume 2, The identification of drugs of interest in highway safety

    Science.gov (United States)

    1980-03-01

    This report presents findings of a workshop on the identification of drugs that should be the focus of near-term highway safety research. Drugs of interest are those that have a potential to increase the likelihood of traffic crashes and their attend...

  3. Special characteristics of the safety analysis of HWRs

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    Two lectures are presented in this report. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor, and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (orig./RW)

  4. Analysis of the reliability of the active injection safety systems of Angra I

    International Nuclear Information System (INIS)

    Frutuoso e Melo, P.F.F.

    1981-01-01

    The reliability of the active emergency core cooling systems of Angra I nuclear power plant is evaluated. The fault tree analysis is employed. The unavailability of the above cited systems, is calculated. A parametric sensitivity analysis has been performed, due to the existing scattering in the failure and repair rate data of these system's components. The minimal cut sets were determined and, as a final step, a reliability importance analysis has been performed. This final step has required the development of a computer program. The methodology and data from the 'Reactor Safety Study' (Wash-1400) (in which the reliability of safety systems of a tipical PWR plant is calculated), is employed. The unavailability values for the safety systems analysed are too low, thus showing that in most cases the systems analysed are available to mitigate the effects of a loss-of-coolant accident. (Author) [pt

  5. Using functional analysis in archival appraisal a practical and effective alternative to traditional appraisal methodologies

    CERN Document Server

    Robyns, Marcus C

    2014-01-01

    In an age of scarcity and the challenge of electronic records, can archivists and records managers continue to rely upon traditional methodology essentially unchanged since the early 1950s? Using Functional Analysis in Archival Appraisal: A Practical and Effective Alternative to Traditional Appraisal Methodologies shows how archivists in other countries are already using functional analysis, which offers a better, more effective, and imminently more practical alternative to traditional appraisal methodologies that rely upon an analysis of the records themselves.

  6. Transit safety & security statistics & analysis 2002 annual report (formerly SAMIS)

    Science.gov (United States)

    2004-12-01

    The Transit Safety & Security Statistics & Analysis 2002 Annual Report (formerly SAMIS) is a compilation and analysis of mass transit accident, casualty, and crime statistics reported under the Federal Transit Administrations (FTAs) National Tr...

  7. Transit safety & security statistics & analysis 2003 annual report (formerly SAMIS)

    Science.gov (United States)

    2005-12-01

    The Transit Safety & Security Statistics & Analysis 2003 Annual Report (formerly SAMIS) is a compilation and analysis of mass transit accident, casualty, and crime statistics reported under the Federal Transit Administrations (FTAs) National Tr...

  8. Safety analysis report 231-Z Building

    Energy Technology Data Exchange (ETDEWEB)

    Powers, C.S.

    1989-03-01

    This report provides an intensive review of the nuclear safety of the operation of the 231-Z Building. For background information complete descriptions of the floor plan, building services, alarm systems, and glove box systems are included in this report. In addition, references are included to The Plutonium Laboratory Radiation Work Procedures, Safety Guides, 231-Z Operating Procedures Manual and Nuclear Materials accountability Procedures. Engineered and administrative features contribute to the overall safety of personnel, the building, and environs. The consequences of credible incidents were considered and are discussed.

  9. Compositional Safety Analysis using Barrier Certificates

    DEFF Research Database (Denmark)

    Sloth, Christoffer; Pappas, George J.; Wisniewski, Rafael

    2012-01-01

    This paper proposes a compositional method for verifying the safety of a dynamical system, given as an interconnection of subsystems. The safety verification is conducted by the use of the barrier certificate method; hence, the contribution of this paper is to show how to obtain compositional...... conditions for safety verification. We show how to formulate the verification problem, as a composition of coupled subproblems, each given for one subsystem. Furthermore, we show how to find the compositional barrier certificates via linear and sum of squares programming problems. The proposed method makes...

  10. Improved Methodology of MSLB M/E Release Analysis for OPR1000

    International Nuclear Information System (INIS)

    Park, Seok Jeong; Kim, Cheol Woo; Seo, Jong Tae

    2006-01-01

    A new mass and energy (M/E) release analysis methodology for the equipment environmental qualification (EEQ) on loss-of-coolant accident (LOCA) has been recently developed and adopted on small break LOCA EEQ. The new methodology for the M/E release analysis is extended to the M/E release analysis for the containment design for large break LOCA and the main steam line break (MSLB) accident, and named KIMERA (KOPEC Improved Mass and Energy Release Analysis) methodology. The computer code systems used in this methodology is RELAP5K/CONTEMPT4 (or RELAP5-ME) which couples RELAP5/MOD3.1/K with enhanced M/E model and LOCA long term model, and CONTEMPT4/ MOD5. This KIMERA methodology is applied to the MSLB M/E release analysis to evaluate the validation of KIMERA methodology for MSLB in containment design. The results are compared with the OPR 1000 FSAR

  11. Component Testing Methodology for Safety Grade PLC(POSAFE-Q) Operating Systems

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Hwang, Sung Jae [Poscon Corp, Seoul (Korea, Republic of); Sung, Ah-Young [Ewha Womans University, Seoul (Korea, Republic of); Son, Han Sung [EnEsys Corp. Taejon (Korea, Republic of)

    2006-07-01

    A Safety-Grade Programmable Logic Controller (POSAFE-Q) can be applied to a Reactor Protection System in nuclear power plants. This POSAFE-Q is being developed in the Korea Nuclear Instrumentation and Control System (KNICS). A Processor Module Operating System is developed to operate a POSAFE-Q. A variety of testing is executing to guarantee perfectly functionalities and performances. This Paper addresses the method, procedure and testing environment that are needed for component testing.

  12. Defining the methodological challenges and opportunities for an effective science of sociotechnical systems and safety

    Science.gov (United States)

    Waterson, Patrick; Robertson, Michelle M.; Cooke, Nancy J.; Militello, Laura; Roth, Emilie; Stanton, Neville A.

    2015-01-01

    An important part of the application of sociotechnical systems theory (STS) is the development of methods, tools and techniques to assess human factors and ergonomics workplace requirements. We focus in this paper on describing and evaluating current STS methods for workplace safety, as well as outlining a set of six case studies covering the application of these methods to a range of safety contexts. We also describe an evaluation of the methods in terms of ratings of their ability to address a set of theoretical and practical questions (e.g. the degree to which methods capture static/dynamic aspects of tasks and interactions between system levels). The outcomes from the evaluation highlight a set of gaps relating to the coverage and applicability of current methods for STS and safety (e.g. coverage of external influences on system functioning; method usability). The final sections of the paper describe a set of future challenges, as well as some practical suggestions for tackling these. Practitioner Summary: We provide an up-to-date review of STS methods, a set of case studies illustrating their use and an evaluation of their strengths and weaknesses. The paper concludes with a ‘roadmap’ for future work. PMID:25832121

  13. Preliminary Safety Analysis Report for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Motloch, C.G.; Bonney, R.F.; Levine, J.D.; Masson, L.S.; Commander, J.C.

    1995-04-01

    This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR

  14. Probabilistic safety analysis and interpretation thereof

    International Nuclear Information System (INIS)

    Steininger, U.; Sacher, H.

    1999-01-01

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

  15. Gap Analysis Approach for Construction Safety Program Improvement

    Directory of Open Access Journals (Sweden)

    Thanet Aksorn

    2007-06-01

    Full Text Available To improve construction site safety, emphasis has been placed on the implementation of safety programs. In order to successfully gain from safety programs, factors that affect their improvement need to be studied. Sixteen critical success factors of safety programs were identified from safety literature, and these were validated by safety experts. This study was undertaken by surveying 70 respondents from medium- and large-scale construction projects. It explored the importance and the actual status of critical success factors (CSFs. Gap analysis was used to examine the differences between the importance of these CSFs and their actual status. This study found that the most critical problems characterized by the largest gaps were management support, appropriate supervision, sufficient resource allocation, teamwork, and effective enforcement. Raising these priority factors to satisfactory levels would lead to successful safety programs, thereby minimizing accidents.

  16. A dynamic probabilistic safety margin characterization approach in support of Integrated Deterministic and Probabilistic Safety Analysis

    International Nuclear Information System (INIS)

    Di Maio, Francesco; Rai, Ajit; Zio, Enrico

    2016-01-01

    The challenge of Risk-Informed Safety Margin Characterization (RISMC) is to develop a methodology for estimating system safety margins in the presence of stochastic and epistemic uncertainties affecting the system dynamic behavior. This is useful to support decision-making for licensing purposes. In the present work, safety margin uncertainties are handled by Order Statistics (OS) (with both Bracketing and Coverage approaches) to jointly estimate percentiles of the distributions of the safety parameter and of the time required for it to reach these percentiles values during its dynamic evolution. The novelty of the proposed approach consists in the integration of dynamic aspects (i.e., timing of events) into the definition of a dynamic safety margin for a probabilistic Quantification of Margin and Uncertainties (QMU). The system here considered for demonstration purposes is the Lead–Bismuth Eutectic- eXperimental Accelerator Driven System (LBE-XADS). - Highlights: • We integrate dynamic aspects into the definition of a safety margins. • We consider stochastic and epistemic uncertainties affecting the system dynamics. • Uncertainties are handled by Order Statistics (OS). • We estimate the system grace time during accidental scenarios. • We apply the approach to an LBE-XADS accidental scenario.

  17. Methodologies for analysis of patterning in the mouse RPE sheet

    Science.gov (United States)

    Boatright, Jeffrey H.; Dalal, Nupur; Chrenek, Micah A.; Gardner, Christopher; Ziesel, Alison; Jiang, Yi; Grossniklaus, Hans E.

    2015-01-01

    Purpose Our goal was to optimize procedures for assessing shapes, sizes, and other quantitative metrics of retinal pigment epithelium (RPE) cells and contact- and noncontact-mediated cell-to-cell interactions across a large series of flatmount RPE images. Methods The two principal methodological advances of this study were optimization of a mouse RPE flatmount preparation and refinement of open-access software to rapidly analyze large numbers of flatmount images. Mouse eyes were harvested, and extra-orbital fat and muscles were removed. Eyes were fixed for 10 min, and dissected by puncturing the cornea with a sharp needle or a stab knife. Four radial cuts were made with iridectomy scissors from the puncture to near the optic nerve head. The lens, iris, and the neural retina were removed, leaving the RPE sheet exposed. The dissection and outcomes were monitored and evaluated by video recording. The RPE sheet was imaged under fluorescence confocal microscopy after staining for ZO-1 to identify RPE cell boundaries. Photoshop, Java, Perl, and Matlab scripts, as well as CellProfiler, were used to quantify selected parameters. Data were exported into Excel spreadsheets for further analysis. Results A simplified dissection procedure afforded a consistent source of images that could be processed by computer. The dissection and flatmounting techniques were illustrated in a video recording. Almost all of the sheet could be routinely imaged, and substantial fractions of the RPE sheet (usually 20–50% of the sheet) could be analyzed. Several common technical problems were noted and workarounds developed. The software-based analysis merged 25 to 36 images into one and adjusted settings to record an image suitable for large-scale identification of cell-to-cell boundaries, and then obtained quantitative descriptors of the shape of each cell, its neighbors, and interactions beyond direct cell–cell contact in the sheet. To validate the software, human- and computer

  18. Methodologies for analysis of patterning in the mouse RPE sheet.

    Science.gov (United States)

    Boatright, Jeffrey H; Dalal, Nupur; Chrenek, Micah A; Gardner, Christopher; Ziesel, Alison; Jiang, Yi; Grossniklaus, Hans E; Nickerson, John M

    2015-01-01

    Our goal was to optimize procedures for assessing shapes, sizes, and other quantitative metrics of retinal pigment epithelium (RPE) cells and contact- and noncontact-mediated cell-to-cell interactions across a large series of flatmount RPE images. The two principal methodological advances of this study were optimization of a mouse RPE flatmount preparation and refinement of open-access software to rapidly analyze large numbers of flatmount images. Mouse eyes were harvested, and extra-orbital fat and muscles were removed. Eyes were fixed for 10 min, and dissected by puncturing the cornea with a sharp needle or a stab knife. Four radial cuts were made with iridectomy scissors from the puncture to near the optic nerve head. The lens, iris, and the neural retina were removed, leaving the RPE sheet exposed. The dissection and outcomes were monitored and evaluated by video recording. The RPE sheet was imaged under fluorescence confocal microscopy after staining for ZO-1 to identify RPE cell boundaries. Photoshop, Java, Perl, and Matlab scripts, as well as CellProfiler, were used to quantify selected parameters. Data were exported into Excel spreadsheets for further analysis. A simplified dissection procedure afforded a consistent source of images that could be processed by computer. The dissection and flatmounting techniques were illustrated in a video recording. Almost all of the sheet could be routinely imaged, and substantial fractions of the RPE sheet (usually 20-50% of the sheet) could be analyzed. Several common technical problems were noted and workarounds developed. The software-based analysis merged 25 to 36 images into one and adjusted settings to record an image suitable for large-scale identification of cell-to-cell boundaries, and then obtained quantitative descriptors of the shape of each cell, its neighbors, and interactions beyond direct cell-cell contact in the sheet. To validate the software, human- and computer-analyzed results were compared. Whether

  19. Iterative transport-diffusion methodology for LWR core analysis

    International Nuclear Information System (INIS)

    Colameco, D.; Beacon, D.; Ivanov, K.N.; Inanov, B.D.

    2013-01-01

    This paper presents an update on the development of an advanced methodology for Light Water Reactor core calculations that uses local heterogeneous solutions for on-the-fly nodal cross-section generation. The Iterative Transport-Diffusion Method (ITDM) is an embedded transport approach that is expected to provide results with near 3D transport accuracy for a fraction of the time required by a full 3D transport method. In this methodology, the infinite environment used for homogenized nodal cross-section generation is replaced with a simulated 3D environment of the diffusion calculation. It is shown that the ITDM methodology provides very promising results when using partial currents as boundary conditions for loosely coupling a 2D lattice transport code to a 3D core nodal solver. The use of partial currents is a major improvement over the albedo concept: the solutions converged in a smoother manner

  20. On integration of probabilistic and deterministic safety analysis

    International Nuclear Information System (INIS)

    Cepin, M.; Wardzinski, A.

    1996-01-01

    The paper presents the case study on probabilistic and deterministic safety analysis of Engineered Safety Features Actuation System. The Fault Tree as a Probabilistic Safety Assessment tool is developed and analysed. The same Fault Tree is specified in a formal way. When formalized, it has a possibility to include the time requirements of the analysed system, which can not be included in a probabilistic approach to Fault Tree Analysis. The feature of inclusion of time is the main advantage of formalized Fault Tree, which extends it to a dynamic tool. Its results are Minimal Cut Sets with time relations, which are the base for the definition of safety requirements. Definition of safety requirements is one of early phases of software lifecycle and it is of special importance designing safety-related computer systems. (author)

  1. A Methodology for the Analysis and Selection of Alternative for the Disposition of Surplus Plutonium

    International Nuclear Information System (INIS)

    1999-01-01

    The Department of Energy (DOE) - Office of Fissile Materials Disposition (OFMD) has announced a Record of Decision (ROD) selecting alternatives for disposition of surplus plutonium. A major objective of this decision was to further U.S. efforts to prevent the proliferation of nuclear weapons. Other concerns that were addressed include economic, technical, institutional, schedule, environmental, and health and safety issues. The technical, environmental, and nonproliferation analyses supporting the ROD are documented in three DOE reports (DOE-TSR 96, DOE-PEIS 96, and DOE-NN 97, respectively). At the request of OFMD, a team of analysts from the Amarillo National Resource Center for Plutonium (ANRCP) provided an independent evaluation of the alternatives for plutonium that were considered during the evaluation effort. This report outlines the methodology used by the ANRCP team. This methodology, referred to as multiattribute utility theory (MAU), provides a structure for assembling results of detailed technical, economic, schedule, environment, and nonproliferation analyses for OFMD, DOE policy makers, other stakeholders, and the general public in a systematic way. The MAU methodology has been supported for use in similar situations by the National Research Council, an agency of the National Academy of Sciences.1 It is important to emphasize that the MAU process does not lead to a computerized model that actually determines the decision for a complex problem. MAU is a management tool that is one component, albeit a key component, of a decision process. We subscribe to the philosophy that the result of using models should be insights, not numbers. The MAU approach consists of four steps: (1) identification of alternatives, objectives, and performance measures, (2) estimation of the performance of the alternatives with respect to the objectives, (3) development of value functions and weights for the objectives, and (4) evaluation of the alternatives and sensitivity

  2. A Methodology for the Analysis and Selection of Alternative for the Disposition of Surplus Plutonium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-08-31

    The Department of Energy (DOE) - Office of Fissile Materials Disposition (OFMD) has announced a Record of Decision (ROD) selecting alternatives for disposition of surplus plutonium. A major objective of this decision was to further U.S. efforts to prevent the proliferation of nuclear weapons. Other concerns that were addressed include economic, technical, institutional, schedule, environmental, and health and safety issues. The technical, environmental, and nonproliferation analyses supporting the ROD are documented in three DOE reports [DOE-TSR 96, DOE-PEIS 96, and DOE-NN 97, respectively]. At the request of OFMD, a team of analysts from the Amarillo National Resource Center for Plutonium (ANRCP) provided an independent evaluation of the alternatives for plutonium that were considered during the evaluation effort. This report outlines the methodology used by the ANRCP team. This methodology, referred to as multiattribute utility theory (MAU), provides a structure for assembling results of detailed technical, economic, schedule, environment, and nonproliferation analyses for OFMD, DOE policy makers, other stakeholders, and the general public in a systematic way. The MAU methodology has been supported for use in similar situations by the National Research Council, an agency of the National Academy of Sciences.1 It is important to emphasize that the MAU process does not lead to a computerized model that actually determines the decision for a complex problem. MAU is a management tool that is one component, albeit a key component, of a decision process. We subscribe to the philosophy that the result of using models should be insights, not numbers. The MAU approach consists of four steps: (1) identification of alternatives, objectives, and performance measures, (2) estimation of the performance of the alternatives with respect to the objectives, (3) development of value functions and weights for the objectives, and (4) evaluation of the alternatives and sensitivity

  3. The analysis of classroom talk: methods and methodologies.

    Science.gov (United States)

    Mercer, Neil

    2010-03-01

    This article describes methods for analysing classroom talk, comparing their strengths and weaknesses. Both quantitative and qualitative methods are described and assessed for their strengths and weaknesses, with a discussion of the mixed use of such methods. It is acknowledged that particular methods are often embedded in particular methodologies, which are based on specific theories of social action, research paradigms, and disciplines; and so a comparison is made of two contemporary methodologies, linguistic ethnography, and sociocultural research. The article concludes with some comments on the current state of development of this field of research and on ways that it might usefully progress.

  4. Multimorbidity and Patient Safety Incidents in Primary Care: A Systematic Review and Meta-Analysis

    Science.gov (United States)

    Panagioti, Maria; Stokes, Jonathan; Esmail, Aneez; Coventry, Peter; Cheraghi-Sohi, Sudeh; Alam, Rahul; Bower, Peter

    2015-01-01

    Background Multimorbidity is increasingly prevalent and represents a major challenge in primary care. Patients with multimorbidity are potentially more likely to experience safety incidents due to the complexity of their needs and frequency of their interactions with health services. However, rigorous syntheses of the link between patient safety incidents and multimorbidity are not available. This review examined the relationship between multimorbidity and patient safety incidents in primary care. Methods We followed our published protocol (PROSPERO registration number: CRD42014007434). Medline, Embase and CINAHL were searched up to May 2015. Study design and quality were assessed. Odds ratios (OR) and 95% confidence intervals (95% CIs) were calculated for the associations between multimorbidity and two categories of patient safety outcomes: ‘active patient safety incidents’ (such as adverse drug events and medical complications) and ‘precursors of safety incidents’ (such as prescription errors, medication non-adherence, poor quality of care and diagnostic errors). Meta-analyses using random effects models were undertaken. Results Eighty six relevant comparisons from 75 studies were included in the analysis. Meta-analysis demonstrated that physical-mental multimorbidity was associated with an increased risk for ‘active patient safety incidents’ (OR = 2.39, 95% CI = 1.40 to 3.38) and ‘precursors of safety incidents’ (OR = 1.69, 95% CI = 1.36 to 2.03). Physical multimorbidity was associated with an increased risk for active safety incidents (OR = 1.63, 95% CI = 1.45 to 1.80) but was not associated with precursors of safety incidents (OR = 1.02, 95% CI = 0.90 to 1.13). Statistical heterogeneity was high and the methodological quality of the studies was generally low. Conclusions The association between multimorbidity and patient safety is complex, and varies by type of multimorbidity and type of safety incident. Our analyses suggest that multimorbidity

  5. Multimorbidity and Patient Safety Incidents in Primary Care: A Systematic Review and Meta-Analysis.

    Science.gov (United States)

    Panagioti, Maria; Stokes, Jonathan; Esmail, Aneez; Coventry, Peter; Cheraghi-Sohi, Sudeh; Alam, Rahul; Bower, Peter

    2015-01-01

    Multimorbidity is increasingly prevalent and represents a major challenge in primary care. Patients with multimorbidity are potentially more likely to experience safety incidents due to the complexity of their needs and frequency of their interactions with health services. However, rigorous syntheses of the link between patient safety incidents and multimorbidity are not available. This review examined the relationship between multimorbidity and patient safety incidents in primary care. We followed our published protocol (PROSPERO registration number: CRD42014007434). Medline, Embase and CINAHL were searched up to May 2015. Study design and quality were assessed. Odds ratios (OR) and 95% confidence intervals (95% CIs) were calculated for the associations between multimorbidity and two categories of patient safety outcomes: 'active patient safety incidents' (such as adverse drug events and medical complications) and 'precursors of safety incidents' (such as prescription errors, medication non-adherence, poor quality of care and diagnostic errors). Meta-analyses using random effects models were undertaken. Eighty six relevant comparisons from 75 studies were included in the analysis. Meta-analysis demonstrated that physical-mental multimorbidity was associated with an increased risk for 'active patient safety incidents' (OR = 2.39, 95% CI = 1.40 to 3.38) and 'precursors of safety incidents' (OR = 1.69, 95% CI = 1.36 to 2.03). Physical multimorbidity was associated with an increased risk for active safety incidents (OR = 1.63, 95% CI = 1.45 to 1.80) but was not associated with precursors of safety incidents (OR = 1.02, 95% CI = 0.90 to 1.13). Statistical heterogeneity was high and the methodological quality of the studies was generally low. The association between multimorbidity and patient safety is complex, and varies by type of multimorbidity and type of safety incident. Our analyses suggest that multimorbidity involving mental health may be a key driver of safety

  6. Multimorbidity and Patient Safety Incidents in Primary Care: A Systematic Review and Meta-Analysis.

    Directory of Open Access Journals (Sweden)

    Maria Panagioti

    Full Text Available Multimorbidity is increasingly prevalent and represents a major challenge in primary care. Patients with multimorbidity are potentially more likely to experience safety incidents due to the complexity of their needs and frequency of their interactions with health services. However, rigorous syntheses of the link between patient safety incidents and multimorbidity are not available. This review examined the relationship between multimorbidity and patient safety incidents in primary care.We followed our published protocol (PROSPERO registration number: CRD42014007434. Medline, Embase and CINAHL were searched up to May 2015. Study design and quality were assessed. Odds ratios (OR and 95% confidence intervals (95% CIs were calculated for the associations between multimorbidity and two categories of patient safety outcomes: 'active patient safety incidents' (such as adverse drug events and medical complications and 'precursors of safety incidents' (such as prescription errors, medication non-adherence, poor quality of care and diagnostic errors. Meta-analyses using random effects models were undertaken.Eighty six relevant comparisons from 75 studies were included in the analysis. Meta-analysis demonstrated that physical-mental multimorbidity was associated with an increased risk for 'active patient safety incidents' (OR = 2.39, 95% CI = 1.40 to 3.38 and 'precursors of safety incidents' (OR = 1.69, 95% CI = 1.36 to 2.03. Physical multimorbidity was associated with an increased risk for active safety incidents (OR = 1.63, 95% CI = 1.45 to 1.80 but was not associated with precursors of safety incidents (OR = 1.02, 95% CI = 0.90 to 1.13. Statistical heterogeneity was high and the methodological quality of the studies was generally low.The association between multimorbidity and patient safety is complex, and varies by type of multimorbidity and type of safety incident. Our analyses suggest that multimorbidity involving mental health may be a key driver of

  7. Review of dynamic reliability methodology in probabilistic safety assessments of nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Yongfa; Tong Jiejuan; Zhou Yu; Cai Qi

    2012-01-01

    Dynamic reliability method can make up the shortage of traditional Event Tree/Fault Tree method. It can be used to complete the current reliability and safety assessment methods of nuclear power plants. In recent 30 years. a relatively perfect theory foundation (probabilistic dynamics) has been developed. On this basis, Monte Carlo simulation and discrete dynamic event tree were formed. The research situation and technical characteristics of the dynamic reliability theory and method were introduced. And the future trend of dynamic reliability was also discussed. (authors)

  8. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  9. Probabilistic safety analysis : a new nuclear power plants licensing method

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de.

    1982-04-01

    After a brief retrospect of the application of Probabilistic Safety Analysis in the nuclear field, the basic differences between the deterministic licensing method, currently in use, and the probabilistic method are explained. Next, the two main proposals (by the AIF and the ACRS) concerning the establishment of the so-called quantitative safety goals (or simply 'safety goals') are separately presented and afterwards compared in their most fundamental aspects. Finally, some recent applications and future possibilities are discussed. (Author) [pt

  10. Safety analysis report for the Waste Storage Facility. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  11. Evaluation of atmospheric dispersion/consequence models supporting safety analysis

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. Nuclear safety in Slovak Republic. Safety analysis reports for WWER 440 reactors

    International Nuclear Information System (INIS)

    Rohar, S.

    1999-01-01

    Implementation of nuclear power program is connected to establishment of regulatory body for safe regulation of siting, construction, operation and decommissioning of nuclear installations. Licensing being one of the most important regulatory surveillance activity is based on independent regulatory review and assessment of information on nuclear safety for particular nuclear facility. Documents required to be submitted to the regulatory body by the licensee in Slovakia for the review and assessment usually named Safety Analysis Report (SAR) are presented in detail in this paper. Current status of Safety Analysis Reports for Bohunice V-1, Bohunice V-2 and Mochovce NPP is shown

  13. Applying importance-performance analysis to patient safety culture.

    Science.gov (United States)

    Lee, Yii-Ching; Wu, Hsin-Hung; Hsieh, Wan-Lin; Weng, Shao-Jen; Hsieh, Liang-Po; Huang, Chih-Hsuan

    2015-01-01

    The Sexton et al.'s (2006) safety attitudes questionnaire (SAQ) has been widely used to assess staff's attitudes towards patient safety in healthcare organizations. However, to date there have been few studies that discuss the perceptions of patient safety both from hospital staff and upper management. The purpose of this paper is to improve and to develop better strategies regarding patient safety in healthcare organizations. The Chinese version of SAQ based on the Taiwan Joint Commission on Hospital Accreditation is used to evaluate the perceptions of hospital staff. The current study then lies in applying importance-performance analysis technique to identify the major strengths and weaknesses of the safety culture. The results show that teamwork climate, safety climate, job satisfaction, stress recognition and working conditions are major strengths and should be maintained in order to provide a better patient safety culture. On the contrary, perceptions of management and hospital handoffs and transitions are important weaknesses and should be improved immediately. Research limitations/implications - The research is restricted in generalizability. The assessment of hospital staff in patient safety culture is physicians and registered nurses. It would be interesting to further evaluate other staff's (e.g. technicians, pharmacists and others) opinions regarding patient safety culture in the hospital. Few studies have clearly evaluated the perceptions of healthcare organization management regarding patient safety culture. Healthcare managers enable to take more effective actions to improve the level of patient safety by investigating key characteristics (either strengths or weaknesses) that healthcare organizations should focus on.

  14. Towards a Methodological Improvement of Narrative Inquiry: A Qualitative Analysis

    Science.gov (United States)

    Abdallah, Mahmoud Mohammad Sayed

    2009-01-01

    The article suggests that though narrative inquiry as a research methodology entails free conversations and personal stories, yet it should not be totally free and fictional as it has to conform to some recognized standards used for conducting educational research. Hence, a qualitative study conducted by Russ (1999) was explored as an exemplar…

  15. Seismic hazard analysis. A methodology for the Eastern United States

    International Nuclear Information System (INIS)

    Bernreuter, D.L.

    1980-08-01

    This report presents a probabilistic approach for estimating the seismic hazard in the Central and Eastern United States. The probabilistic model (Uniform Hazard Methodology) systematically incorporates the subjective opinion of several experts in the evaluation of seismic hazard. Subjective input, assumptions and associated hazard are kept separate for each expert so as to allow review and preserve diversity of opinion. The report is organized into five sections: Introduction, Methodology Comparison, Subjective Input, Uniform Hazard Methodology (UHM), and Uniform Hazard Spectrum. Section 2 Methodology Comparison, briefly describes the present approach and compares it with other available procedures. The remainder of the report focuses on the UHM. Specifically, Section 3 describes the elicitation of subjective input; Section 4 gives details of various mathematical models (earthquake source geometry, magnitude distribution, attenuation relationship) and how these models re combined to calculate seismic hazard. The lost section, Uniform Hazard Spectrum, highlights the main features of typical results. Specific results and sensitivity analyses are not presented in this report. (author)

  16. Qualitative Analysis of Comic Strip Culture: A Methodological Inquiry.

    Science.gov (United States)

    Newman, Isadore; And Others

    The paper is a methodological inquiry into the interpretation of qualitative data. It explores a grounded-theory approach to the synthesis of data and examines, in particular, the construction of categories. It focuses on ways of organizing and attaching meaning to data, as research problems embedded in a cultural context are explored. A…

  17. Grounded Theory and Educational Ethnography: A Methodological Analysis and Critique.

    Science.gov (United States)

    Smith, Louis M.; Pohland, Paul A.

    This paper analyzes and evaluates the methodological approach developed by B. G. Glaser and A. L. Strauss in THE DISCOVERY OF GROUNDED THEORY (Chicago: Aldine, 1967). Smith and Pohland's major intent is to raise Glaser and Strauss' most significant concepts and issues, analyze them in the context of seven of their own studies, and in conclusion…

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

    Directory of Open Access Journals (Sweden)

    Putra Perdana Suteja

    2018-01-01

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

  19. Attack Methodology Analysis: Emerging Trends in Computer-Based Attack Methodologies and Their Applicability to Control System Networks

    Energy Technology Data Exchange (ETDEWEB)

    Bri Rolston

    2005-06-01

    Threat characterization is a key component in evaluating the threat faced by control systems. Without a thorough understanding of the threat faced by critical infrastructure networks, adequate resources cannot be allocated or directed effectively to the defense of these systems. Traditional methods of threat analysis focus on identifying the capabilities and motivations of a specific attacker, assessing the value the adversary would place on targeted systems, and deploying defenses according to the threat posed by the potential adversary. Too many effective exploits and tools exist and are easily accessible to anyone with access to an Internet connection, minimal technical skills, and a significantly reduced motivational threshold to be able to narrow the field of potential adversaries effectively. Understanding how hackers evaluate new IT security research and incorporate significant new ideas into their own tools provides a means of anticipating how IT systems are most likely to be attacked in the future. This research, Attack Methodology Analysis (AMA), could supply pertinent information on how to detect and stop new types of attacks. Since the exploit methodologies and attack vectors developed in the general Information Technology (IT) arena can be converted for use against control system environments, assessing areas in which cutting edge exploit development and remediation techniques are occurring can provide significance intelligence for control system network exploitation, defense, and a means of assessing threat without identifying specific capabilities of individual opponents. Attack Methodology Analysis begins with the study of what exploit technology and attack methodologies are being developed in the Information Technology (IT) security research community within the black and white hat community. Once a solid understanding of the cutting edge security research is established, emerging trends in attack methodology can be identified and the gap between

  20. Galileo and Ulysses missions safety analysis and launch readiness status

    International Nuclear Information System (INIS)

    Cork, M.J.; Turi, J.A.

    1989-01-01

    The Galileo spacecraft will explore the Jupiter system and Ulysses will fly by Jupiter en route to a polar orbit of the sun. Both spacecraft are powered by general purpose heat source radioisotope thermoelectric generators (RTGs). As a result of the Challenger accident and subsequent mission reprogramming, the Galileo and Ulysses missions' safety analysis had to be repeated. In addition to presenting an overview of the safety analysis status for the missions, this paper presents a brief review of the missions' objectives and design approaches, RTG design characteristics and development history, and a description of the safety analysis process. (author)