WorldWideScience

Sample records for safety evaluation methodology

  1. PWR reload safety evaluation methodology

    International Nuclear Information System (INIS)

    Doshi, P.K.; Chapin, D.L.; Love, D.S.

    1993-01-01

    The current practice for WWER safety analysis is to prepare the plant Safety Analysis Report (SAR) for initial plant operation. However, the existing safety analysis is typically not evaluated for reload cycles to confirm that all safety limits are met. In addition, there is no systematic reanalysis or reevaluation of the safety analyses after there have been changes made to the plant. The Westinghouse process is discussed which is in contrast to this and in which the SAR conclusions are re-validated through evaluation and/or analysis of each reload cycle. (Z.S.)

  2. Safety analysis and evaluation methodology for fusion systems

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  3. 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

  4. A Methodology for Evaluating Quantitative Nuclear Safety Culture Impact

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  5. A progressive methodology for seismic safety evaluation of gravity dams

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  6. 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)

  7. Methodology development for statistical evaluation of reactor safety analyses

    International Nuclear Information System (INIS)

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

    1976-07-01

    In February 1975, Westinghouse Electric Corporation, under contract to Electric Power Research Institute, started a one-year program to develop methodology for statistical evaluation of nuclear-safety-related engineering analyses. The objectives of the program were to develop an understanding of the relative efficiencies of various computational methods which can be used to compute probability distributions of output variables due to input parameter uncertainties in analyses of design basis events for nuclear reactors and to develop methods for obtaining reasonably accurate estimates of these probability distributions at an economically feasible level. A series of tasks was set up to accomplish these objectives. Two of the tasks were to investigate the relative efficiencies and accuracies of various Monte Carlo and analytical techniques for obtaining such estimates for a simple thermal-hydraulic problem whose output variable of interest is given in a closed-form relationship of the input variables and to repeat the above study on a thermal-hydraulic problem in which the relationship between the predicted variable and the inputs is described by a short-running computer program. The purpose of the report presented is to document the results of the investigations completed under these tasks, giving the rationale for choices of techniques and problems, and to present interim conclusions

  8. 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

  9. 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

  10. 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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sukyoung; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Jung Taek [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Tae Wan [Kepco International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-05-15

    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

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

    International Nuclear Information System (INIS)

    Esteves, R.G.

    1981-01-01

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

  14. 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.

  15. 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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  17. Blanket safety by GEMSAFE methodology

    International Nuclear Information System (INIS)

    Sawada, Tetsuo; Saito, Masaki

    2001-01-01

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

  18. 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.)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  20. 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

  1. 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)

  2. Effectiveness evaluation methodology for safety processes to enhance organisational culture in hazardous installations

    International Nuclear Information System (INIS)

    Mengolini, A.; Debarberis, L.

    2008-01-01

    Safety performance indicators are widely collected and used in hazardous installations. The IAEA, OECD and other international organisations have developed approaches that strongly promote deployment of safety performance indicators. These indicators focus mainly on operational performance, but some of them also address organisational and safety culture aspects. However, operators of hazardous installations, in particular those with limited resources and time constraints, often find it difficult to collect the large number of different safety performance indicators. Moreover, they also have difficulties with giving a meaning to the numbers and trends recorded, especially to those that should reflect a positive safety culture. In this light, the aim of this article is to address the need to monitor and assess progress on implementation of a programme to enhance safety and organisational culture. It proposes a specific process-view approach to effectiveness evaluation of organisational and safety culture indicators by means of a multi-level system in which safety processes and staff involvement in defining improvement activities are central. In this way safety becomes fully embedded in staff activities. Key members of personnel become directly involved in identifying and supplying leading indicators relating to their own daily activity and become responsible and accountable for keeping the measurement system alive. Besides use of lagging indicators, particular emphasis is placed on the importance of identifying and selecting leading indicators which can be used to drive safety performance for organisational and safety culture aspects as well

  3. Effectiveness evaluation methodology for safety processes to enhance organisational culture in hazardous installations.

    Science.gov (United States)

    Mengolini, A; Debarberis, L

    2008-06-30

    Safety performance indicators are widely collected and used in hazardous installations. The IAEA, OECD and other international organisations have developed approaches that strongly promote deployment of safety performance indicators. These indicators focus mainly on operational performance, but some of them also address organisational and safety culture aspects. However, operators of hazardous installations, in particular those with limited resources and time constraints, often find it difficult to collect the large number of different safety performance indicators. Moreover, they also have difficulties with giving a meaning to the numbers and trends recorded, especially to those that should reflect a positive safety culture. In this light, the aim of this article is to address the need to monitor and assess progress on implementation of a programme to enhance safety and organisational culture. It proposes a specific process-view approach to effectiveness evaluation of organisational and safety culture indicators by means of a multi-level system in which safety processes and staff involvement in defining improvement activities are central. In this way safety becomes fully embedded in staff activities. Key members of personnel become directly involved in identifying and supplying leading indicators relating to their own daily activity and become responsible and accountable for keeping the measurement system alive. Besides use of lagging indicators, particular emphasis is placed on the importance of identifying and selecting leading indicators which can be used to drive safety performance for organisational and safety culture aspects as well.

  4. 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

  5. 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

  6. Safety analysis methodology for OPR 1000

    International Nuclear Information System (INIS)

    Hwang-Yong, Jun

    2005-01-01

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

  7. 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

  8. Quantifying reactor safety margins: Part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology

    International Nuclear Information System (INIS)

    Boyack, B.E.; Duffey, R.B.; Griffith, P.

    1988-01-01

    In August 1988, the Nuclear Regulatory Commission (NRC) approved the final version of a revised rule on the acceptance of emergency core cooling systems (ECCS) entitled ''Emergency Core Cooling System; Revisions to Acceptance Criteria.'' The revised rule states an alternate ECCS performance analysis, based on best-estimate methods, may be used to provide more realistic estimates of plant safety margins, provided the licensee quantifies the uncertainty of the estimates and included that uncertainty when comparing the calculated results with prescribed acceptance limits. To support the revised ECCS rule, the NRC and its contractors and consultants have developed and demonstrated a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. It is an auditable, traceable, and practical method for combining quantitative analyses and expert opinions to arrive at computed values of uncertainty. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17 /times/ 17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3. 23 refs., 5 figs., 1 tab

  9. The status of nuclear safety of Ukrainian NPPS and evaluations by ASSET methodology

    Energy Technology Data Exchange (ETDEWEB)

    Koltakov, V [State Scientific and Technical Center on Nuclear and Radiational Safety, Kiev (Ukraine)

    1997-10-01

    The presentation discusses the following issues: nuclear power plants and research reactors; electricity production and operating personnel; status of nuclear safety at NPPs of Ukraine; operation of Ukrainian NPPs; violations of NPPs operation; short review of ASSET missions; experience of feedback process. Figs, tabs.

  10. The status of nuclear safety of Ukrainian NPPS and evaluations by ASSET methodology

    International Nuclear Information System (INIS)

    Koltakov, V.

    1997-01-01

    The presentation discusses the following issues: nuclear power plants and research reactors; electricity production and operating personnel; status of nuclear safety at NPPs of Ukraine; operation of Ukrainian NPPs; violations of NPPs operation; short review of ASSET missions; experience of feedback process. Figs, tabs

  11. Review of decision methodologies for evaluating regulatory actions affecting public health and safety

    International Nuclear Information System (INIS)

    Hendrickson, P.L.; McDonald, C.L.; Schilling, A.H.

    1976-12-01

    This report examines several aspects of the problems and choices facing the governmental decision maker who must take regulatory actions with multiple decision objectives and attributes. Particular attention is given to the problems facing the U.S. Nuclear Regulatory Commission (NRC) and to the decision attribute of chief concern to NRC, the protection of human health and safety, with emphasis on nuclear power plants. The study was undertaken to provide background information for NRC to use in refining its process of value/impact assessment of proposed regulatory actions. The principal conclusion is that approaches to rationally consider the value and impact of proposed regulatory actions are available. These approaches can potentially improve the decision-making process and enable the agency to better explain and defend its decisions. They also permit consistent examination of the impacts, effects of uncertainty and sensitivity to various assumptions of the alternatives being considered. Finally, these approaches can help to assure that affected parties are heard and that technical information is used appropriately and to the extent possible. The principal aspects of the regulatory decision problem covered in the report are: the legal setting for regulatory decisions which affect human health and safety, elements of the decision-making process, conceptual approaches to decision making, current approaches to decision making in several Federal agencies, and the determination of acceptable risk levels

  12. Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

    International Nuclear Information System (INIS)

    Hashim Muhammad; Yoshikawa, Hidekazu; Matsuoka, Takeshi; Yang Ming

    2014-01-01

    The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR. For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. (author)

  13. 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)

  14. An Integrated Safety Assessment Methodology for Generation IV Nuclear Systems

    International Nuclear Information System (INIS)

    Leahy, Timothy J.

    2010-01-01

    The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Early work of the RSWG focused on defining a safety philosophy founded on lessons learned from current and prior generations of nuclear technologies, and on identifying technology characteristics that may help achieve Generation IV safety goals. More recent RSWG work has focused on the definition of an integrated safety assessment methodology for evaluating the safety of Generation IV systems. The methodology, tentatively called ISAM, is an integrated 'toolkit' consisting of analytical techniques that are available and matched to appropriate stages of Generation IV system concept development. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time.

  15. A Methodology for Safety Culture Impact Assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  16. A Methodology for Safety Culture Impact Assessment

    International Nuclear Information System (INIS)

    Han, Kiyoon; Jae, Moosung

    2014-01-01

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

  17. Quantifying reactor safety margins: Application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

    International Nuclear Information System (INIS)

    Boyack, B.; Duffey, R.; Wilson, G.; Griffith, P.; Lellouche, G.; Levy, S.; Rohatgi, U.; Wulff, W.; Zuber, N.

    1989-12-01

    The US Nuclear Regulatory Commission (NRC) has issued a revised rule for loss-of-coolant accident/emergency core cooling system (ECCS) analysis of light water reactors to allow the use of best-estimate computer codes in safety analysis as an option. A key feature of this option requires the licensee to quantify the uncertainty of the calculations and include that uncertainty when comparing the calculated results with acceptance limits provided in 10 CFR Part 50. To support the revised ECCS rule and illustrate its application, the NRC and its contractors and consultants have developed and demonstrated an uncertainty evaluation methodology called code scaling, applicability, and uncertainty (CSAU). The CSAU methodology and an example application described in this report demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. The methodology is systematic and comprehensive as it addresses and integrates the scenario, experiments, code, and plant to resolve questions concerned with: (a) code capability to scale-up processes from test facility to full-scale nuclear power plants; (b) code applicability to safety studies of a postulated accident scenario in a specified nuclear power plant; and (c) quantifying uncertainties of calculated results. 127 refs., 55 figs., 40 tabs

  18. Safety methodology and risk targets

    International Nuclear Information System (INIS)

    Kazimi, M.S.

    1983-01-01

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

  19. A new methodology for non-contact accurate crack width measurement through photogrammetry for automated structural safety evaluation

    International Nuclear Information System (INIS)

    Jahanshahi, Mohammad R; Masri, Sami F

    2013-01-01

    In mechanical, aerospace and civil structures, cracks are important defects that can cause catastrophes if neglected. Visual inspection is currently the predominant method for crack assessment. This approach is tedious, labor-intensive, subjective and highly qualitative. An inexpensive alternative to current monitoring methods is to use a robotic system that could perform autonomous crack detection and quantification. To reach this goal, several image-based crack detection approaches have been developed; however, the crack thickness quantification, which is an essential element for a reliable structural condition assessment, has not been sufficiently investigated. In this paper, a new contact-less crack quantification methodology, based on computer vision and image processing concepts, is introduced and evaluated against a crack quantification approach which was previously developed by the authors. The proposed approach in this study utilizes depth perception to quantify crack thickness and, as opposed to most previous studies, needs no scale attachment to the region under inspection, which makes this approach ideal for incorporation with autonomous or semi-autonomous mobile inspection systems. Validation tests are performed to evaluate the performance of the proposed approach, and the results show that the new proposed approach outperforms the previously developed one. (paper)

  20. 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)

  1. A review of the current state-of-the-art methodology for handling bias and uncertainty in performing criticality safety evaluations. Final report

    International Nuclear Information System (INIS)

    Disney, R.K.

    1994-10-01

    The methodology for handling bias and uncertainty when calculational methods are used in criticality safety evaluations (CSE's) is a rapidly evolving technology. The changes in the methodology are driven by a number of factors. One factor responsible for changes in the methodology for handling bias and uncertainty in CSE's within the overview of the US Department of Energy (DOE) is a shift in the overview function from a ''site'' perception to a more uniform or ''national'' perception. Other causes for change or improvement in the methodology for handling calculational bias and uncertainty are; (1) an increased demand for benchmark criticals data to expand the area (range) of applicability of existing data, (2) a demand for new data to supplement existing benchmark criticals data, (3) the increased reliance on (or need for) computational benchmarks which supplement (or replace) experimental measurements in critical assemblies, and (4) an increased demand for benchmark data applicable to the expanded range of conditions and configurations encountered in DOE site restoration and remediation

  2. Safety significance evaluation system

    International Nuclear Information System (INIS)

    Lew, B.S.; Yee, D.; Brewer, W.K.; Quattro, P.J.; Kirby, K.D.

    1991-01-01

    This paper reports that the Pacific Gas and Electric Company (PG and E), in cooperation with ABZ, Incorporated and Science Applications International Corporation (SAIC), investigated the use of artificial intelligence-based programming techniques to assist utility personnel in regulatory compliance problems. The result of this investigation is that artificial intelligence-based programming techniques can successfully be applied to this problem. To demonstrate this, a general methodology was developed and several prototype systems based on this methodology were developed. The prototypes address U.S. Nuclear Regulatory Commission (NRC) event reportability requirements, technical specification compliance based on plant equipment status, and quality assurance assistance. This collection of prototype modules is named the safety significance evaluation system

  3. A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sunghyun; Kim, Namyeong; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2015-10-15

    In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant.

  4. A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Park, Sunghyun; Kim, Namyeong; Jae, Moosung

    2015-01-01

    In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant

  5. 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

  6. Risk-based Regulatory Evaluation Program methodology

    International Nuclear Information System (INIS)

    DuCharme, A.R.; Sanders, G.A.; Carlson, D.D.; Asselin, S.V.

    1987-01-01

    The objectives of this DOE-supported Regulatory Evaluation Progrwam are to analyze and evaluate the safety importance and economic significance of existing regulatory guidance in order to assist in the improvement of the regulatory process for current generation and future design reactors. A risk-based cost-benefit methodology was developed to evaluate the safety benefit and cost of specific regulations or Standard Review Plan sections. Risk-based methods can be used in lieu of or in combination with deterministic methods in developing regulatory requirements and reaching regulatory decisions

  7. LNG Safety Assessment Evaluation Methods

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

  8. Review of decision methodologies for evaluating regulatory actions affecting public health and safety. [Nuclear industry site selection

    Energy Technology Data Exchange (ETDEWEB)

    Hendrickson, P.L.; McDonald, C.L.; Schilling, A.H.

    1976-12-01

    This report examines several aspects of the problems and choices facing the governmental decision maker who must take regulatory actions with multiple decision objectives and attributes. Particular attention is given to the problems facing the U.S. Nuclear Regulatory Commission (NRC) and to the decision attribute of chief concern to NRC, the protection of human health and safety, with emphasis on nuclear power plants. The study was undertaken to provide background information for NRC to use in refining its process of value/impact assessment of proposed regulatory actions. The principal conclusion is that approaches to rationally consider the value and impact of proposed regulatory actions are available. These approaches can potentially improve the decision-making process and enable the agency to better explain and defend its decisions. They also permit consistent examination of the impacts, effects of uncertainty and sensitivity to various assumptions of the alternatives being considered. Finally, these approaches can help to assure that affected parties are heard and that technical information is used appropriately and to the extent possible. The principal aspects of the regulatory decision problem covered in the report are: the legal setting for regulatory decisions which affect human health and safety, elements of the decision-making process, conceptual approaches to decision making, current approaches to decision making in several Federal agencies, and the determination of acceptable risk levels.

  9. Application of realistic (best- estimate) methodologies for large break loss of coolant (LOCA) safety analysis: licensing of Westinghouse ASTRUM evaluation model in Spain

    International Nuclear Information System (INIS)

    Lage, Carlos; Frepoli, Cesare

    2010-01-01

    When the LOCA Final Acceptance Criteria for Light Water Reactors was issued in Appendix K of 10CFR50 both the USNRC and the industry recognized that the rule was highly conservative. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the USNRC began a research program to identify the degree of conservatism in those models permitted in the Appendix K rule and to develop improved thermal-hydraulic computer codes so that realistic accident analysis calculations could be performed. The overall results of this research program quantified the conservatism in the Appendix K rule and confirmed that some relaxation of the rule can be made without a loss in safety to the public. Also, from a risk-informed perspective it is recognized that conservatism is not always a complete defense for lack of sophistication in models. In 1988, as a result of the improved understanding of LOCA phenomena, the USNRC staff amended the requirements of 10 CFR 50.46 and Appendix K, 'ECCS Evaluation Models', so that a realistic evaluation model may be used to analyze the performance of the ECCS during a hypothetical LOCA. Under the amended rules, best-estimate plus uncertainty (BEPU) thermal-hydraulic analysis may be used in place of the overly prescriptive set of models mandated by Appendix K rule. Further guidance for the use of best-estimate codes was provided in Regulatory Guide 1.157 To demonstrate use of the revised ECCS rule, the USNRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology as an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis. More recently the CSAU principles have been generalized in the Evaluation Model Development and Assessment Process (EMDAP) of Regulatory Guide 1.203. ASTRUM is the Westinghouse Best Estimate Large Break LOCA evaluation model applicable to two-, three

  10. 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.

  11. 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)

  12. Methodology and recruitment for a randomised controlled trial to evaluate the safety of wahakura for infant bedsharing.

    Science.gov (United States)

    Tipene-Leach, David; Baddock, Sally; Williams, Sheila; Jones, Raymond; Tangiora, Angeline; Abel, Sally; Taylor, Barry

    2014-09-28

    Sudden Unexpected Death in Infancy (SUDI) has persistent high rates in deprived indigenous communities and much of this mortality is attributable to unsafe sleep environments. Whilst health promotion worldwide has concentrated on avoidance of bedsharing, the indigenous Māori community in New Zealand has reproduced a traditional flax bassinet (wahakura) designed to be used in ways that include bedsharing. To date there has been no assessment of the safety of this traditional sleeping device. This two arm randomised controlled trial is being conducted with 200 mother-baby dyads recruited from Māori communities in areas of high deprivation in the Hawkes Bay, New Zealand. They are randomised to wahakura or bassinet use and investigation includes questionnaires at baseline (pregnancy), when baby is 1, 3, and 6 months, and an overnight video sleep study at 1 month with monitoring of baby temperature and oxygen saturation, and measurement of baby urinary cotinine and maternal salivary oxytocin. Outcome measures are amount of time head covered, amount of time in thermal comfort zone, number of hypoxic events, amount of time in the assigned sleep device, amount of time breastfeeding, number of parental (non-feed related) touching infant events, amount of time in the prone sleep position, the number of behavioural arousals and the amount of time infant is awake overnight. Survey data will compare breastfeeding patterns at 1, 3, and 6 months as well as data on maternal mind-mindedness, maternal wellbeing, attachment to baby, and maternal sleep patterns. Indigenous communities require creative SUDI interventions that fit within their prevailing world view. This trial, and its assessment of the safety of a wahakura relative to a standard bassinet, is an important contribution to the range of SUDI prevention research being undertaken worldwide. Australian New Zealand Clinical Trials Registry: ACTRN12610000993099 Registered 16th November 2010.

  13. A methodology for Level 2 PSA evaluation with consideration of specific features for Low Power Shutdown Probabilistic Safety Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Gab; Seok, Ho [KEPCO-ENC, Yongin (Korea, Republic of)

    2015-05-15

    The primary objective of the Level 2 PSA during Lower Power/Shutdown (LPSD) operation is to provide insights into potential plant vulnerabilities with regard to accident progression. The shutdown risk information can be used to provide the information to develop outage risk management guidelines. The LPSD Level 2 analysis utilizes much of the at-power Level 2 analysis for bounding, conservative treatment of severe accident phenomena. But, for some portions of the analysis including Plant Operational States (POSs), LPSD-specific evaluations such as UPC related to the containment Equipment Hatch (E/H) with 4 bolts, Reactor Coolant System (RCS) Not Intact for severe accident phenomena are desired for realistic evaluation. All POSs are evaluated for their Large Release Frequency (LRF). Some POSs are evaluated conservatively utilizing the at-power models, and other POSs are evaluated in specific analysis. The overall LPSD Level 2 model is evaluated. If the containment E/H and one of the two doors on each of the personal air locks are closed as containment is operable at reduced RCS inventory operation, LRF is expected to be less than 10% of LPSD CDF.

  14. Screening Risk Evaluation methodology

    International Nuclear Information System (INIS)

    Hopper, K.M.

    1994-01-01

    The Screening Risk Evaluation (SRE) Guidance document is a set of guidelines provided for the uniform implementation of SREs performed on D ampersand D facilities. These guidelines are designed specifically for the completion of the second (semi-quantitative screening) phase of the D ampersand D Risk-Based Process. The SRE Guidance produces screening risk scores reflecting levels of risk through the use of risk ranking indices. Five types of possible risk are calculated from the SRE: current releases, worker exposures, future releases, physical hazards, and criticality. The Current Release Index (CRI) calculates the risk to human health and the environment from ongoing or probable releases within a one year time period. The Worker Exposure Index (WEI) calculates the risk to workers, occupants, and visitors in D ampersand D facilities of contaminant exposure. The Future Release Index (FRI) calculates the risk of future releases of contaminants, after one year, to human health and the environment. The Physical Hazards Index (PHI) calculates the risk-to human health due to factors other than that of contaminants. The index of Criticality is approached as a modifying factor to the entire SRE, due to the fact that criticality issues are strictly regulated under DOE. Screening risk results will be tabulated in matrix form and Total Risk will be calculated (weighted equation) to produce a score on which to base early action recommendations. Other recommendations from the screening risk scores will be made based either on individual index scores or from reweighted Total Risk calculations. All recommendations based on the SRE will be made based on a combination of screening risk scores, decision drivers, and other considerations, determined on a project by project basis. The SRE is the first and most important step in the overall D ampersand D project level decision making process

  15. Evaluation and assessment of nuclear power plant seismic methodology

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-03-01

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

  16. Evaluation and assessment of nuclear power plant seismic methodology

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  17. Evaluation of a Propolis Water Extract Using a Reliable RP-HPLC Methodology and In Vitro and In Vivo Efficacy and Safety Characterisation

    Science.gov (United States)

    Rocha, Bruno Alves; Bueno, Paula Carolina Pires; Vaz, Mirela Mara de Oliveira Lima Leite; Nascimento, Andresa Piacezzi; Ferreira, Nathália Ursoli; Moreno, Gabriela de Padua; Rodrigues, Marina Rezende; Costa-Machado, Ana Rita de Mello; Barizon, Edna Aparecida; Campos, Jacqueline Costa Lima; de Oliveira, Pollyanna Francielli; Acésio, Nathália de Oliveira; Martins, Sabrina de Paula Lima; Tavares, Denise Crispim; Berretta, Andresa Aparecida

    2013-01-01

    Since the beginning of propolis research, several groups have studied its antibacterial, antifungal, and antiviral properties. However, most of these studies have only employed propolis ethanolic extract (PEE) leading to little knowledge about the biological activities of propolis water extract (PWE). Based on this, in a previous study, we demonstrated the anti-inflammatory and immunomodulatory activities of PWE. In order to better understand the equilibrium between effectiveness and toxicity, which is essential for a new medicine, the characteristics of PWE were analyzed. We developed and validated an RP-HPLC method to chemically characterize PWE and PEE and evaluated the in vitro antioxidant/antimicrobial activity for both extracts and the safety of PWE via determining genotoxic potential using in vitro and in vivo mammalian micronucleus assays. We have concluded that the proposed analytical methodology was reliable, and both extracts showed similar chemical composition. The extracts presented antioxidant and antimicrobial effects, while PWE demonstrated higher antioxidant activity and more efficacious for the most of the microorganisms tested than PEE. Finally, PWE was shown to be safe using micronucleus assays. PMID:23710228

  18. Evaluation of a Propolis Water Extract Using a Reliable RP-HPLC Methodology and In Vitro and In Vivo Efficacy and Safety Characterisation

    Directory of Open Access Journals (Sweden)

    Bruno Alves Rocha

    2013-01-01

    Full Text Available Since the beginning of propolis research, several groups have studied its antibacterial, antifungal, and antiviral properties. However, most of these studies have only employed propolis ethanolic extract (PEE leading to little knowledge about the biological activities of propolis water extract (PWE. Based on this, in a previous study, we demonstrated the anti-inflammatory and immunomodulatory activities of PWE. In order to better understand the equilibrium between effectiveness and toxicity, which is essential for a new medicine, the characteristics of PWE were analyzed. We developed and validated an RP-HPLC method to chemically characterize PWE and PEE and evaluated the in vitro antioxidant/antimicrobial activity for both extracts and the safety of PWE via determining genotoxic potential using in vitro and in vivo mammalian micronucleus assays. We have concluded that the proposed analytical methodology was reliable, and both extracts showed similar chemical composition. The extracts presented antioxidant and antimicrobial effects, while PWE demonstrated higher antioxidant activity and more efficacious for the most of the microorganisms tested than PEE. Finally, PWE was shown to be safe using micronucleus assays.

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

    International Nuclear Information System (INIS)

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

    2005-07-01

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

  20. Safety on Judo Children: Methodology and Results

    OpenAIRE

    Sacripanti, Attilio; De Blasis, Tania

    2017-01-01

    Many doctors although they have not firsthand experience of judo, describe it as a sport unsuitable for children. Theoretically speaking falls derived by Judo throwing techniques,could be potentially dangerous,especially for kids,if poorly managed.A lot of researches were focalized on trauma or injuries taking place in judo, both during training and competition The goal of this Research is to define and apply a scientific methodology to evaluate the hazard in falls by judo throws for children...

  1. Latest developments on safety analysis methodologies at the Juzbado plant

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  2. Inventory differences: An evaluation methodology

    International Nuclear Information System (INIS)

    Heinberg, C.L.; Roberts, N.J.

    1987-01-01

    This paper discusses an evaluation methodology which is used for inventory differences at the Los Alamos National Laboratory. It is recognized that there are various methods which can be, and are being, used to evaluate process inventory differences at DOE facilities. The purpose of this paper is to share our thoughts on the subject and our techniques with those who are responsible for the evaluation of inventory differences at their facility. One of the most dangerous aspects of any evaluation technique, especially one as complex as most inventory difference evaluations tend to be, is to fail to look at the tools being used as indicators. There is a tendency to look at the results of an evaluation by one technique as an absolute. At the Los Alamos National Laboratory, several tools are used and the final evaluation is based on a combination of the observed results of a many-faceted evaluation. The tools used and some examples are presented

  3. 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

  4. Safety assessment methodology for waste repositories in deep geological formations

    International Nuclear Information System (INIS)

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

    1986-06-01

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

  5. 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.

  6. Development of Audit Calculation Methodology for RIA Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  7. Development of a new methodology for quantifying nuclear safety culture

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  8. 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.

  9. Methodology for flammable gas evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Hopkins, J.D., Westinghouse Hanford

    1996-06-12

    There are 177 radioactive waste storage tanks at the Hanford Site. The waste generates flammable gases. The waste releases gas continuously, but in some tanks the waste has shown a tendency to trap these flammable gases. When enough gas is trapped in a tank`s waste matrix, it may be released in a way that renders part or all of the tank atmosphere flammable for a period of time. Tanks must be evaluated against previously defined criteria to determine whether they can present a flammable gas hazard. This document presents the methodology for evaluating tanks in two areas of concern in the tank headspace:steady-state flammable-gas concentration resulting from continuous release, and concentration resulting from an episodic gas release.

  10. 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.

  11. 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

  12. 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

  13. MAPLE research reactor safety uncertainty assessment methodology

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  14. Integrated Deterministic-Probabilistic Safety Assessment Methodologies

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-01

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

  15. 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

  16. Evaluation of reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-04-15

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

  17. Evaluation of proliferation resistance using the INPRO methodology

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Park, Joo Hwan; Ko, Won Il; Song, Kee Chan; Choi, Kun Mo; Kim, Jin Kyoung

    2007-01-01

    The IAEA launched the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and developed the INPRO Methodology to provide guidelines and to assess the characteristics of a future innovative nuclear energy system in areas such as safety, economics, waste management, and proliferation resistance. The proliferation resistance area of the INPRO Methodology is reviewed here, and modifications for further improvements are proposed. The evaluation metrics including the evaluation parameters, evaluation scales and acceptance limits are developed for a practical application of the methodology to assess the proliferation resistance. The proliferation resistant characteristics of the DUPIC fuel cycle are assessed by applying the modified INPRO Methodology based on the developed evaluation metrics and acceptance criteria. The evaluation procedure and the metrics can be utilized as a reference for an evaluation of the proliferation resistance of a future innovative nuclear energy system

  18. The methodological seminar “Psychological Safety in Transport”

    Directory of Open Access Journals (Sweden)

    Sviridenko I.N.

    2018-03-01

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

  19. Evaluation of repository safety

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  20. Evaluation of repository safety

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    Science.gov (United States)

    1995-01-01

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

  2. An overview to CERSSO's self evaluation of the cost-benefit on the investment in occupational safety and health in the textile factories: "a step by step methodology".

    Science.gov (United States)

    Amador-Rodezno, Rafael

    2005-01-01

    The Pan American Health Organization (PAHO) and CERSSO collaborated to develop a new Tool Kit (TK), which became available in May 2002. PAHO already had a TK in place, and CERSSO requested that one be developed for their needs. CERSSO wanted to enable managers and line workers in garment factories to self-diagnose plant and workstation hazards and to estimate the costs and benefits of investing in occupational safety and health (OSH) as a way to improve productivity and competitiveness. For consistency, the collaborating organizations agreed to construct the TK according to PAHO's methodology. The instrument was developed to be comprehensive enough that any user can collect the data easily. It integrates epidemiologic, risk assessment, clinic, engineering, and accountability issues, organized to include step-by-step training in: (a) performing risk assessments in the workplaces (risk factors); (b) making cause-effect relationships; (c) improving decision making on OSH interventions; (d) doing calculations of direct and indirect costs and savings; and (e) doing calculation of the overall cost-benefit of OSH interventions. Since July 2002, about 2,400 employees and officials from 736 garment factories, Ministries of Labor, Health, Social Security Institutes, and Technical Training Institutions of Central America and the Dominican Republic have used this instrument. Systematically, they have calculated a positive relationship of the investment (3 to 33 times). Employers are now aware of the financial rewards of investing in OSH. The TK is available in Spanish, Korean, and English. In July 2003, a software program in Spanish and English was developed (180 persons have been trained in the region), which requires less time to execute with better reliability.

  3. WNP-2 outage safety review methodology

    International Nuclear Information System (INIS)

    Chiang, Albert; Fu, James

    2004-01-01

    A practical and versatile method was developed in the flow chart and checklist forms to show the defense-in-depth for various key safety functions of a nuclear power plant during shutdown. Using four different colors (green, yellow, orange, and red) for indication of levels of defense-in-depth is visually impressive, easy to understand, and was adopted by the outage management personnel as a convenient reference tool for maintenance activity planning before the outage, and schedule changes during the outage. This paper describes the method and its application at Washington Public Power Supply System's Nuclear Project 2 (WNP-2). (author)

  4. Overview of seismic re-evaluation methodologies

    International Nuclear Information System (INIS)

    Campbell, R.D.; Johnson, J.J.

    1993-01-01

    Several seismic licensing and safety issues have emerged over the past fifteen years for commercial U.S. Nuclear Power Plants and U.S. Government research reactors, production reactors and process facilities. The methodologies for resolution of these issues have been developed in numerous government and utility sponsored research programs. The resolution criteria have included conservative deterministic design criteria, deterministic seismic margins assessments criteria (SMA) and seismic probabilistic safety assessment criteria (SPSA). The criteria for SMAs and SPSAs have been based on realistically considering the inelastic energy absorption capability of ductile structures, equipment and piping and have incorporated the use of earthquake and testing experience to evaluate the operability of complex mechanical and electrical equipment. Most of the applications to date have been confined to the U.S. but there have been several applications to Asian, Western and Eastern Europe reactors. This paper summarizes the major issues addressed, the development of reevaluation criteria and selected applications to non U.S. reactors including WWER reactors. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  6. Performance Testing Methodology for Safety-Critical Programmable Logic Controller

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  8. Comparing in vivo biodistribution with radiolabeling and Franz cell permeation assay to validate the efficacy of both methodologies in the evaluation of nanoemulsions: a safety approach

    International Nuclear Information System (INIS)

    Cerqueira-Coutinho, C S; De Campo, V E B; Mansur, C R E; Rossi, A L; Veiga, V F; Holandino, C; Freitas, Z M F; Ricci-Junior, E; Santos, E P; Santos-Oliveira, R

    2016-01-01

    The Franz cells permeation assay has been performed for over 25 years. However, the advent of nanotechnology created a whole new world, especially with regard to topical products. In this new global scenario an increasing number of nanostructure-based delivery systems (NDSs) have emerged and a global warning relating to the safety of these NDSs is arising. This work studied the efficacy of the Franz cells assay, comparing it with the radiolabeling biodistribution test. For this purpose a formulation of sunscreen based on an NDS was developed and characterized. The results demonstrated both that the NDS did not present in vitro cytotoxicity and that the radiolabeling biodistribution test is more precise for the evaluation of NDS cosmetics than the Franz cells assay, since it detected the permeation of the NDS at a picogram order. Due to this fact, and considering all the concerns related to NDSs and nanoparticles in general, more precise methods must be used in order to guarantee the safe use of these new classes of products. (paper)

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

    International Nuclear Information System (INIS)

    Hwang-Yong, Jun; Yo-Han, Kim

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    2003-05-01

    The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

  11. Methodology of formal software evaluation

    International Nuclear Information System (INIS)

    Tuszynski, J.

    1998-01-01

    Sydkraft AB, the major Swedish utility, owner of ca 6000 MW el installed in nuclear (NPP Barsebaeck and NPP Oskarshamn), fossil fuel and hydro Power Plants is facing modernization of the control systems of the plants. Standards applicable require structured, formal methods for implementation of the control functions in the modem, real time software systems. This presentation introduces implementation methodology as discussed presently at the Sydkraft organisation. The approach suggested is based upon the process of co-operation of three parties taking part in the implementation; owner of the plant, vendor and Quality Assurance (QA) organisation. QA will be based on tools for formal software validation and on systematic gathering by the owner of validated and proved-by-operation control modules for the concern-wide utilisation. (author)

  12. Nuclear power plant simulation facility evaluation methodology

    International Nuclear Information System (INIS)

    Haas, P.M.; Carter, R.J.; Laughery, K.R. Jr.

    1985-01-01

    A methodology for evaluation of nuclear power plant simulation facilities with regard to their acceptability for use in the US Nuclear Regulatory Commission (NRC) operator licensing exam is described. The evaluation is based primarily on simulator fidelity, but incorporates some aspects of direct operator/trainee performance measurement. The panel presentation and paper discuss data requirements, data collection, data analysis and criteria for conclusions regarding the fidelity evaluation, and summarize the proposed use of direct performance measurment. While field testing and refinement of the methodology are recommended, this initial effort provides a firm basis for NRC to fully develop the necessary methodology

  13. 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

  14. 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.

  15. 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)

  16. Covariance Evaluation Methodology for Neutron Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.

    2008-09-01

    We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.

  17. Objectives of safety evaluation

    International Nuclear Information System (INIS)

    Rosen, M.

    1980-01-01

    An examination of the safety aspects of exported nuclear power plants demonstrates that additional and somewhat special considerations exist for these plants. In view of this and the generally small regulatory staffs of importing coutnries, suggestions are given for measures which should be taken by various organizations involved in the export and import of nuclear power facilities to raise the level of the very essential safety assessment. (orig.)

  18. Health economic evaluation: important principles and methodology.

    Science.gov (United States)

    Rudmik, Luke; Drummond, Michael

    2013-06-01

    To discuss health economic evaluation and improve the understanding of common methodology. This article discusses the methodology for the following types of economic evaluations: cost-minimization, cost-effectiveness, cost-utility, cost-benefit, and economic modeling. Topics include health-state utility measures, the quality-adjusted life year (QALY), uncertainty analysis, discounting, decision tree analysis, and Markov modeling. Economic evaluation is the comparative analysis of alternative courses of action in terms of both their costs and consequences. With increasing health care expenditure and limited resources, it is important for physicians to consider the economic impact of their interventions. Understanding common methodology involved in health economic evaluation will improve critical appraisal of the literature and optimize future economic evaluations. Copyright © 2012 The American Laryngological, Rhinological and Otological Society, Inc.

  19. Safety design and evaluation policy for future FBRs in Japan

    International Nuclear Information System (INIS)

    Aizawa, Kiyoto

    1991-01-01

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

  20. Framework of Comprehensive Proliferation Resistance Evaluation Methodology

    International Nuclear Information System (INIS)

    Kim, Min Su; Jo, Seong Youn; Kim, Min Soo; Kim, Jae San; Lee, Hyun Kyung

    2007-01-01

    Civilian nuclear programs can be used as a pretext to acquire technologies, materials, equipment for military weapon programs. Consequently, international society has a strong incentive to develop a nuclear system more proliferation resistant to assure that the civilian nuclear energy system is an unattractive and least desirable route for diversion of weapon usable material. The First step developing a more proliferation resistant nuclear energy system is to develop a systematic and standardized evaluation methodology to ensure that any future nuclear energy system satisfies the proliferation resistance goals. Many attempts to develop systematic evaluation methodology have been proposed and many systems for assessing proliferation resistance have been previously studied. However, a comprehensive proliferation resistance evaluation can not be achieved by simply applying one method since complicated proliferation resistance characteristics, including inherent features and extrinsic features, should be completely evaluated. Therefore, it is necessary to develop one incorporated evaluation methodology to make up for weak points of each evaluation method. The objective of this study is to provide a framework of comprehensive proliferation resistance evaluation methodology by incorporating two generally used evaluation methods, attribute and scenario analysis

  1. Methodology for quantitative evaluation of diagnostic performance

    International Nuclear Information System (INIS)

    Metz, C.

    1981-01-01

    Of various approaches that might be taken to the diagnostic performance evaluation problem, Receiver Operating Characteristic (ROC) analysis holds great promise. Further development of the methodology for a unified, objective, and meaningful approach to evaluating the usefulness of medical imaging procedures is done by consideration of statistical significance testing, optimal sequencing of correlated studies, and analysis of observer performance

  2. Quantified reliability and risk assessment methodology in safety evaluation and licensing: survey of practice and trends in E.C. countries; partial contribution in decision making, perpective of safety goals

    International Nuclear Information System (INIS)

    Vinck, W.F.

    1982-01-01

    Quantified reliability analysis of structures and systems and the quantified risk-concept is increasingly developed and applied in safety evaluation and in the licensing/regulatory process where deterministic approaches are however still predominant. A description of the types of application and a survey of the diversified opinions and the problem areas (e.g. the validity of input data, uncertainties in consequence modelling, human factors, common mode failures, etc.) are given. The significance of quantified risk assessment and comparisons, as one of the contributors in the solution to acceptability of modern technology such as nuclear power production, is discussed. Other contributions, such as benefit assessment and cost-efficiency of risk reduction, are also put into perspective within the decision-making process and in the problem of actual acceptance of new technologies. The growing need of developing and agreeing on overall safety objectives (how safe is safe enough) is finally discussed, in the light of the increasing diversity of approaches in the interconnected areas of accident hypotheses/sequences, siting parameters and technical bases for emergency planning; the latter problem being also closely connected to decisional processes for acceptability and to actual acceptance

  3. 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

  4. Methodology for evaluation of industrial CHP production

    International Nuclear Information System (INIS)

    Pavlovic, Nenad V.; Studovic, Milovan

    2000-01-01

    At the end of the century industry switched from exclusive power consumer into power consumer-producer which is one of the players on the deregulated power market. Consequently, goals of industrial plant optimization have to be changed, making new challenges that industrial management has to be faced with. In the paper is reviewed own methodology for evaluation of industrial power production on deregulated power market. The methodology recognizes economic efficiency of industrial CHP facilities as a main criterion for evaluation. Energy and ecological efficiency are used as additional criteria, in which implicit could be found social goals. Also, methodology recognizes key and limit factors for CHP production in industry. It could be successful applied, by use of available commercial software for energy simulation in CHP plants and economic evaluation. (Authors)

  5. Experiment to evaluate software safety

    International Nuclear Information System (INIS)

    Soubies, B.; Henry, J.Y.

    1994-01-01

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

  6. 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

  7. Review and evaluation of paleohydrologic methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Foley, M.G.; Zimmerman, D.A.; Doesburg, J.M.; Thorne, P.D.

    1982-12-01

    A literature review was conducted to identify methodologies that could be used to interpret paleohydrologic environments. Paleohydrology is the study of past hydrologic systems or of the past behavior of an existing hydrologic system. The purpose of the review was to evaluate how well these methodologies could be applied to the siting of low-level radioactive waste facilities. The computer literature search queried five bibliographical data bases containing over five million citations of technical journals, books, conference papers, and reports. Two data-base searches (United States Geological Survey - USGS) and a manual search were also conducted. The methodologies were examined for data requirements and sensitivity limits. Paleohydrologic interpretations are uncertain because of the effects of time on hydrologic and geologic systems and because of the complexity of fluvial systems. Paleoflow determinations appear in many cases to be order-of-magnitude estimates. However, the methodologies identified in this report mitigate this uncertainty when used collectively as well as independently. That is, the data from individual methodologies can be compared or combined to corroborate hydrologic predictions. In this manner, paleohydrologic methodologies are viable tools to assist in evaluating the likely future hydrology of low-level radioactive waste sites.

  8. Review and evaluation of paleohydrologic methodologies

    International Nuclear Information System (INIS)

    Foley, M.G.; Zimmerman, D.A.; Doesburg, J.M.; Thorne, P.D.

    1982-12-01

    A literature review was conducted to identify methodologies that could be used to interpret paleohydrologic environments. Paleohydrology is the study of past hydrologic systems or of the past behavior of an existing hydrologic system. The purpose of the review was to evaluate how well these methodologies could be applied to the siting of low-level radioactive waste facilities. The computer literature search queried five bibliographical data bases containing over five million citations of technical journals, books, conference papers, and reports. Two data-base searches (United States Geological Survey - USGS) and a manual search were also conducted. The methodologies were examined for data requirements and sensitivity limits. Paleohydrologic interpretations are uncertain because of the effects of time on hydrologic and geologic systems and because of the complexity of fluvial systems. Paleoflow determinations appear in many cases to be order-of-magnitude estimates. However, the methodologies identified in this report mitigate this uncertainty when used collectively as well as independently. That is, the data from individual methodologies can be compared or combined to corroborate hydrologic predictions. In this manner, paleohydrologic methodologies are viable tools to assist in evaluating the likely future hydrology of low-level radioactive waste sites

  9. 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 ...

  10. Hybrid probabilistic and possibilistic safety assessment. Methodology and application

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  11. Methodology for evaluation of diagnostic performance

    International Nuclear Information System (INIS)

    Metz, C.E.

    1992-01-01

    Effort in this project during the past year has focused on the development, refinement, and distribution of computer software that will allow current Receiver Operating Characteristic (ROC) methodology to be used conveniently and reliably by investigators in a variety of evaluation tasks in diagnostic medicine; and on the development of new ROC methodology that will broaden the spectrum of evaluation tasks and/or experimental settings to which the fundamental approach can be applied. Progress has been limited by the amount of financial support made available to the project

  12. Demonstration of an infiltration evaluation methodology

    International Nuclear Information System (INIS)

    Smyth, J.D.; Gee, G.W.; Kincaid, C.T.; Nichols, W.M.; Bresler, E.

    1990-07-01

    An Infiltration Evaluation Methodology (IEM) was developed for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL) to provide a consistent, well formulated approach for evaluating drainage through engineered covers at low-level radioactive waste (LLW) sites. The methodology is designed to help evaluate the ability of proposed waste site covers to minimize drainage for LLW site license applications and for sites associated with the Uranium Mill Tailings Remedial Action (UMTRA) program. The objective of this methodology is to estimate the drainage through an engineered burial site cover system. The drainage estimate can be used as an input to a broader performance assessment methodology currently under development by the NRC. The methodology is designed to simulate, at the field scale, significant factors and hydrologic conditions which determine or influence estimates of infiltration, long-term moisture content profiles, and drainage from engineered covers and barriers. The IEM developed under this study acknowledges the uncertainty inherent in soil properties and quantifies the influence of such uncertainty on the estimates of drainage in engineered cover systems at waste disposal sites. 6 refs., 1 fig

  13. 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)

  14. 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)

  15. 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)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  18. Status of Methodology Development for the Evaluation of Proliferation Resistance

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Ko, Won Il; Lee, Jung Won

    2010-01-01

    Concerning the increasing energy demand and green house effect, nuclear energy is now the most feasible option. Therefore, recently, oil countries even have a plan to build the nuclear power plant for energy production. If nuclear systems are to make a major and sustainable contribution to the worlds energy supply, future nuclear energy systems must meet specific requirements. One of the requirements is to satisfy the proliferation resistance condition in an entire nuclear system. Therefore, from the beginning of future nuclear energy system development, it is important to consider a proliferation resistance to prevent the diversion of nuclear materials. The misuse of a nuclear system must be considered as well. Moreover, in the import and export of nuclear system, the evaluation of the proliferation resistance on the nuclear system becomes a key factor The INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles) program initiated by the IAEA proposed proliferation resistance (PR) as a key component of a future innovative nuclear system (INS) with a sustainability, economics, safety of nuclear installation and waste management. The technical goal for Generation IV (Gen IV) nuclear energy systems (NESs) highlights a Proliferation Resistance and Physical Protection (PR and PP), sustainability, safety, reliability and economics as well. Based on INPRO and Gen IV study, the methodology development for the evaluation of proliferation resistance has been carried out in KAERI. Finally, the systematic procedure for methodology was setup and the indicators for the procedure were decided. The methodology involves the evaluation from total nuclear system to individual process. Therefore, in this study, the detailed procedure for the evaluation of proliferation resistance and the newly proposed additional indicators are described and several conditions are proposed to increase the proliferation resistance in the future nuclear system. The assessment of PR

  19. Gamma ray auto absorption correction evaluation methodology

    International Nuclear Information System (INIS)

    Gugiu, Daniela; Roth, Csaba; Ghinescu, Alecse

    2010-01-01

    Neutron activation analysis (NAA) is a well established nuclear technique, suited to investigate the microstructural or elemental composition and can be applied to studies of a large variety of samples. The work with large samples involves, beside the development of large irradiation devices with well know neutron field characteristics, the knowledge of perturbing phenomena and adequate evaluation of correction factors like: neutron self shielding, extended source correction, gamma ray auto absorption. The objective of the works presented in this paper is to validate an appropriate methodology for gamma ray auto absorption correction evaluation for large inhomogeneous samples. For this purpose a benchmark experiment has been defined - a simple gamma ray transmission experiment, easy to be reproduced. The gamma ray attenuation in pottery samples has been measured and computed using MCNP5 code. The results show a good agreement between the computed and measured values, proving that the proposed methodology is able to evaluate the correction factors. (authors)

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

    Science.gov (United States)

    1995-09-01

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

  1. A study on safety assessment methodology for a vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-15

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

  2. Common methodological flaws in economic evaluations.

    Science.gov (United States)

    Drummond, Michael; Sculpher, Mark

    2005-07-01

    Economic evaluations are increasingly being used by those bodies such as government agencies and managed care groups that make decisions about the reimbursement of health technologies. However, several reviews of economic evaluations point to numerous deficiencies in the methodology of studies or the failure to follow published methodological guidelines. This article, written for healthcare decision-makers and other users of economic evaluations, outlines the common methodological flaws in studies, focussing on those issues that are likely to be most important when deciding on the reimbursement, or guidance for use, of health technologies. The main flaws discussed are: (i) omission of important costs or benefits; (ii) inappropriate selection of alternatives for comparison; (iii) problems in making indirect comparisons; (iv) inadequate representation of the effectiveness data; (v) inappropriate extrapolation beyond the period observed in clinical studies; (vi) excessive use of assumptions rather than data; (vii) inadequate characterization of uncertainty; (viii) problems in aggregation of results; (ix) reporting of average cost-effectiveness ratios; (x) lack of consideration of generalizability issues; and (xi) selective reporting of findings. In each case examples are given from the literature and guidance is offered on how to detect flaws in economic evaluations.

  3. Methodologies for evaluation of AECB regulatory program

    International Nuclear Information System (INIS)

    Yarranton, G.A.; Gray, B.J.; Yarranton, M.

    1986-05-01

    AECB (Atomic Energy Control Board) commissioned this report to obtain information about methods of planning and conducting evaluation of its regulatory program. The report begins with a bibliography consisting of 280 abstracts assembled from an extensive search of international literature. Each cited publication describes or uses methods applicable to the evaluation of regulatory programs. The report continues with a review of the methodologies found in the literature. It identifies the most relevant references for each step in program evaluation: the commissioning of evaluation; the identification of evaluation issues; the defining of questions; the answering of questions; the reporting of reslts, and the implementation of recommendations. Finally, the report examines the applicability, advantages and disadvantages of the different evaluation methods and makes recommendations about the selection of methods and their application to the AECB program

  4. Inclusion of Premeditated Threats in the Safety Methodology for NPPs

    International Nuclear Information System (INIS)

    Levanon, I.

    2014-01-01

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

  5. Economic evaluations of occupational health interventions from a corporate perspective - A systematic review of methodological quality

    NARCIS (Netherlands)

    Uegaki, K.; Bruijne, M.C. de; Lambeek, L.; Anema, J.R.; Beek, A.J. van der; Mechelen, W. van; Tulder, M.W. van

    2010-01-01

    Objective: Using a standardized quality criteria list, we appraised the methodological quality of economic evaluations of occupational safety and health (OSH) interventions conducted from a corporate perspective. Methods: The primary literature search was conducted in Medline and Embase.

  6. The Evaluation Methodology of Information Support

    Directory of Open Access Journals (Sweden)

    Lubos Necesal

    2016-01-01

    Full Text Available Knowledge, information and people are the motive force in today's organizations. Successful organizations need to find the right employees and provide them with the right and highquality information. This is a complex problem. In the world where information plays more and more important role, employees have to be skilled at information activities (searching, processing, saving, etc. of information and information system/-s (IS they work with. Organizations have to cover both these areas. Therefore, we need an effective instrument, which could be used to evaluate new employees within admission or as regular evaluating of current employees, to evaluate information system, whether it is an appropriate tool for fulfilling the employee’s tasks within the organization, and to evaluate how the organization covers the foregoing areas. Such instrument is the “Evaluation methodology of information support in organization”. This paper defines the term “information support“ and its role in organization. The body of the paper proposes the “Evaluation methodology of information support in organization”. The conclusion discusses contributions of information support evaluation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  9. 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

  10. Research methodological issues in evaluating herbal interventions

    Directory of Open Access Journals (Sweden)

    Dipika Bansal

    2010-02-01

    Full Text Available Dipika Bansal, Debasish Hota, Amitava ChakrabartiPostgraduate Institute of Medical Education and Research, Chandigarh, IndiaAbstract: Randomized controlled trials provide the best evidence, and is seen as the gold standard for allopathic research. Herbal therapies are not an integral part of conventional care although they are still used by patients in their health care management. These medicines need to be subjected to rigorous research to establish their effectiveness and safety. Clearly defined treatments are required and should be recorded in a manner that enables other suitably trained researchers to reproduce them reliably. Quality control of herbal products is also a prerequisite of credible clinical trials. Methodological strategies for investigating the herbal interventions and the issues regarding appropriate patient selection, randomization and blinding, placebo effects and choice of comparator, occupational standardization and the selection of appropriate study endpoints to prove efficacy are being discussed. This paper will review research options and propose some suggestions for future research design.Keywords: CAM research, herbal therapies, methodology, clinical trial

  11. Unification of the methodology of competitiveness evaluation

    Directory of Open Access Journals (Sweden)

    Ivanova Marina

    2016-05-01

    Full Text Available The purpose of the study is the comparison of the modern methods of competitiveness evaluation and testing the results in practice at a production plant in order to unify the methodology. Analysis of the publications shows that the existing diversity in methods of the competitiveness evaluation complicates the choice of a method and interpretation of the results, which confirms the need to classify and unify the methods available today. The authors suggest that the standardized method of competitiveness evaluation should combine matrix and graphical methods. In our study, we have shown that SWOT-analysis should be carried out in two stages: firstly, it is necessary to determine the strengths and weaknesses of the company, taking into account the following factors: marketing, production, economic and financial situation, technology, the administration and personnel; secondly, it is essential to assess the opportunities and threats with the allowance for the political, legal, and economic situation, the impact of market as well as social, cultural and natural environmental factors. The proposed methodology of competitiveness evaluation has been tested at the industrial enterprise "Olvita ltd", which is dynamically developing and successfully operates on the food market of Ukraine; it specializes in processing, logistics and distribution of fresh and frozen vegetables, fruits and berries, as well as ready-to cook meat products. The results have shown that the competitiveness evaluation remains a pressing issue that requires simple and fairly demonstrative methods, one of which is a method that combines both graphical and matrix approaches. The advantage of the proposed method is its sufficiently high visualizations and record of personal approaches in evaluating the scores for strengths and weaknesses as well as opportunities and threats. In the future, the authors are planning to examine and systemize the indicators used in the competitiveness

  12. 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.

  13. 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

  14. 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…

  15. Safety evaluation of food flavorings

    International Nuclear Information System (INIS)

    Schrankel, Kenneth R.

    2004-01-01

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

  16. 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)

  17. Evaluation of periodic safety status analyses

    International Nuclear Information System (INIS)

    Faber, C.; Staub, G.

    1997-01-01

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

  18. 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.

  19. 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.

  20. 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

  1. Towards a Public Sector GIS Evaluation Methodology | Kurwakumire ...

    African Journals Online (AJOL)

    Towards a Public Sector GIS Evaluation Methodology. ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search · USING AJOL · RESOURCES ... However, evaluation methodologies for public sector GIS are largely lacking.

  2. Evaluation methodology for generator refurbishment decisions

    International Nuclear Information System (INIS)

    Moore, W.G.; Ulm, S.F.

    1991-01-01

    The Electrical Power Industry is undergoing tremendous change due to deregulation, aging equipment, environmental concerns, and investment/risk considerations. Public utility commissions, along with shareholders and end consumers, are closely monitoring utilities; decisions, especially in the area of costs-both Operation and Maintenance, and Capital. Increasing emphasis, within the conventional utility environment, has been and continue to be, placed on controlling expenditures. To be responsive to these industry and competitive pressures, utilities must make equipment refurbishment decisions. These decisions should be based on input from many sources, including the severity of the failure, cost of replacement versus refurbishment, risks and safety considerations, the expected remaining life of the unit, operational mode (base or peak), fuel type, initial costs, system capacity, available budgets, and financing options. Many times, however, refurbishment decisions are base don an abstract understanding of the above, but feel, or emotional attachment to a particular option. This paper describes a general methodology for refurbishment decision making, applied specifically to generators. Also included in a case history of one utility's progression through this process

  3. The study of evaluation methodology of the aging and degradation researches

    International Nuclear Information System (INIS)

    Cho, C. J.; Park, Z. H.; Jeong, I. S.

    2001-01-01

    To judge the usefulness of aging related researches like PLIM (Plant lifetime Management) and aging related degradation, et. al. in PSR(Periodic Safety Review), the evaluation methodology of the R and D have been proposed up to now are reviewed. The infometric methodology is considered to be the optimum method for the evaluation of the nuclear related researches. And finally, to increase the objectiveness and reliability of the infometric methodology in the aging and degradation researches, the indexes of safety, technology and economics are introduced. From this study, the infometric methodology has the advantage of the actual engineering evaluation in the nuclear related researches with other methodologies, but for the further research, the effective construction of DB and survey of various statistics in the technical reports and papers are needed

  4. Safety climate and attitude as evaluation measures of organizational safety.

    Science.gov (United States)

    Isla Díaz, R; Díaz Cabrera, D

    1997-09-01

    The main aim of this research is to develop a set of evaluation measures for safety attitudes and safety climate. Specifically it is intended: (a) to test the instruments; (b) to identify the essential dimensions of the safety climate in the airport ground handling companies; (c) to assess the quality of the differences in the safety climate for each company and its relation to the accident rate; (d) to analyse the relationship between attitudes and safety climate; and (e) to evaluate the influences of situational and personal factors on both safety climate and attitude. The study sample consisted of 166 subjects from three airport companies. Specifically, this research was centered on ground handling departments. The factor analysis of the safety climate instrument resulted in six factors which explained 69.8% of the total variance. We found significant differences in safety attitudes and climate in relation to type of enterprise.

  5. Econometric Methodology of Monopolization Process Evaluation

    Directory of Open Access Journals (Sweden)

    Dmitrijs Skoruks

    2014-06-01

    Full Text Available The research “Econometric Methodology of Monopolization Process Evaluation” gives a perspective description of monopolization process’ nature, occurrence source, development procedure and internal conjuncture specifics, as well as providing an example of modern econometrical method application within a unified framework of market competition analysis for the purpose of conducting a quantitative competition evaluation on an industry level for practical use in both private and public sectors. The main question of the aforementioned research is the definition and quantitative analysis of monopolization effects in modern day globalized markets, while con- structing an empirical model of the econometric analysis, based on the use of in- ternational historical experience of monopoly formations standings, with the goal of introducing a further development scheme for the use of both econometrical and statistical instruments in line with the forecasting and business research need of enterprises and regulatory functions of the public sector. The current research uses a vast variety of monopolization evaluation ratios and their econometrical updates on companies that are involved in the study procedure in order to detect and scallar measure their market monopolizing potential, based on the implemented acquired market positions, turnover shares and competition policies.

  6. 2005 dossier: clay. Tome: safety evaluation of the geologic disposal

    International Nuclear Information System (INIS)

    2005-01-01

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

  7. 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)

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

    Directory of Open Access Journals (Sweden)

    Franco Pierro

    2009-01-01

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

  9. Evaluation methodologies for security testing biometric systems beyond technological evaluation

    OpenAIRE

    Fernández Saavedra, María Belén

    2013-01-01

    The main objective of this PhD Thesis is the specification of formal evaluation methodologies for testing the security level achieved by biometric systems when these are working under specific contour conditions. This analysis is conducted through the calculation of the basic technical biometric system performance and its possible variations. To that end, the next two relevant contributions have been developed. The first contribution is the definition of two independent biometric performance ...

  10. Methodology for evaluation of railroad technology research projects

    Science.gov (United States)

    1981-04-01

    This Project memorandum presents a methodology for evaluating railroad research projects. The methodology includes consideration of industry and societal benefits, with special attention given to technical risks, implementation considerations, and po...

  11. EVALUATION OF THE GRAI INTEGRATED METHODOLOGY AND THE IMAGIM SUPPORTWARE

    Directory of Open Access Journals (Sweden)

    J.M.C. Reid

    2012-01-01

    Full Text Available This paper describes the GRAI Integrated Methodology and identifies the need for computer tools to support enterprise modelling,design and integration. The IMAGIM tool is then evaluated in terms of its ability to support the GRAI Integrated Methodology. The GRAI Integrated Methodology is an Enterprise Integration methodology developed to support the design of CIM systems . The GRAI Integrated Methodology consists of the GRAI model and a structured approach. The latest addition to the methodology is the IMAGIM software tool developed by the GRAI research group for the specific purpose of supporting the methodology.

  12. 10CFR50.59 safety evaluations

    International Nuclear Information System (INIS)

    Grime, L.; Page, E.

    1987-01-01

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

  13. A methodological study to compare survey-based and observation-based evaluations of organisational and safety cultures and then compare both approaches with markers of the quality of care.

    Science.gov (United States)

    Freeth, D; Sandall, J; Allan, T; Warburton, F; Berridge, E J; Mackintosh, N; Rogers, M; Abbott, S

    2012-05-01

    Patient safety concerns have focused attention on organisational and safety cultures, in turn directing attention to the measurement of organisational and safety climates. First, to compare levels of agreement between survey- and observation-based measures of organisational and safety climates/cultures and to compare both measures with criterion-based audits of the quality of care, using evidence-based markers drawn from national care standards relating to six common clinical conditions. (This required development of an observation-based instrument.) Second, to examine whether observation-based evaluations could replace or augment survey measurements to mitigate concerns about declining response rates and increasing social desirability bias. Third, to examine mediating factors in safety and organisational climate scores. The study had three strands: (A) a postal questionnaire survey to elicit staff perceptions of organisational and safety climates, using six prevalidated scales; (B) semistructured non-participant observation of clinical teams; and (C) a retrospective criterion-based audit carried out by non-clinical auditors to minimise hindsight bias. Standardised summary scores were created for each strand, and pairs of measurements were compared (strand A with strand B, strand A with strand C, and strand B with strand C) using Bland-Altman plots to evaluate agreement. Correlations were also examined. Multilevel modelling of Strand A scores explored mediating factors. Eight emergency departments and eight maternity units in England, UK. None. Examination of feasibility, correlation and agreement. Strand A: the overall response rate was 27.6%, whereas site-specific rates ranged from 9% to 47%. We identified more mediating factors than previous studies; variable response rates had little effect on the results. Organisational and safety climate scores were strongly correlated (r = 0.845) and exhibited good agreement [standard deviation (SD) differences 0.449; 14 (88

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    T. A. Bjornard; M. D. Zentner

    2006-01-01

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

  16. 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.

  17. 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

  18. Criticality safety evaluation in Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

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

    2000-04-01

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

  19. Development of a Seismic Setpoint Calculation Methodology Using a Safety System Approach

    International Nuclear Information System (INIS)

    Lee, Chang Jae; Baik, Kwang Il; Lee, Sang Jeong

    2013-01-01

    The Automatic Seismic Trip System (ASTS) automatically actuates reactor trip when it detects seismic activities whose magnitudes are comparable to a Safe Shutdown Earthquake (SSE), which is the maximum hypothetical earthquake at the nuclear power plant site. To ensure that the reactor is tripped before the magnitude of earthquake exceeds the SSE, it is crucial to reasonably determine the seismic setpoint. The trip setpoint and allowable value for the ASTS for Advanced Power Reactor (APR) 1400 Nuclear Power Plants (NPPs) were determined by the methodology presented in this paper. The ASTS that trips the reactor when a large earthquake occurs is categorized as a non safety system because the system is not required by design basis event criteria. This means ASTS has neither specific analytical limit nor dedicated setpoint calculation methodology. Therefore, we developed the ASTS setpoint calculation methodology by conservatively considering that of PPS. By incorporating the developed methodology into the ASTS for APR1400, the more conservative trip setpoint and allowable value were determined. In addition, the ZPA from the Operating Basis Earthquake (OBE) FRS of the floor where the sensor module is located is 0.1g. Thus, the allowance of 0.17g between OBE of 0.1 g and ASTS trip setpoint of 0.27 g is sufficient to prevent the reactor trip before the magnitude of the earthquake exceeds the OBE. In result, the developed ASTS setpoint calculation methodology is evaluated as reasonable in both aspects of the safety and performance of the NPPs. This will be used to determine the ASTS trip setpoint and allowable for newly constructed plants

  20. 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

  1. Evaluating safety management system implementation

    International Nuclear Information System (INIS)

    Preuss, M.

    2009-01-01

    Canada is committed to not only maintaining, but also improving upon our record of having one of the safest aviation systems in the world. The development, implementation and maintenance of safety management systems is a significant step towards improving safety performance. Canada is considered a world leader in this area and we are fully engaged in implementation. By integrating risk management systems and business practices, the aviation industry stands to gain better safety performance with less regulatory intervention. These are important steps towards improving safety and enhancing the public's confidence in the safety of Canada's aviation system. (author)

  2. A methodology for spectral wave model evaluation

    Science.gov (United States)

    Siqueira, S. A.; Edwards, K. L.; Rogers, W. E.

    2017-12-01

    Model evaluation is accomplished by comparing bulk parameters (e.g., significant wave height, energy period, and mean square slope (MSS)) calculated from the model energy spectra with those calculated from buoy energy spectra. Quality control of the observed data and choice of the frequency range from which the bulk parameters are calculated are critical steps in ensuring the validity of the model-data comparison. The compared frequency range of each observation and the analogous model output must be identical, and the optimal frequency range depends in part on the reliability of the observed spectra. National Data Buoy Center 3-m discus buoy spectra are unreliable above 0.3 Hz due to a non-optimal buoy response function correction. As such, the upper end of the spectrum should not be included when comparing a model to these data. Bioufouling of Waverider buoys must be detected, as it can harm the hydrodynamic response of the buoy at high frequencies, thereby rendering the upper part of the spectrum unsuitable for comparison. An important consideration is that the intentional exclusion of high frequency energy from a validation due to data quality concerns (above) can have major implications for validation exercises, especially for parameters such as the third and fourth moments of the spectrum (related to Stokes drift and MSS, respectively); final conclusions can be strongly altered. We demonstrate this by comparing outcomes with and without the exclusion, in a case where a Waverider buoy is believed to be free of biofouling. Determination of the appropriate frequency range is not limited to the observed spectra. Model evaluation involves considering whether all relevant frequencies are included. Guidance to make this decision is based on analysis of observed spectra. Two model frequency lower limits were considered. Energy in the observed spectrum below the model lower limit was calculated for each. For locations where long swell is a component of the wave

  3. ECSIN's methodological approach for hazard evaluation of engineered nanomaterials

    Science.gov (United States)

    Bregoli, Lisa; Benetti, Federico; Venturini, Marco; Sabbioni, Enrico

    2013-04-01

    The increasing production volumes and commercialization of engineered nanomaterials (ENM), together with data on their higher biological reactivity when compared to bulk counterpart and ability to cross biological barriers, have caused concerns about their potential impacts on the health and safety of both humans and the environment. A multidisciplinary component of the scientific community has been called to evaluate the real risks associated with the use of products containing ENM, and is today in the process of developing specific definitions and testing strategies for nanomaterials. At ECSIN we are developing an integrated multidisciplinary methodological approach for the evaluation of the biological effects of ENM on the environment and human health. While our testing strategy agrees with the most widely advanced line of work at the European level, the choice of methods and optimization of protocols is made with an extended treatment of details. Our attention to the methodological and technical details is based on the acknowledgment that the innovative characteristics of matter at the nano-size range may influence the existing testing methods in a partially unpredictable manner, an aspect which is frequently recognized at the discussion level but oftentimes disregarded at the laboratory bench level. This work outlines the most important steps of our testing approach. In particular, each step will be briefly discussed in terms of potential technical and methodological pitfalls that we have encountered, and which are often ignored in nanotoxicology research. The final aim is to draw attention to the need of preliminary studies in developing reliable tests, a crucial aspect to confirm the suitability of the chosen analytical and toxicological methods to be used for the specific tested nanoparticle, and to express the idea that in nanotoxicology,"devil is in the detail".

  4. ECSIN's methodological approach for hazard evaluation of engineered nanomaterials

    International Nuclear Information System (INIS)

    Bregoli, Lisa; Benetti, Federico; Venturini, Marco; Sabbioni, Enrico

    2013-01-01

    The increasing production volumes and commercialization of engineered nanomaterials (ENM), together with data on their higher biological reactivity when compared to bulk counterpart and ability to cross biological barriers, have caused concerns about their potential impacts on the health and safety of both humans and the environment. A multidisciplinary component of the scientific community has been called to evaluate the real risks associated with the use of products containing ENM, and is today in the process of developing specific definitions and testing strategies for nanomaterials. At ECSIN we are developing an integrated multidisciplinary methodological approach for the evaluation of the biological effects of ENM on the environment and human health. While our testing strategy agrees with the most widely advanced line of work at the European level, the choice of methods and optimization of protocols is made with an extended treatment of details. Our attention to the methodological and technical details is based on the acknowledgment that the innovative characteristics of matter at the nano-size range may influence the existing testing methods in a partially unpredictable manner, an aspect which is frequently recognized at the discussion level but oftentimes disregarded at the laboratory bench level. This work outlines the most important steps of our testing approach. In particular, each step will be briefly discussed in terms of potential technical and methodological pitfalls that we have encountered, and which are often ignored in nanotoxicology research. The final aim is to draw attention to the need of preliminary studies in developing reliable tests, a crucial aspect to confirm the suitability of the chosen analytical and toxicological methods to be used for the specific tested nanoparticle, and to express the idea that in nanotoxicology,'devil is in the detail'.

  5. Final report of the accident phenomenology and consequence (APAC) methodology evaluation. Spills Working Group

    Energy Technology Data Exchange (ETDEWEB)

    Brereton, S.; Shinn, J. [Lawrence Livermore National Lab., CA (United States); Hesse, D [Battelle Columbus Labs., OH (United States); Kaninich, D. [Westinghouse Savannah River Co., Aiken, SC (United States); Lazaro, M. [Argonne National Lab., IL (United States); Mubayi, V. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The Spills Working Group was one of six working groups established under the Accident Phenomenology and Consequence (APAC) methodology evaluation program. The objectives of APAC were to assess methodologies available in the accident phenomenology and consequence analysis area and to evaluate their adequacy for use in preparing DOE facility safety basis documentation, such as Basis for Interim Operation (BIO), Justification for Continued Operation (JCO), Hazard Analysis Documents, and Safety Analysis Reports (SARs). Additional objectives of APAC were to identify development needs and to define standard practices to be followed in the analyses supporting facility safety basis documentation. The Spills Working Group focused on methodologies for estimating four types of spill source terms: liquid chemical spills and evaporation, pressurized liquid/gas releases, solid spills and resuspension/sublimation, and resuspension of particulate matter from liquid spills.

  6. Safety evaluation of advance street name signs

    Science.gov (United States)

    2009-06-01

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

  7. A methodology for nuclear power plant operational events evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Jeferson, E-mail: jeferson@cnen.gov.br [Comissao Nacional de Energia Nuclear (CGRC/CNEN), Rio de janeiro, RJ (Brazil). Coordenacao Geral de Reatores e do Ciclo de Combustivel; Costa, Sergio Dias, E-mail: sergiodiascosta@hotmail.com [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    to identify trends that might pass unnoticed. There are several methodologies for evaluation of operational events, specifically, for the determination of the causes of the event, where the concepts of root cause, direct cause and causal factors, among others. However, the most recent methodology in use in Brazil, date of 2003. The subject is so important that there is national legislation dealing with the theme, also it need of update. The actual regulations establishes criteria for notification of the occurrence of events considered significant for safety and establishes criteria and deadlines for the issuance of the report. Is also discussed the relationship between occurrence of operational events and a potential application for the life extension of the nuclear power plants, considering the regulatory focus, theme quite present in the international field. The present study intends to present a modern, appropriate methodology for evaluation of operational events, focusing on regulatory aspects and verify its applicability to nuclear power plants. Will also be presented a study of the occurrence of events, focusing on the last five years of operation. Finally will be also presented to international trends in development on this area. (author)

  8. A methodology for nuclear power plant operational events evaluation

    International Nuclear Information System (INIS)

    Araujo, Jeferson

    2015-01-01

    to identify trends that might pass unnoticed. There are several methodologies for evaluation of operational events, specifically, for the determination of the causes of the event, where the concepts of root cause, direct cause and causal factors, among others. However, the most recent methodology in use in Brazil, date of 2003. The subject is so important that there is national legislation dealing with the theme, also it need of update. The actual regulations establishes criteria for notification of the occurrence of events considered significant for safety and establishes criteria and deadlines for the issuance of the report. Is also discussed the relationship between occurrence of operational events and a potential application for the life extension of the nuclear power plants, considering the regulatory focus, theme quite present in the international field. The present study intends to present a modern, appropriate methodology for evaluation of operational events, focusing on regulatory aspects and verify its applicability to nuclear power plants. Will also be presented a study of the occurrence of events, focusing on the last five years of operation. Finally will be also presented to international trends in development on this area. (author)

  9. Status of Nuclear Safety evaluation in China

    International Nuclear Information System (INIS)

    Tian Jiashu

    1999-01-01

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

  10. 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

  11. Evaluation on safety issues of SMART

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. 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.

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. Methodology of evaluation of value created in the productive processes

    OpenAIRE

    M.T. Roszak

    2008-01-01

    Purpose: Of this paper was to present the methodology of analysis of the productive processes with applicationof value analysis and multi-criterion-analysis which allow to evaluate the technology and organization of theproductive processes.Design/methodology/approach: Presented in the paper methodology of evaluation of the productive processesis based on analysis of activities in the productive processes and their characteristics with reference to createdvalue in the productive chain.Findings...

  18. 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

  19. Stress Optical Coefficient, Test Methodology, and Glass Standard Evaluation

    Science.gov (United States)

    2016-05-01

    ARL-TN-0756 ● MAY 2016 US Army Research Laboratory Stress Optical Coefficient, Test Methodology , and Glass Standard Evaluation...Stress Optical Coefficient, Test Methodology , and Glass Standard Evaluation by Clayton M Weiss Oak Ridge Institute for Science and Education...ORISE), Belcamp, MD Parimal J Patel Weapons and Materials Research Directorate, ARL Approved for public release; distribution is

  20. Evaluation Methodologies for Estimating the Likelihood of Program Implementation Failure

    Science.gov (United States)

    Durand, Roger; Decker, Phillip J.; Kirkman, Dorothy M.

    2014-01-01

    Despite our best efforts as evaluators, program implementation failures abound. A wide variety of valuable methodologies have been adopted to explain and evaluate the "why" of these failures. Yet, typically these methodologies have been employed concurrently (e.g., project monitoring) or to the post-hoc assessment of program activities.…

  1. TAPS safety evaluation criteria for reload fueling

    International Nuclear Information System (INIS)

    Mahendra Nath; Veeraraghavan, N.

    1976-01-01

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

  2. An Evaluation Methodology for Protocol Analysis Systems

    Science.gov (United States)

    2007-03-01

    Main Memory Requirement NS: Needham-Schroeder NSL: Needham-Schroeder-Lowe OCaml : Objective Caml POSIX: Portable Operating System...methodology is needed. A. PROTOCOL ANALYSIS FIELD As with any field, there is a specialized language used within the protocol analysis community. Figure...ProVerif requires that Objective Caml ( OCaml ) be installed on the system, OCaml version 3.09.3 was installed. C. WINDOWS CONFIGURATION OS

  3. Parameters Evaluation of PLC Dependability and Safety

    Directory of Open Access Journals (Sweden)

    Juraj Zdansky

    2006-01-01

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

  4. Proposal of risk evaluation methodology for hazardous materials transportation

    International Nuclear Information System (INIS)

    Hartman, Luiz Carlos

    2009-01-01

    The increasing concern with the level of risk associated with the transportation of hazardous materials took some international institutions to pledge efforts in the evaluation of risk in regional level. Following this trend, the objective of this work was to analyze the most recent processes of analysis of risks from road transportation of hazardous materials. In the present work 21 methodologies of analysis of risks, developed by some authors and for diverse localities have been evaluated. Two of them, in special, have been reviewed and discussed: a method recently developed by the Swiss Federal Institute of Technology (Nicolet-Monnier and Gheorghe, 1996) and the strategy delineated by the Center for Chemical Process Safety CCPS (1995), taking into consideration the estimate of the individual and social risk. Also, the models of Harwood et al. (1990) and of Ramos (1997), adapted by Hartman (2003) have been applied to the reality of the roads of the state of Sao Paulo. The extension of these methodologies was explored, in order to find its advantages and disadvantages. As a study case the present work considered the ammonia transportation throughout two routes evaluating the reality of the roads of the state of Sao Paulo, including a significant parcel of evaluation in a densely populated area, getting the results using risk, at least, one of the methodologies mentioned above. The innovation proposed by this work was the research, the development and the introduction of two variables to the model considered by Harwood et al. (1990). These variables that influence in the value of the risk are: the age of the driver of truck and the zone of impact that is function type of product, period of the day where the transport was carried and the volume that has been transported. The aim of the proposed modifications is to let the value of the risk more sensible in relation to the type of the product carried and the age of the truck driver. The main related procedural stages

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

    Science.gov (United States)

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

    2012-06-01

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

  6. 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...

  7. 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...

  8. Methodology of site generation for evaluation of the behaviour of radioactive waste storage

    International Nuclear Information System (INIS)

    Ruiz Rivas, C.; Eguilior Diez, S.

    1997-01-01

    The present report summarizes the purpose of methodology for the site generation in the evaluation of high-level radioactive waste storage for long-term. This work is developed into the project Safety analysis long-term of high-level radioactive waste. This project is carried on for CIEMAT and ENRESA

  9. Uncertainty Evaluation of a Postulated LBLOCA for APR+ using KINS Realistic Evaluation Methodology and MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Min Jeong; Marigomena, Ralph; Yoo, Tae Ho; Kim, Y. S.; Sim, S. K. [Environment and Energy Technology, Inc., Daejeon (Korea, Republic of); Bang, Young Seok [KINS, Daejeon (Korea, Republic of)

    2014-05-15

    As a part of the regulatory safety research, Korea Institute of Nuclear Safety(KINS) also developed a best estimate safety analysis regulatory audit code, MARS-KS, to realistically predict and better understand the physical phenomena of the design basis accidents. KINS improved uncertainty propagation methodology using MARS-KS and applied the improved uncertainty evaluation method for the Shinkori Units 3 and 4 LBLOC. This study is to evaluate the uncertainty propagation of a postulated LBLOCA and quantify the safety margin using KINS-REM and MARS-KS code for the APR+ (Advanced Pressurizer Reactor Plus) Standard Safety Analysis Report(SSAR) which is under regulatory review by the KINS for its design approval. KINS-REM LBLOCA realistic evaluation methodology was used for the regulatory assessment of the APR+ LBLOCA using MARS-KS to evaluate the uncertainty propagation of the uncertainty variables as well as to assess the safety margin during the limiting case of the APR+ double ended guillotine cold leg LBLOCA. Uncertainty evaluation for the APR+ LBLOCA shows that the reflood PCT with upper limit of 95% probability at 95% confidence level is 1363.2 K and is higher than the blowdown PCT95/95 of 1275.3 K. The result shows that the current evaluation of APR+ LBLOCA PCT is within the acceptance criteria of 1477 K ECCS.

  10. 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)

  11. Creating and evaluating a new clicker methodology

    Science.gov (United States)

    Li, Pengfei

    "Clickers", an in-class polling system, has been used by many instructors to add active learning and formative assessment to previously passive traditional lectures. While considerable research has been conducted on clicker increasing student interaction in class, less research has been reported on the effectiveness of using clicker to help students understand concepts. This thesis reported a systemic project by the OSU Physics Education group to develop and test a new clicker methodology. Clickers question sequences based on a constructivist model of learning were used to improve classroom dynamics and student learning. They also helped students and lecturers understand in real time whether a concept had been assimilated or more effort was required. Chapter 1 provided an introduction to the clicker project. Chapter 2 summarized widely-accepted teaching principles that have arisen from a long history of research and practice in psychology, cognitive science and physics education. The OSU clicker methodology described in this thesis originated partly from our years of teaching experience, but mostly was based on these teaching principles. Chapter 3 provided an overview of the history of clicker technology and different types of clickers. Also, OSU's use of clickers was summarized together with a list of common problems and corresponding solutions. These technical details may be useful for those who want to use clickers. Chapter 4 discussed examples of the type and use of question sequences based on the new clicker methodology. In several years of research, we developed a base of clicker materials for calculus-based introductory physics courses at OSU. As discussed in chapter 5, a year-long controlled quantitative study was conducted to determine whether using clickers helps students learn, how using clickers helps students learn and whether students perceive that clicker has a positive effect on their own learning process. The strategy for this test was based on

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

  13. Drug safety evaluation of defibrotide.

    Science.gov (United States)

    Richardson, Paul G; Corbacioglu, Selim; Ho, Vincent Trien-Vinh; Kernan, Nancy A; Lehmann, Leslie; Maguire, Craig; Maglio, Michelle; Hoyle, Margaret; Sardella, Marco; Giralt, Sergio; Holler, Ernst; Carreras, Enric; Niederwieser, Dietger; Soiffer, Robert

    2013-01-01

    Hepatic veno-occlusive disease (VOD), also known as sinusoidal obstruction syndrome (SOS), is a potentially life-threatening complication of chemotherapeutic conditioning used in preparation for hematopoietic stem-cell transplantation (SCT). Defibrotide (DF) has been shown in Phase II and III trials to improve complete response in patients with severe VOD (sVOD). None of the articles, to date, provide a comprehensive review of the safety of DF in VOD and/or a range of other conditions. This article reviews current clinical findings on DF, primarily in terms of safety for use in treatment and prophylaxis of VOD, and relevant safety data for its use in other diseases. The literature review was conducted using a PubMed search with the fixed term 'defibrotide' in combination with ≥ 1 of 'safety', 'veno-occlusive disease' (with and without 'treatment', 'prevention'), 'oncology', 'myeloma', 'microangiopathy', 'anti-thrombotic' and 'peripheral vascular disorder'. Related articles from the EBMT and ASH conference websites were also included. DF was well tolerated in majority of the studies. The safety profile of DF is largely favourable with toxicities comparable to control populations in the setting of SCT complicated by sVOD.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-02-15

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

  15. Methodology to identify, review, and evaluate components for license renewal

    International Nuclear Information System (INIS)

    Carlson, D.D.; Gregor, F.E.; Walker, R.S.

    1988-01-01

    A methodology has been developed to systematically identify, review, and evaluate plant equipment for license renewal. The method builds upon the existing licensing basis, operating history, and accepted deterministic and probabilistic techniques. Use of these approaches provides a focus for license renewal upon those safety-significant systems and components that are not routinely replaced, refurbished, or subject to detailed inspection as part of the plant's existing test, maintenance, and surveillance programs. Application of the method identified the PWR and BWR systems that should be subjected to detailed license renewal review. Detailed examination of two example systems demonstrates the approach. The review and evaluation of plant equipment for license renewal differ from the initial licensing of the plant. A substantial operating history has been established, the licensing basis has evolved from the original one, and plant equipment has been subject to periodic maintenance and surveillance throughout its life. In consideration of these differences, a basis for license renewal is needed. License renewal should be based upon continuation of the existing licensing basis and recognition of existing programs and operating history

  16. 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.

  17. 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

  18. Methodological issues and pitfalls of short safety culture questionnaires

    NARCIS (Netherlands)

    Jagtman, H.M.; Koornneef, F.; Akselsson, R.; Stewart, S.

    2013-01-01

    Safety culture surveys have been fielded in many different sectors of industrial activities. Many of these surveys consist of a long list of questions which is time consuming for the respondents. As part of the FP6 HILAS project a shorter survey has been developed, which aimed at getting a high

  19. The dynamic flowgraph methodology as a safety analysis tool : programmable electronic system design and verification

    NARCIS (Netherlands)

    Houtermans, M.J.M.; Apostolakis, G.E.; Brombacher, A.C.; Karydas, D.M.

    2002-01-01

    The objective of this paper is to demonstrate the use of the Dynamic Flowgraph Methodology (DFM) during the design and verification of programmable electronic safety-related systems. The safety system consists of hardware as well as software. This paper explains and demonstrates the use of DFM, and

  20. Approach to uncertainty evaluation for safety analysis

    International Nuclear Information System (INIS)

    Ogura, Katsunori

    2005-01-01

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

  1. Safeguards Automated Facility Evaluation (SAFE) methodology

    International Nuclear Information System (INIS)

    Chapman, L.D.; Grady, L.M.; Bennett, H.A.; Sasser, D.W.; Engi, D.

    1978-08-01

    An automated approach to facility safeguards effectiveness evaluation has been developed. This automated process, called Safeguards Automated Facility Evaluation (SAFE), consists of a collection of a continuous stream of operational modules for facility characterization, the selection of critical paths, and the evaluation of safeguards effectiveness along these paths. The technique has been implemented on an interactive computer time-sharing system and makes use of computer graphics for the processing and presentation of information. Using this technique, a comprehensive evaluation of a safeguards system can be provided by systematically varying the parameters that characterize the physical protection components of a facility to reflect the perceived adversary attributes and strategy, environmental conditions, and site operational conditions. The SAFE procedure has broad applications in the nuclear facility safeguards field as well as in the security field in general. Any fixed facility containing valuable materials or components to be protected from theft or sabotage could be analyzed using this same automated evaluation technique

  2. Safety evaluation of a hydrogen fueled transit bus

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-08-01

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

  7. Safeguards Automated Facility Evaluation (SAFE) methodology

    International Nuclear Information System (INIS)

    Chapman, L.D.; Grady, L.M.; Bennett, H.A.; Sasser, D.W.; Engi, D.

    1978-01-01

    The SAFE procedure is an efficient method of evaluating the physical protection system of a nuclear facility. Since the algorithms used in SAFE for path generation and evaluation are analytical, many paths can be evaluated with a modest investment in computer time. SAFE is easy to use because the information required is well-defined and the interactive nature of this procedure lends itself to straightforward operation. The modular approach that has been taken allows other functionally equivalent modules to be substituted as they become available. The SAFE procedure has broad applications in the nuclear facility safeguards field as well as in the security field in general. Any fixed facility containing valuable materials or components to be protected from theft or sabotage could be analyzed using this same automated evaluation technique

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. 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.)

  11. Site evaluation for nuclear installations. Safety requirements

    International Nuclear Information System (INIS)

    2003-01-01

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

  12. A framework of risk-informed seismic safety evaluation of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Kondo, S.; Sakagami, M.; Hirano, M.; Shiba, M.

    2001-01-01

    A framework of risk-informed seismic design and safety evaluation of nuclear power plants is under consideration in Japan so as to utilize the progress in the seismic probabilistic safety assessment methodology. Issues resolved to introduce this framework are discussed after the concept, evaluation process and characteristics of the framework are described. (author)

  13. 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

  14. Fault-tolerant architecture: Evaluation methodology

    International Nuclear Information System (INIS)

    Battle, R.E.; Kisner, R.A.

    1992-08-01

    The design and reliability of four fault-tolerant architectures that may be used in nuclear power plant control systems were evaluated. Two architectures are variations of triple-modular-redundant (TMR) systems, and two are variations of dual redundant systems. The evaluation includes a review of methods of implementing fault-tolerant control, the importance of automatic recovery from failures, methods of self-testing diagnostics, block diagrams of typical fault-tolerant controllers, review of fault-tolerant controllers operating in nuclear power plants, and fault tree reliability analyses of fault-tolerant systems

  15. 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...

  16. Present stage evaluation of Furnas calculus methodology qualification

    International Nuclear Information System (INIS)

    1987-07-01

    This technical note is about the present stage evaluation of FURNAS Calculus Methodology Qualification related to reload licensing process and licensing support of operation questions of Angra 1 NPP concerning transient and Core ThermalHydraulic areas. (Author) [pt

  17. Methodology for Evaluating Encapsulated Beneficial Uses of Coal Combustion Residuals

    Science.gov (United States)

    The primary purpose of this document is to present an evaluation methodology developed by the EPA for making determinations about environmental releases from encapsulated products containing coal combustion residuals.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. THE INTRODUCTION OF THE METHODOLOGY TO IMPROVE ROAD SAFETY

    Directory of Open Access Journals (Sweden)

    D. V. Kapsky

    2013-01-01

    Full Text Available Recommendations for improving the road safety and quality of road traffic controlled junctions(crossings on individual parameters of traffic light control, improvement of traffic light control by optimizing the length of the transition interval in the traffic light cycle, increase awareness and early warning drivers about the upcoming change traffic lights division of transport and pedestrian traffic, road conditions , transportation planning and technical aids of road  traffic, as well as recommendations for the use of the hump in the settlements, etc.

  20. Interviewing: Methodological Briefs - Impact Evaluation No. 12

    OpenAIRE

    Bronwen McDonald; Patricia Rogers

    2014-01-01

    Interviews are easy to do badly and hard to do well - good planning, adequate time and appropriate skills are required. The type of interview should be carefully chosen to suit the situation rather than choosing a type of interview (such as focus groups) simply because it is commonly used. Interviews with children raise particular ethical issues that need to be carefully considered and fully addressed. This brief outlines key issues to consider in planning interviews for impact evaluation, ta...

  1. Experiment on safety software evaluation

    International Nuclear Information System (INIS)

    Soubies, B.; Henry, J.Y.

    1994-06-01

    The licensing procedures process of nuclear plants includes compulsory steps which bring about a thorough exam of the commands control system. In this context the IPSN uses a tool called MALPAS to carry out an analysis of the quality of the software involved in safety control. The IPSN also try to obtain the automation of the generation of test games necessary for dynamical analysis. The MALPAS tool puts forward the particularities of programing which can influence the testability and the upholding of the studied software. (TEC). 4 refs

  2. An overview of IPSN research on the evolution of the natural systems in support of the French methodology for the safety evaluation of radwaste disposal in deep geological formations

    International Nuclear Information System (INIS)

    Escalier des Orres, P.; Granier, T.; Mohammadioun, B.

    1992-01-01

    A regulatory guidance has been recently set up in France for the safety assessment of radwaste deep geological disposal: the present paper concerns the requirements related to bedrock stability issues and their technical background. This regulation relies in particular on a long term effort of the Protection and Nuclear Safety Institute (IPSN) of the French Atomic Energy Commission (CEA), which ensures two main duties: it carries out research programs in the area of protection and nuclear safety and provides expertise to the safety authorities. It should be noted that ANDRA (the French National Radioactive Waste Management Agency) is responsible for the safety of radioactive waste management and relies, for safety demonstration purposes, on its research programs. IPSN, in support of the safety authorities, is in charge of the verification of the applicant's safety demonstration and develops its own research programs in order to achieve an independent capability in safety analysis expertise. We present here the major axes of the Institute research program devoted to the assessment of seismic events consequences on the groundwater system. 19 refs., 8 figs

  3. An overview of IPSN research on the evolution of the natural systems in support of the French methodology for the safety evaluation of radwaste disposal in deep geological formations

    International Nuclear Information System (INIS)

    Escalier des Orres, P.; Granier, T.; Mohammadioun, B.

    1992-01-01

    A regulatory guidance has been recently set up in France for the safety assessment of radwaste deep in geological disposal: the present paper concerns the requirements related to bedrock stability issues and their technical background. This regulation relies in particular on a long term effort of the Protection and Nuclear Safety Institute (IPSN) of the French Atomic Energy Commission (CEA), which ensures two main duties: it carries out research programs in the area of protection and nuclear safety and provides expertise to the safety authorities. It should be noted that ANDRA (the French National Radioactive Waste Management Agency) is responsible for the safety of radioactive waste management and relies, for safety demonstration purposes, on its research programs. IPSN, in support of the safety authorities, is in charge of the verification of the applicant's safety demonstration and develops its own research programs in order to achieve an independent capability in safety analysis expertise. We present here the major axes of the Institute research program devoted to the assessment of seismic events consequences on the groundwater system. 19 refs., 8 figs

  4. 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.

  5. Methodology for safety optimization of highway cross-sections for horizontal curves with restricted sight distance.

    Science.gov (United States)

    Ibrahim, Shewkar E; Sayed, Tarek; Ismail, Karim

    2012-11-01

    Several earlier studies have noted the shortcomings with existing geometric design guides which provide deterministic standards. In these standards the safety margin of the design output is generally unknown and there is little knowledge of the safety implications of deviating from the standards. To mitigate these shortcomings, probabilistic geometric design has been advocated where reliability analysis can be used to account for the uncertainty in the design parameters and to provide a mechanism for risk measurement to evaluate the safety impact of deviations from design standards. This paper applies reliability analysis for optimizing the safety of highway cross-sections. The paper presents an original methodology to select a suitable combination of cross-section elements with restricted sight distance to result in reduced collisions and consistent risk levels. The purpose of this optimization method is to provide designers with a proactive approach to the design of cross-section elements in order to (i) minimize the risk associated with restricted sight distance, (ii) balance the risk across the two carriageways of the highway, and (iii) reduce the expected collision frequency. A case study involving nine cross-sections that are parts of two major highway developments in British Columbia, Canada, was presented. The results showed that an additional reduction in collisions can be realized by incorporating the reliability component, P(nc) (denoting the probability of non-compliance), in the optimization process. The proposed approach results in reduced and consistent risk levels for both travel directions in addition to further collision reductions. Copyright © 2012 Elsevier Ltd. All rights reserved.

  6. 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:

  7. 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:

  8. 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

  9. Researches on nuclear criticality safety evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-10-01

    For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

  10. Researches on nuclear criticality safety evaluation

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi

    2003-01-01

    For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

  11. Status of the Gen-IV Proliferation Resistance and Physical Protection (PRPP) Evaluation Methodology

    International Nuclear Information System (INIS)

    Whitlock, J.; Bari, R.; Peterson, P.; Padoani, F.; Cojazzi, G.G.M.; Renda, G.; ); Cazalet, J.; Haas, E.; Hori, K.; Kawakubo, Y.; Chang, S.; Kim, H.; Kwon, E.-H.; Yoo, H.; Chebeskov, A.; Pshakin, G.; Pilat, J.F.; Therios, I.; Bertel, E.

    2015-01-01

    Methodologies have been developed within the Generation IV International Forum (GIF) to support the assessment and improvement of system performance in the areas safeguards, security, economics and safety. Of these four areas, safeguards and security are the subjects of the GIF working group on Proliferation Resistance and Physical Protection (PRPP). Since the PRPP methodology (now at Revision 6) represents a mature, generic, and comprehensive evaluation approach, and is freely available on the GIF public website, several non-GIF technical groups have chosen to utilize the PRPP methodology for their own goals. Indeed, the results of the evaluations performed with the methodology are intended for three types of generic users: system designers, programme policy makers, and external stakeholders. The PRPP Working Group developed the methodology through a series of demonstration and case studies. In addition, over the past few years various national and international groups have applied the methodology to inform nuclear energy system designs, as well as to support the development of approaches to advanced safeguards. A number of international workshops have also been held which have introduced the methodology to design groups and other stakeholders. In this paper we summarize the technical progress and accomplishments of the PRPP evaluation methodology, including applications outside GIF, and we outline the PRPP methodology's relationship with the IAEA's INPRO methodology. Current challenges with the efficient implementation of the methodology are outlined, along with our path forward for increasing its accessibility to a broader stakeholder audience - including supporting the next generation of skilled professionals in the nuclear non-proliferation field. (author)

  12. Prospective safety performance evaluation on construction sites.

    Science.gov (United States)

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

    2015-05-01

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

  13. Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

    Science.gov (United States)

    Mkhabela, Peter Tshepo

    The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis

  14. Plutonium Finishing Plant safety evaluation report

    International Nuclear Information System (INIS)

    1995-01-01

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

  15. Empirical evaluation and justification of methodologies in psychological science.

    Science.gov (United States)

    Proctor, R W; Capaldi, E J

    2001-11-01

    The purpose of this article is to describe a relatively new movement in the history and philosophy of science, naturalism, a form of pragmatism emphasizing that methodological principles are empirical statements. Thus, methodological principles must be evaluated and justified on the same basis as other empirical statements. On this view, methodological statements may be less secure than the specific scientific theories to which they give rise. The authors examined the feasibility of a naturalistic approach to methodology using logical and historical analysis and by contrasting theories that predict new facts versus theories that explain already known facts. They provide examples of how differences over methodological issues in psychology and in science generally may be resolved using a naturalistic, or empirical, approach.

  16. Probabilistic safety analysis for FRJ-2 motivation, methodology and results

    International Nuclear Information System (INIS)

    Wolters, J.

    1993-01-01

    A PSA of the Research Reactor FRJ-2 was performed to check the twenty-year-old safety system for weak points and to develop accident management as a 'fourth line of defence' against severe accidents according to a German initiative. The total core damage frequency proved to be 1.5·10 -4 /a, which is consistent with figures found for other research reactors. Minor plant modifications will reduce the value by roughly a factor of 4, resulting in a frequency of 3·10 -7 /a for a major release of fission products into the environment caused by an independent failure of the containment. The integrity of the gas-tight steel containment proved not to be endangered by any core damage accident. From the results and insights gained by the PSA, many accident management measures could be identified and defined for the emergency handbook. The most important measure is primary feed and bleed, for which the feed line already exists. (author)

  17. Probabilistic safety analysis for FRJ-2 motivation, methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Wolters, J [Institute for Safety Research and Reactor Technology, Research Center Juelich (Germany)

    1993-07-01

    A PSA of the Research Reactor FRJ-2 was performed to check the twenty-year-old safety system for weak points and to develop accident management as a 'fourth line of defence' against severe accidents according to a German initiative. The total core damage frequency proved to be 1.5{center_dot}10{sup -4}/a, which is consistent with figures found for other research reactors. Minor plant modifications will reduce the value by roughly a factor of 4, resulting in a frequency of 3{center_dot}10{sup -7}/a for a major release of fission products into the environment caused by an independent failure of the containment. The integrity of the gas-tight steel containment proved not to be endangered by any core damage accident. From the results and insights gained by the PSA, many accident management measures could be identified and defined for the emergency handbook. The most important measure is primary feed and bleed, for which the feed line already exists. (author)

  18. 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

  19. Probabilistic safety analysis for FRJ-2 motivation, methodology and results

    International Nuclear Information System (INIS)

    Wolters, J.

    1994-01-01

    A PSA of the Research Reactor FRJ-2 was performed to check the twenty-year-old safety system for weak points and to develop accident management as a 'fourth line of defence' against severe accidents according to a German initiative. The total core damage frequency proved to be 1.5·10 -4 /a, which is consistent with figures found for other research reactors. Minor plant modifications will reduce the value by roughly a factor of 4, resulting in a frequency of 3·10 -7 /a for a major release of fission products into the environment caused by an independent failure of the containment. The integrity of the gas-tight steel containment proved not to be endangered by any core damage accident. From the results and insights gained by the PSA, many accident management measures could be identified and defined for the emergency handbook. The most important measure is primary feed and bleed, for which the feed line already exists. (author)

  20. Physical protection evaluation methodology program development and application

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Janghoon; Yoo, Hosik [Korea Institute of Nuclear Non-proliferation and Control, Daejeon (Korea, Republic of)

    2015-10-15

    It is essential to develop a reliable physical protection evaluation methodology for applying physical protection concept to the design stage. The methodology can be used to assess weak points and improve performance not only for the design stage but also for nuclear facilities in operation. Analyzing physical protection property of nuclear facilities is not a trivial work since there are many interconnected factors affecting overall performance. Therefore several international projects have been organized to develop a systematic physical protection evaluation methodology. INPRO (The International Project on Innovative Nuclear Reactors and Fuel Cycles) and GIF PRPP (Generation IV International Forum Proliferation Resistance and Physical Protection) methodology are among the most well-known evaluation methodologies. INPRO adopts a checklist type of questionnaire and has a strong point in analyzing overall characteristic of facilities in a qualitative way. COMPRE program has been developed to help general users apply COMPRE methodology to nuclear facilities. In this work, COMPRE program development and a case study of the hypothetical nuclear facility are presented. The development of COMPRE program and a case study for hypothetic facility is presented in this work. The case study shows that COMPRE PP methodology can be a useful tool to assess the overall physical protection performance of nuclear facilities. To obtain meaningful results from COMPRE PP methodology, detailed information and comprehensive analysis are required. Especially, it is not trivial to calculate reliable values for PPSE (Physical Protection System Effectiveness) and C (Consequence), while it is relatively straightforward to evaluate LI (Legislative and Institutional framework), MC (Material Control) and HR (Human Resources). To obtain a reliable PPSE value, comprehensive information about physical protection system, vital area analysis and realistic threat scenario assessment are required. Like

  1. Physical protection evaluation methodology program development and application

    International Nuclear Information System (INIS)

    Seo, Janghoon; Yoo, Hosik

    2015-01-01

    It is essential to develop a reliable physical protection evaluation methodology for applying physical protection concept to the design stage. The methodology can be used to assess weak points and improve performance not only for the design stage but also for nuclear facilities in operation. Analyzing physical protection property of nuclear facilities is not a trivial work since there are many interconnected factors affecting overall performance. Therefore several international projects have been organized to develop a systematic physical protection evaluation methodology. INPRO (The International Project on Innovative Nuclear Reactors and Fuel Cycles) and GIF PRPP (Generation IV International Forum Proliferation Resistance and Physical Protection) methodology are among the most well-known evaluation methodologies. INPRO adopts a checklist type of questionnaire and has a strong point in analyzing overall characteristic of facilities in a qualitative way. COMPRE program has been developed to help general users apply COMPRE methodology to nuclear facilities. In this work, COMPRE program development and a case study of the hypothetical nuclear facility are presented. The development of COMPRE program and a case study for hypothetic facility is presented in this work. The case study shows that COMPRE PP methodology can be a useful tool to assess the overall physical protection performance of nuclear facilities. To obtain meaningful results from COMPRE PP methodology, detailed information and comprehensive analysis are required. Especially, it is not trivial to calculate reliable values for PPSE (Physical Protection System Effectiveness) and C (Consequence), while it is relatively straightforward to evaluate LI (Legislative and Institutional framework), MC (Material Control) and HR (Human Resources). To obtain a reliable PPSE value, comprehensive information about physical protection system, vital area analysis and realistic threat scenario assessment are required. Like

  2. Economic evaluation in stratified medicine: methodological issues and challenges

    Directory of Open Access Journals (Sweden)

    Hans-Joerg eFugel

    2016-05-01

    Full Text Available Background: Stratified Medicine (SM is becoming a practical reality with the targeting of medicines by using a biomarker or genetic-based diagnostic to identify the eligible patient sub-population. Like any healthcare intervention, SM interventions have costs and consequences that must be considered by reimbursement authorities with limited resources. Methodological standards and guidelines exist for economic evaluations in clinical pharmacology and are an important component for health technology assessments (HTAs in many countries. However, these guidelines have initially been developed for traditional pharmaceuticals and not for complex interventions with multiple components. This raises the issue as to whether these guidelines are adequate to SM interventions or whether new specific guidance and methodology is needed to avoid inconsistencies and contradictory findings when assessing economic value in SM.Objective: This article describes specific methodological challenges when conducting health economic (HE evaluations for SM interventions and outlines potential modifications necessary to existing evaluation guidelines /principles that would promote consistent economic evaluations for SM.Results/Conclusions: Specific methodological aspects for SM comprise considerations on the choice of comparator, measuring effectiveness and outcomes, appropriate modelling structure and the scope of sensitivity analyses. Although current HE methodology can be applied for SM, greater complexity requires further methodology development and modifications in the guidelines.

  3. Safety assessment of a vault-based disposal facility using the ISAM methodology

    International Nuclear Information System (INIS)

    Kelly, E.; Kim, C.-L.; Lietava, P.; Little, R.; Simon, I.

    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 testing the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Vault Test Case (VTC), related to the disposal of low level radioactive waste (LLW) to a hypothetical facility comprising a set of above surface vaults. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the VTC and provides some general conclusions that can be drawn from the findings of their work. (author)

  4. Optimized Evaluation System to Athletic Food Safety

    OpenAIRE

    Shanshan Li

    2015-01-01

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

  5. SGHWR safety design and evaluation

    International Nuclear Information System (INIS)

    Smith, D.R.; Merrett, D.J.; Ward, D.A.

    1977-01-01

    The paper discusses the characteristic features of the S.G.H.W.R. and identifies the single channel concept as of considerable importance. The unique feature of the design is the provision of individual spray cooling E.C.C.S. to each channel. This spray cooling occupies a prominent position in the main line safety arguments. The reliance on this form of spray cooling leads to provision of a comprehensive E.C.C.S. system of high reliability. Duplicate systems with diverse power and water sources cover the complete pressure range to give very high confidence that spray cooling is available in all major L.O.C.A.s. On the other hand hydraulic analysis of the blowdown phase demonstrates that significant convective flow is available as an alternative/supplementary cooling regime for most faults. The reactor shutdown mechanisms have also been duplicated and will be designed to high reliabilities to give surety of reactor trip in all credible faults. The comparative performance of the two systems is considered. Extent of diversity and redundancy in trip parameters is also discussed. A feature of channel concept is that the pipe sizes can be made relatively small thus restricting rates of blowdown, and the paper discusses effects of this upon long term cooling and flooding arguments. The quantities of pipework in the primary circuit introduce considerations of integrity and the paper goes on to list the measures introduced to improve segregation and protection of individual sections of the plant so that the extent of possible L.O.C.A.s is minimised. The achievement of high standards of reliability by use of in-service inspection is covered, with particular reference to the steam drums. The impact of these inspection requirements upon the very low man-rem exposures required by U.K. utilities is also included. Finally, it is noted that the provision of containment in common with other L.W.R. practice also provides a valuable engineered safety feature. The principles of

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

    International Nuclear Information System (INIS)

    Wahlstroem, B.; Gunsell, L.

    1998-03-01

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

  7. Safety evaluation of large ventilation networks

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  8. Squale: evaluation criteria of functioning safety

    International Nuclear Information System (INIS)

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

    1998-05-01

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

  9. Evaluation of safeguards procedures: a summary of a methodology

    International Nuclear Information System (INIS)

    Salisbury, J.D.; Savage, J.W.

    1979-01-01

    A methodology for the evaluation of safeguards procedures is described. As presently conceptualized, the methodology will consist of the following steps: (1) expansion of the general protection requirements that are contained in the NRC regulations into more detailed but still generic requirements for use at the working level; (2) development of techniques and formats for using the working-level requirements in an evaluation; (3) development of a technique for converting specific facility protection procedures into a format that will allow comparison with the working-level requirements; (4) development of an evaluation technique for comparing the facility protection procedures to determine if they meet the protection requirements

  10. 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

  11. Safety culture management and quantitative indicator evaluation

    International Nuclear Information System (INIS)

    Mandula, J.

    2002-01-01

    This report discuses a relationship between safety culture and evaluation of quantitative indicators. It shows how a systematic use of generally shared operational safety indicators may contribute to formation and reinforcement of safety culture characteristics in routine plant operation. The report also briefly describes the system of operational safety indicators used at the Dukovany plant. It is a PC database application enabling an effective work with the indicators and providing all users with an efficient tool for making synoptic overviews of indicator values in their links and hierarchical structure. Using color coding, the system allows quick indicator evaluation against predefined limits considering indicator value trends. The system, which has resulted from several-year development, was completely established at the plant during the years 2001 and 2002. (author)

  12. Safety evaluation of synthetic β-carotene

    NARCIS (Netherlands)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Kirillovich, A.P.

    2000-01-01

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

  14. 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.

  15. Green light for the methodology file. Periodical safety review 2016 has begun

    International Nuclear Information System (INIS)

    2014-01-01

    Every ten years, the organisational framework and the facilities of the Belgian Nuclear Research Center SCK-CEN are subject to an encompassing safety evaluation. Together with initiatives arising from the stress tests, a large number of safety studies and actions are scheduled until 2018. The article discusses the ongoing safety review.

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

    Science.gov (United States)

    Lores, Marta; 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. Copyright © 2016

  17. 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.

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

    OpenAIRE

    Bennett, Anthea; Hellier, Elizabeth; Weyman, Andrew

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Hedin, A.

    2008-01-01

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

  20. An investigative study towards constructing anthropocentric Man-Machine System design evaluation methodology

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Gofuku, A.; Itoh, T.; Sasaki, K.

    1992-01-01

    A methodological investigation has been conducted for evaluating the reliability of man-machine interaction in the total Man-Machine System (MMS) from the view-point of safety maintenance for emergent situations of nuclear power plant. Basic considerations in our study are: (i) what are the MMS design data to be evaluated, (ii) how are those MMS design data should be treated, and (iii) how the introduction effects of various operator support tools can be evaluated. The methods of both qualitative and quantitative MMS design evaluation are summarized in this paper, with the system architecture based on man-machine interaction simulation and the related cognitive human error factor analysis. (author)

  1. 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)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  3. Development of evaluation method for software safety analysis techniques

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  4. Methodology of investment effectiveness evaluation in the local energy market

    Energy Technology Data Exchange (ETDEWEB)

    Kamrat, W.

    1999-07-01

    The paper presents issues of investment effectiveness evaluation in the local energy market. Results of research presented in the paper are mainly proposing a concept of a methodology which allows the evaluation of investment processes in regional power markets at the decision-making stage. In this respect, selecting a rational investment strategy is an important stage of the entire investment process. In view of criteria of various nature, the construction of a methodology of investment effectiveness bears an especially important meaning for a local decision-maker or investor. It is of particular significance to countries that are undergoing a transition from a centrally planned economy to a market economy. (orig.)

  5. 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

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

    NARCIS (Netherlands)

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

    2017-01-01

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

  7. 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.

  8. 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.

  9. Methodology for safety assessment of near-surface radioactive waste disposal facilities

    International Nuclear Information System (INIS)

    Mateeva, M.

    1998-01-01

    The objective of the work is to present the conceptual model of the methodology of safety assessment of near-surface radioactive disposal facilities. The widely used mathematical models and approaches are presented. The emphasis is given on the mathematical models and approaches, which are applicable for the conditions in our country. The different transport models for analysis and safety assessment of migration processes are presented. The parallel between the Mixing-Cell Cascade model and model of Finite-Differences is made. In the methodology the basic physical and chemical processes and events, concerning mathematical modelling of the flow and the transport of radionuclides from the Near Field to Far Field and Biosphere are analyzed. Suitable computer codes corresponding to the ideology and appropriate for implementing of the methodology are shown

  10. Problems of nuclear power plant safety evaluation

    International Nuclear Information System (INIS)

    Suchomel, J.

    1977-01-01

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

  11. Evaluation Methodology. The Evaluation Exchange. Volume 11, Number 2, Summer 2005

    Science.gov (United States)

    Coffman, Julia, Ed.

    2005-01-01

    This is the third issue of "The Evaluation Exchange" devoted entirely to the theme of methodology, though every issue tries to identify new methodological choices, the instructive ways in which people have applied or combined different methods, and emerging methodological trends. For example, lately "theories of change" have gained almost…

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

    International Nuclear Information System (INIS)

    Aldemir, Tunc

    2013-01-01

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

  13. A Theoretical and Methodological Evaluation of Leadership Research.

    Science.gov (United States)

    Lashbrook, Velma J.; Lashbrook, William B.

    This paper isolates some of the strengths and weaknesses of leadership research by evaluating it from both a theoretical and methodological perspective. The seven theories or approaches examined are: great man, trait, situational, style, functional, social influence, and interaction positions. General theoretical, conceptual, and measurement…

  14. Guidelines for reporting evaluations based on observational methodology.

    Science.gov (United States)

    Portell, Mariona; Anguera, M Teresa; Chacón-Moscoso, Salvador; Sanduvete-Chaves, Susana

    2015-01-01

    Observational methodology is one of the most suitable research designs for evaluating fidelity of implementation, especially in complex interventions. However, the conduct and reporting of observational studies is hampered by the absence of specific guidelines, such as those that exist for other evaluation designs. This lack of specific guidance poses a threat to the quality and transparency of these studies and also constitutes a considerable publication hurdle. The aim of this study thus was to draw up a set of proposed guidelines for reporting evaluations based on observational methodology. The guidelines were developed by triangulating three sources of information: observational studies performed in different fields by experts in observational methodology, reporting guidelines for general studies and studies with similar designs to observational studies, and proposals from experts in observational methodology at scientific meetings. We produced a list of guidelines grouped into three domains: intervention and expected outcomes, methods, and results. The result is a useful, carefully crafted set of simple guidelines for conducting and reporting observational studies in the field of program evaluation.

  15. A multiple feedback methodology for the study of creativity evaluations

    DEFF Research Database (Denmark)

    Glaveanu, Vlad Petre

    2012-01-01

    The cultural psychology perspective emphasizes the role of communication and social interaction in the construction of representations and evaluations concerning creativity. In terms of creativity assessment, this approach leads to the use of a multiple feedback methodology selecting several groups...... useful for investigating the social and cultural dynamics of creativity....

  16. Evaluation and testing methodology for evolving entertainment systems

    NARCIS (Netherlands)

    Jurgelionis, A.; Bellotti, F.; IJsselsteijn, W.A.; Kort, de Y.A.W.; Bernhaupt, R.; Tscheligi, M.

    2007-01-01

    This paper presents a testing and evaluation methodology for evolving pervasive gaming and multimedia systems. We introduce the Games@Large system, a complex gaming and multimedia architecture comprised of a multitude of elements: heterogeneous end user devices, wireless and wired network

  17. Operational safety evaluation for minor reactor accidents

    International Nuclear Information System (INIS)

    Wang, O.S.

    1981-01-01

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

  18. Evaluation of methodologies for remunerating wind power's reliability in Colombia

    International Nuclear Information System (INIS)

    Botero B, Sergio; Isaza C, Felipe; Valencia, Adriana

    2010-01-01

    Colombia strives to have enough firm capacity available to meet unexpected power shortages and peak demand; this is clear from mechanisms currently in place that provide monetary incentives (in the order of nearly US$ 14/MW h) to power producers that can guarantee electricity provision during scarcity periods. Yet, wind power in Colombia is not able to currently guarantee firm power because an accepted methodology to calculate its potential firm capacity does not exist. In this paper we argue that developing such methodology would provide an incentive to potential investors to enter into this low carbon technology. This paper analyzes three methodologies currently used in energy markets around the world to calculate firm wind energy capacity: PJM, NYISO, and Spain. These methodologies are initially selected due to their ability to accommodate to the Colombian energy regulations. The objective of this work is to determine which of these methodologies makes most sense from an investor's perspective, to ultimately shed light into developing a methodology to be used in Colombia. To this end, the authors developed a methodology consisting on the elaboration of a wind model using the Monte-Carlo simulation, based on known wind behaviour statistics of a region with adequate wind potential in Colombia. The simulation gives back random generation data, representing the resource's inherent variability and simulating the historical data required to evaluate the mentioned methodologies, thus achieving the technology's theoretical generation data. The document concludes that the evaluated methodologies are easy to implement and that these do not require historical data (important for Colombia, where there is almost no historical wind power data). It is also found that the Spanish methodology provides a higher Capacity Value (and therefore a higher return to investors). The financial assessment results show that it is crucial that these types of incentives exist to make viable

  19. Development of Testing Methodologies to Evaluate Postflight Locomotor Performance

    Science.gov (United States)

    Mulavara, A. P.; Peters, B. T.; Cohen, H. S.; Richards, J. T.; Miller, C. A.; Brady, R.; Warren, L. E.; Bloomberg, J. J.

    2006-01-01

    Crewmembers experience locomotor and postural instabilities during ambulation on Earth following their return from space flight. Gait training programs designed to facilitate recovery of locomotor function following a transition to a gravitational environment need to be accompanied by relevant assessment methodologies to evaluate their efficacy. The goal of this paper is to demonstrate the operational validity of two tests of locomotor function that were used to evaluate performance after long duration space flight missions on the International Space Station (ISS).

  20. 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

  1. Performance evaluation methodology for historical document image binarization.

    Science.gov (United States)

    Ntirogiannis, Konstantinos; Gatos, Basilis; Pratikakis, Ioannis

    2013-02-01

    Document image binarization is of great importance in the document image analysis and recognition pipeline since it affects further stages of the recognition process. The evaluation of a binarization method aids in studying its algorithmic behavior, as well as verifying its effectiveness, by providing qualitative and quantitative indication of its performance. This paper addresses a pixel-based binarization evaluation methodology for historical handwritten/machine-printed document images. In the proposed evaluation scheme, the recall and precision evaluation measures are properly modified using a weighting scheme that diminishes any potential evaluation bias. Additional performance metrics of the proposed evaluation scheme consist of the percentage rates of broken and missed text, false alarms, background noise, character enlargement, and merging. Several experiments conducted in comparison with other pixel-based evaluation measures demonstrate the validity of the proposed evaluation scheme.

  2. Evaluating fuel cycle safety for CITa

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  3. Evaluation of criteria for developing traffic safety materials for Latinos.

    Science.gov (United States)

    Streit-Kaplan, Erica L; Miara, Christine; Formica, Scott W; Gallagher, Susan Scavo

    2011-03-01

    This quantitative study assessed the validity of guidelines that identified four key characteristics of culturally appropriate Spanish-language traffic safety materials: language, translation, formative evaluation, and credible source material. From a sample of 190, the authors randomly selected 12 Spanish-language educational materials for analysis by 15 experts. Hypotheses included that the experts would rate materials with more of the key characteristics as more effective (likely to affect behavioral change) and rate materials originally developed in Spanish and those that utilized formative evaluation (e.g., pilot tests, focus groups) as more culturally appropriate. Although results revealed a weak association between the number of key characteristics in a material and the rating of its effectiveness, reviewers rated materials originally created in Spanish and those utilizing formative evaluation as significantly more culturally appropriate. The findings and methodology demonstrated important implications for developers and evaluators of any health-related materials for Spanish speakers and other population groups.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  6. 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)

  7. 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)

  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. Safety insights from forensics evaluations at Daiichi

    Directory of Open Access Journals (Sweden)

    J. Rempe

    2017-01-01

    Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company (TEPCO to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This paper reports initial results from the US Forensics Effort to utilize examination information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. In this paper, three examples are presented in which examination information, such as visual images, dose surveys, sample evaluations, and muon tomography examinations, along with data from plant instrumentation, are used to obtain significant safety insights in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights confirm actions, such as the importance of water addition and containment venting, that are emphasized in updated guidance for severe accident prevention, mitigation, and emergency planning.

  10. Evaluation methodology for fixed-site physical protection systems

    International Nuclear Information System (INIS)

    Bennett, H.A.; Olascoaga, M.T.

    1980-01-01

    A system performance evaluation methodology has been developed to aid the Nuclear Regulatory Commission (NRC) in the implementation of new regulations designed to upgrade the physical protection of nuclear fuel cycle facilities. The evaluation methodology, called Safeguards Upgrade Rule Evaluation (SURE), provides a means of explicitly incorporating measures for highly important and often difficult to quantify performance factors, e.g., installation, maintenance, training and proficiency levels, compatibility of components in subsystems, etc. This is achieved by aggregating responses to component and system questionaires through successive levels of a functional hierarchy developed for each primary performance capability specified in the regulations, 10 CFR 73.45. An overall measure of performance for each capability is the result of this aggregation process. This paper provides a descripton of SURE

  11. Methodology For Evaluation Of Regulatory Effectiveness In Physical Protection

    International Nuclear Information System (INIS)

    Izmaylov, Alexander; Valente, John; Griggs, James R.; Rexroth, Paul; Piskarev, Alexander; Babkin, Vladimir; Sokolov, Egor; Melton, Ronald B.; Cunningham, Mitchel E.; Baker, Kathryn A.; Brothers, Alan J.

    2005-01-01

    Material protection, control, and accounting (MPC and A) regulatory documents play an important role in securing and protecting nuclear material by regulating a variety of activities at different hierarchical levels. The development, implementation, and practical application of these regulatory documents requires a significant investment of financial and material resources. Therefore, it is important to evaluate the effectiveness of the regulatory development process and the extent to which regulations improve the effectiveness of MPC and A at nuclear sites. The joint Russian and U.S. Regulatory Development Project has a goal of evaluating the effectiveness of regulatory documents developed for MPC and A. As part of this joint Project, a methodology for evaluating effectiveness has been developed. This methodology was developed around physical protection objectives. The developed methodology specifies physical protection objectives to be accomplished through the implementation of a regulatory system based on the physical protection goals at the nuclear sites. It includes approaches to assessing regulatory effectiveness, the hierarchical structure of physical protection objectives to be accomplished through implementing regulations, a 'mapping' of the physical protection objectives to the regulatory framework, a list of criteria for evaluating the effectiveness of physical protection regulations and effectiveness indicators, as well as means and methods for gathering information and implementation of this evaluation.

  12. 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.

  13. 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

  14. The International Criticality Safety Benchmark Evaluation Project

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Software Dependability and Safety Evaluations ESA's Initiative

    Science.gov (United States)

    Hernek, M.

    ESA has allocated funds for an initiative to evaluate Dependability and Safety methods of Software. The objectives of this initiative are; · More extensive validation of Safety and Dependability techniques for Software · Provide valuable results to improve the quality of the Software thus promoting the application of Dependability and Safety methods and techniques. ESA space systems are being developed according to defined PA requirement specifications. These requirements may be implemented through various design concepts, e.g. redundancy, diversity etc. varying from project to project. Analysis methods (FMECA. FTA, HA, etc) are frequently used during requirements analysis and design activities to assure the correct implementation of system PA requirements. The criticality level of failures, functions and systems is determined and by doing that the critical sub-systems are identified, on which dependability and safety techniques are to be applied during development. Proper performance of the software development requires the development of a technical specification for the products at the beginning of the life cycle. Such technical specification comprises both functional and non-functional requirements. These non-functional requirements address characteristics of the product such as quality, dependability, safety and maintainability. Software in space systems is more and more used in critical functions. Also the trend towards more frequent use of COTS and reusable components pose new difficulties in terms of assuring reliable and safe systems. Because of this, its dependability and safety must be carefully analysed. ESA identified and documented techniques, methods and procedures to ensure that software dependability and safety requirements are specified and taken into account during the design and development of a software system and to verify/validate that the implemented software systems comply with these requirements [R1].

  16. 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

  17. Nuclear safety culture evaluation model based on SSE-CMM

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. A new methodology for repository site suitability evaluation

    International Nuclear Information System (INIS)

    Miller, I.; Kossik, R.; Cunnane, M.

    1992-01-01

    Golder Associates Inc. (GAI) has developed a probabilistic total system performance assessment and strategy evaluation model (RIP) which can be applied in an iterative manner to evaluate repository site suitability and guide site characterization. The major portion of the software is the performance assessment model, which consists of a series of coupled component models for radionuclide transfer. The performance model itself is embedded within a decision analysis model which allows the user to evaluate alternative site characterization strategies. This paper provides an overview of the methodology, and summarizes the basic concepts of RIP

  19. Application of a new methodology to evaluate Dnb limits based on statistical propagation of uncertainties

    International Nuclear Information System (INIS)

    Machado, Marcio Dornellas

    1998-09-01

    One of the most important thermalhydraulics safety parameters is the DNBR (Departure from Nucleate Boiling Ratio). The current methodology in use at Eletronuclear to determine DNBR is extremely conservative and may result in penalties to the reactor power due to an increase plugging level of steam generator tubes. This work uses a new methodology to evaluate DNBR, named mini-RTDP. The standard methodology (STDP) currently in use establishes a limit design value which cannot be surpassed. This limit value is determined taking into account the uncertainties of the empirical correlation used in COBRA IIC/MIT code, modified to Angra 1 conditions. The correlation used is the Westinghouse's W-3 and the minimum DNBR (MDBR) value cannot be less than 1.3. The new methodology reduces the excessive level of conservatism associated with the parameters used in the DNBR calculation, which take most unfavorable values in the STDP methodology, by using their best estimate values. The final goal is to obtain a new DNBR design limit which will provide a margin gain due to more realistic parameters values used in the methodology. (author)

  20. 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

  1. 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.

  2. Evaluation of BOR-60 operation safety

    International Nuclear Information System (INIS)

    Minakov, A.A.; Antipin, G.K.; Efimov, V.N.; Kuzin, G.G.; Eschenko, L.V.; Eschenko, S.N.

    1987-12-01

    In this communication, BOR-60 reactor operation anomalies capable to produce a dangerous overheating of the core (SDC) is examined. On bases of calculations and reactor operation experience an event tree for SDC is built. Evaluations of probable anomalies entering in the event tree and reactor parameters modifications in case of anomalies are presented. In conclusion BOR-60 agree with the sovietic nuclear safety [fr

  3. Development of a comprehensive management site evaluation methodology

    International Nuclear Information System (INIS)

    Rodgers, J.C.; Onishi, Y.

    1981-01-01

    The Nuclear Regulatory Commission is in the process of preparing regulations that will define the necessary conditions for adequate disposal of low-level waste (LLW) by confinement in an LLW disposal facility. These proposed regulations form the context in which the motivation for the joint Los Alamos National Laboratory Battelle Pacific Northwest Laboratory program to develop a site-specific, LLW site evaluation methodology is discussed. The overall effort is divided into three development areas: land-use evaluation, environmental transport modelling, and long term scenario development including long-range climatology projections. At the present time four steps are envisioned in the application of the methodology to a site: site land use suitability assessment, land use-ecosystem interaction, contaminant transport simulation, and sensitivity analysis. Each of these steps is discussed in the paper. 12 refs

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

    International Nuclear Information System (INIS)

    2004-07-01

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

  5. 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

  6. METHODOLOGICAL PRINCIPLES AND METHODS OF TERMS OF TRADE STATISTICAL EVALUATION

    Directory of Open Access Journals (Sweden)

    N. Kovtun

    2014-09-01

    Full Text Available The paper studies the methodological principles and guidance of the statistical evaluation of terms of trade for the United Nations classification model – Harmonized Commodity Description and Coding System (HS. The practical implementation of the proposed three-stage model of index analysis and estimation of terms of trade for Ukraine's commodity-members for the period of 2011-2012 are realized.

  7. Safety evaluation of cation-exchange resins

    International Nuclear Information System (INIS)

    Kalkwarf, D.R.

    1977-08-01

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

  8. 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)

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

    International Nuclear Information System (INIS)

    Galia, A.V.

    2011-01-01

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

  10. Fusion integral experiments and analysis and the determination of design safety factors - I: Methodology

    International Nuclear Information System (INIS)

    Youssef, M.Z.; Kumar, A.; Abdou, M.A.; Oyama, Y.; Maekawa, H.

    1995-01-01

    The role of the neutronics experimentation and analysis in fusion neutronics research and development programs is discussed. A new methodology was developed to arrive at estimates to design safety factors based on the experimental and analytical results from design-oriented integral experiments. In this methodology, and for a particular nuclear response, R, a normalized density function (NDF) is constructed from the prediction uncertainties, and their associated standard deviations, as found in the various integral experiments where that response, R, is measured. Important statistical parameters are derived from the NDF, such as the global mean prediction uncertainty, and the possible spread around it. The method of deriving safety factors from many possible NDFs based on various calculational and measuring methods (among other variants) is also described. Associated with each safety factor is a confidence level, designers may choose to have, that the calculated response, R, will not exceed (or will not fall below) the actual measured value. An illustrative example is given on how to construct the NDFs. The methodology is applied in two areas, namely the line-integrated tritium production rate and bulk shielding integral experiments. Conditions under which these factors could be derived and the validity of the method are discussed. 72 refs., 17 figs., 4 tabs

  11. Evaluation of analytical performance based on partial order methodology.

    Science.gov (United States)

    Carlsen, Lars; Bruggemann, Rainer; Kenessova, Olga; Erzhigitov, Erkin

    2015-01-01

    Classical measurements of performances are typically based on linear scales. However, in analytical chemistry a simple scale may be not sufficient to analyze the analytical performance appropriately. Here partial order methodology can be helpful. Within the context described here, partial order analysis can be seen as an ordinal analysis of data matrices, especially to simplify the relative comparisons of objects due to their data profile (the ordered set of values an object have). Hence, partial order methodology offers a unique possibility to evaluate analytical performance. In the present data as, e.g., provided by the laboratories through interlaboratory comparisons or proficiency testings is used as an illustrative example. However, the presented scheme is likewise applicable for comparison of analytical methods or simply as a tool for optimization of an analytical method. The methodology can be applied without presumptions or pretreatment of the analytical data provided in order to evaluate the analytical performance taking into account all indicators simultaneously and thus elucidating a "distance" from the true value. In the present illustrative example it is assumed that the laboratories analyze a given sample several times and subsequently report the mean value, the standard deviation and the skewness, which simultaneously are used for the evaluation of the analytical performance. The analyses lead to information concerning (1) a partial ordering of the laboratories, subsequently, (2) a "distance" to the Reference laboratory and (3) a classification due to the concept of "peculiar points". Copyright © 2014 Elsevier B.V. All rights reserved.

  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. 21 CFR 315.6 - Evaluation of safety.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 5 2010-04-01 2010-04-01 false Evaluation of safety. 315.6 Section 315.6 Food and... USE DIAGNOSTIC RADIOPHARMACEUTICALS § 315.6 Evaluation of safety. (a) Factors considered in the safety...)(1) To establish the safety of a diagnostic radiopharmaceutical, FDA may require, among other...

  14. INPRO Methodology to evaluate the Mexico nuclear energy system

    International Nuclear Information System (INIS)

    Cruz S, R. R.; Martin del C, C.

    2016-09-01

    The International Atomic Energy Agency has developed the so-called International Project on Fuel Cycles and Innovative Nuclear Reactors (INPRO), in order to make nuclear energy available to meet the energy needs of the 21 century, in a sustainable way. One of the tasks of the project is the evaluation of the nuclear systems, to check whether they meet the objectives of the project and whether they are sustainable. This paper explains the rationale and general characteristics of the project in the evaluation of nuclear energy systems based on the concept of sustainable development. It describes the methodology developed to carry out this evaluation, divided into seven areas, such as economic, environmental, security, etc., which together make up the sustainable development of energy through nuclear systems. The economic area is analyzed and the evaluation criteria and parameters established by INPRO are discussed, in order to evaluate the Mexican nuclear energy system using Nest (software developed within the same project) as a tool to support the economic evaluation of nuclear systems. Based on the energy strategy proposed by the Energy Secretary of the Mexican Government which seeks to reduce the greenhouse gas emissions from the national electricity generation park, two types of reactor of currently available technology (A BWR and AP1000), were compared and these in turn with other alternative energy generation technologies, such as combined cycle, geothermal and wind plants. Also, the results of the application of the INPRO methodology are presented. Finally, the recommendations on actions that could lead the Mexican nuclear energy system towards sustainable development and conclusions on the application of the methodology to the Mexican case are mentioned. (Author)

  15. Multinational Design Evaluation Programme (MDEP) - Safety Goals

    International Nuclear Information System (INIS)

    Vaughan, G.J.

    2011-01-01

    One of the aims of the NEA's Multinational Design Evaluation Programme (MDEP) is to work towards greater harmonisation of regulatory requirements. To achieve this aim, it is necessary that there is a degree of convergence on the safety goals that are required to be met by designers and operators. The term 'safety goals' is defined to cover all health and safety requirements which must be met: these may be deterministic rules and/or probabilistic targets. They should cover the safety of workers, public and the environment in line with the IAEA's Basic Safety Objective; encompassing safety in normal operation through to severe accidents. MDEP is also interested in how its work can be extended to future reactors, which may use significantly different technology to the almost ubiquitous LWRs used today and in the next generation, building on the close co-operation within MDEP between the regulators who are currently engaged in constructing or carrying out design reviews on new designs. For two designs this work has involved several regulators sharing their safety assessments and in some cases issuing statements on issues that need to be addressed. Work is also progressing towards joint regulatory position statements on specific assessment areas. Harmonisation of safety goals will enhance the cooperation between regulators as further developments in design and technology occur. All regulators have safety goals, but these are expressed in many different ways and exercises in comparing them frequently are done at a very low level eg specific temperatures in the reactor vessel of a specific reactor type. The differences in the requirements from different regulators are difficult to resolve as the goals are derived using different principles and assumptions and are often for a specific technology. Therefore a different approach is being investigated, starting with the top-level safety goals and try to derive a structure and means of deriving lower tier

  16. Methodology to evaluate the insecticide potential of forest tree species

    International Nuclear Information System (INIS)

    Morales Soto, Leon; Garcia P, Carlos Mario

    2000-01-01

    The flora diversity of Colombia has an enormous potential in the rational use of its forest resources. Trees with biocidal effects to control pests and diseases need to be investigated. The objective of this research was to develop a methodology with low costs, easy application and quick results. The methodology employed was as follows: selection of tree species based on bibliography, ancestral reports and personal observations. The process was as follows: field collection of plants, preparation of plants extracts and test with Artemia salina Leach to detect biological activity of the extracts using LC50. Bioassays with those extract more promising (LC50 less than 1000 ppm) Determination of active compounds. The methodology was employed with 5 forest tree species: guarea guidonia (L) Sleumer and trichia hirta L. (Meliaceae), Machaerium Moritzianum Benth. (Fabaceae), Swinglea glutinosa Merrill (rutaceae) and Mammea americana L. (Clusiaceae). Using Artemia salina Leach as indicator of biocidal potential, two species were selected as the most promising, those were: Swinglea glutinosa Merril and Machaerium moritzianum Benth. In addition bioassays were made to evaluate fagoinhibition on Atta cephalotes (L.) (Hym: Formicidae) and control of Alconeura. This methodology is recommended for this kind of research

  17. Methodology for the insecticide potential evaluation of forest species

    International Nuclear Information System (INIS)

    Morales Soto, Leon; Garcia P, Carlos Mario

    2000-01-01

    The flora diversity of Colombia has an enormous potential in the rational use of its forest resources. Trees with biocides effects to pest control and diseases need to be investigated. The objective of this research was to develop a methodology with low costs, easy application and quick results. The methodology employed was as follows: selection of tree species based on bibliography, ancestral reports and personal observations. The process was as follows: field collection of plants, preparation of plants extracts, and test with Artemia Salina, leach to detect biological activity of the extracts using LC50. Bioassays with those extract more promising (LC50 less than 1000 ppm). The methodology was employed with 5 forest tree species: Guarea guidonia (L.) Sleumer and Trichia hirta L. (Meliaceae), machaerium moritzianum Benth. (Fabaceae), Swinglea glutinosa Merrill. (Rutaceae) and Mammea americana L. (Clusiaceae). Using Artemia salina Leach as indicator of biocidal potential, two species were selected as the most promising, those were: Swinglea glutinosa Merril and Machaerium moritzianum Benth. In addition bioassays were made to evaluate fagoinhibition on Atta cephalotes (L.) and control of Alconeura. This methodology is recommended for this kind of research

  18. 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

  19. 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

    There is much common ground between the nuclear industry and other major hazard industries such as those subject to the Seveso II regulations, e.g. oil, gas and chemicals. They are all subject to legal requirements to identify and control hazards, and to demonstrate that all necessary measures have been taken to minimise risks posed by the site with regard to people and the environment. This places a requirement on the Operators of major hazard installations, whether nuclear or conventional, to understand and identify the hazards of their operations, the initiating events, the consequences, the prevention and mitigation measures. However, in the UK, nuclear and 'Seveso' type facilities seem to adopt a different approach to the presentation of their safety cases. Given the magnitude of the hazards, safety cases developed for nuclear fuel cycle facilities are rigorous, detailed and complex, which can have the effect of reducing the visibility of the key hazards and corresponding protective measures. In contrast, on installations in the oil and gas and chemical industries, a real attempt has been made over recent years to improve the visibility and accessibility of the safety case to all operating personnel, through the use of visual aids / diagrams. In particular, many Operators are choosing to use 'bow tie methodology', in which very simple overview diagrams are produced to illustrate, in a form understandable by all: - what the key hazards are; - the initiating events; - the consequences of an incident; - the barriers or 'Layers of Protection' which prevent an initiating event from developing into an incident; - the barriers or 'Layers of Defence' which mitigate the consequences of an incident, i.e. which prevent the incident from escalating into major consequences. The bow tie method is one of a number of methodologies that can be used to make safety cases more accessible. It is used in this paper to illustrate ways to

  20. Evaluation of methodology for QRA in offshore operations

    International Nuclear Information System (INIS)

    Vinnem, Jan Erik

    1998-01-01

    A brief summary is presented of the development of Quantified Risk Analysis in the offshore oil and gas industry for nearly 20 years. From a modest start in Norway as a research activity, QRA has today become one of the focal points of oil and gas management in all countries around the North Sea, and also in other areas. This was mainly initiated by authority regulations and requirements, but has now for some time been driven by the industry itself, as it sees the QRA as a vehicle to gain extended flexibility with respect to optimization of offshore safety standards. This article discusses how QRAs are used, and examines in this light how methodology developments that are and should be carried out by R and D institutions and the industry. Recommendations with respect to how and where knowledge should be developed are expressed and discussed. Lastly visions for future use and developments within the QRA technology are outlined

  1. 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

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, S. S. [BNF Technology, Taejon (Korea, Republic of); Cha, K. H.; Lee, C. S.; Kwon, K. C.; Han, H. B. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-01

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

  5. A probabilistic bridge safety evaluation against floods.

    Science.gov (United States)

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

    2016-01-01

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

  6. Evaluating the effectiveness of active vehicle safety systems.

    Science.gov (United States)

    Jeong, Eunbi; Oh, Cheol

    2017-03-01

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

  7. Model evaluation methodology applicable to environmental assessment models

    International Nuclear Information System (INIS)

    Shaeffer, D.L.

    1979-08-01

    A model evaluation methodology is presented to provide a systematic framework within which the adequacy of environmental assessment models might be examined. The necessity for such a tool is motivated by the widespread use of models for predicting the environmental consequences of various human activities and by the reliance on these model predictions for deciding whether a particular activity requires the deployment of costly control measures. Consequently, the uncertainty associated with prediction must be established for the use of such models. The methodology presented here consists of six major tasks: model examination, algorithm examination, data evaluation, sensitivity analyses, validation studies, and code comparison. This methodology is presented in the form of a flowchart to show the logical interrelatedness of the various tasks. Emphasis has been placed on identifying those parameters which are most important in determining the predictive outputs of a model. Importance has been attached to the process of collecting quality data. A method has been developed for analyzing multiplicative chain models when the input parameters are statistically independent and lognormally distributed. Latin hypercube sampling has been offered as a promising candidate for doing sensitivity analyses. Several different ways of viewing the validity of a model have been presented. Criteria are presented for selecting models for environmental assessment purposes

  8. A methodology to evaluate the fatigue life of flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Sousa, Fernando J.M. de; Sousa, Jose Renato M. de; Siqueira, Marcos Q. de; Sagrilo, Luis V.S. [Coordenacao dos Programas de Pos-graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Lemos, Carlos Alberto D. de [Petroleo Brasileiro S.A. (PETROBRAS), Rio de Janeiro, RJ (Brazil)

    2009-07-01

    This paper focus on a methodology to perform the fatigue analysis of flexible pipes. This methodology employs functions that convert forces and moments obtained in global analyses into stresses. The stresses are then processed by well-known cycle counting methods, and S-N curves evaluate the damage at several points in the pipe cross-section. Palmgren-Miner linear damage hypothesis is assumed in order to calculate the accumulated fatigue damage. A parametric study on the fatigue life of a flexible pipe employing this methodology is presented. The main points addressed in the study are the influence of friction between layers in the results, the importance of evaluating the fatigue life in various points of the pipe cross-section and the effect of different mean stress levels. The obtained results suggest that the consideration of friction effects strongly influences the fatigue life of flexible risers and these effects have to be accounted both in the global and local analyses of the riser. Moreover, mean stress effects are also significant and at least 8 equally spaced wires in each analyzed section of the riser must be considered in fatigue analyses. (author)

  9. Study on GIF PR/PP evaluation methodology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. H.; Kwon, E. H.; Kim, H. D. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Proliferation resistance (PR) and physical protection (PP) is one of the four technology goals of generation IV nuclear energy systems (NESs). The PR component of the goal focuses on providing strong assurance that generation IV NESs are the least desirable sources for the diversion or undeclared production of nuclear materials, whereas the PP portion of the goal ensures that generation IV NESs are robust against theft and sabotage. In 2002, the road map of the Generation IV International Forum (GIF) envisioned that the R and D program for PR and PP would have three areas: 1) safeguards and physical protection technology R and D for each GIF system; 2) formulation of PR and PP criteria and metrics; and 3) evaluation of the criteria and metrics. To cover these R and D items, the PR and PP Working Group (PRPPWG) was formed in late 2002 and has since developed a methodology for PR and PP evaluation. In a succession of revisions beginning in 2004, consensus was achieved amongst all participating GIF countries and related organizations (i.e., IAEA and EU), and Revision 6 of the methodology report was approved by GIF for open distribution in 2011. The paper describes in detail the methodology developed by the PRPPWG and discusses its applicability to the sodium-cooled fast reactor (SFR) fuel cycle with pyro processing currently under development in Korea.

  10. Barrier performance researches for the safety evaluation

    International Nuclear Information System (INIS)

    Niibori, Yuichi

    2004-01-01

    So far, many researches were conducted to propose a scientific evidence (a safety case) for the realization of geological disposal in Japan. In order to regulate the geological disposal system of radioactive wastes, on the other hand, we need also a holistic approach to integrate various data related for the performance evaluations of the engineered barrier system and the natural barrier system. However, the scientific bases are not sufficient to establish the safety regulation for such a natural system. For example, we often apply the specific probability density function (PDF) to the uncertainty of barrier system due to the essential heterogeneity. However, the applicability is not clear in the regulation point of view. A viewpoint to understand such an applicability of PDFs has been presented. (author)

  11. 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

    The Department of Energy has assigned to Sandia National Laboratories the responsibility of producing a Safety Analysis Report (SAR) for the plutonium-dioxide fueled Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) proposed to be used in the Mars Science Laboratory (MSL) mission. The National Aeronautic and Space Administration (NASA) is anticipating a launch in fall of 2009, and the SAR will play a critical role in the launch approval process. As in past safety evaluations of MMRTG missions, a wide range of potential accident conditions differing widely in probability and seventy must be considered, and the resulting risk to the public will be presented in the form of probability distribution functions of health effects in terms of latent cancer fatalities. The basic descriptions of accident cases will be provided by NASA in the MSL SAR Databook for the mission, and on the basis of these descriptions, Sandia will apply a variety of sophisticated computational simulation tools to evaluate the potential release of plutonium dioxide, its transport to human populations, and the consequent health effects. The first step in carrying out this project is to evaluate the existing computational analysis tools (computer codes) for suitability to the analysis and, when appropriate, to identify areas where modifications or improvements are warranted. The overall calculation of health risks can be divided into three levels of analysis. Level A involves detailed simulations of the interactions of the MMRTG or its components with the broad range of insults (e.g., shrapnel, blast waves, fires) posed by the various accident environments. There are a number of candidate codes for this level; they are typically high resolution computational simulation tools that capture details of each type of interaction and that can predict damage and plutonium dioxide release for a range of choices of controlling parameters. Level B utilizes these detailed results to study many

  12. Methodologies for evaluating the proliferation resistance of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Shiotani, Hiroki; Hori, Kei-ichiro; Takeda, Hiroshi

    2001-01-01

    The Japan Nuclear Cycle Development Institute (JNC) believes that the development of future nuclear fuel cycle technology should be conducted with careful consideration given to non-proliferation. JNC is studying methodologies for evaluating proliferation resistance of nuclear fuel cycle technologies. However, it is difficult to establish the methodology for evaluating proliferation resistance since the results greatly depend on the assumption for the evaluation and the surrounding conditions. This study grouped factors of proliferation resistance into categories through reviewing past studies and studied the relationships between the factors. Then, this study tried to find vulnerable nuclear material (plutonium) in some FBR fuel cycles from the proliferation perspective, and calculated the time it takes to convert the materials from various nuclear fuel cycles into pure plutonium metal under some assumptions. The result showed that it would take a long time to convert the nuclear materials from the FBR fuel cycles without plutonium separation. While it is a preliminary attempt to evaluate a technical factor of proliferation resistance as the basis of the institutional proliferation resistance, the JNC hopes that it will contribute to future discussions in this area. (author)

  13. The use of fracture mechanics methodologies for NDT results evaluation and comparison

    International Nuclear Information System (INIS)

    Reale, S.

    1995-01-01

    In the general frame of analysing the interactions amongst the information from non destructive evaluation (NDE) and the methodologies to assess the integrity of a defective structure (such as fracture mechanics), the aim of the paper is to analyse and compare, in terms of indices related to safety margins, NDE results from round robin testing trials to acheive assessments of capabilities and limitations.A structural integrity/fracture mechanics approach for evaluating and comparing results from non destructive techniques is presented. Safety factors can be associated to flaws detected and characterized by inspections (estimated flaws) and to flaws actually present (reference flaws). The mismatch between safety factors associated to estimated flaws and safety factors associated to reference flaws can be used to assess capabilities and limitations of procedures and techniques in use for inspections.As an example, to show how the above procedure is applied and its potential as a method of data evaluation and comparison, the NDE results produced by the PISC (project for the inspection of steel components) activity have been considered. (orig.)

  14. Methodology for evaluating port vulnerability to nuclear smuggling

    International Nuclear Information System (INIS)

    Ek, D.; Gronager, J.R.; Blankenship, J.A.; Martin, D.

    2001-01-01

    Full text: Background: Fueled by an increase in intercepted nuclear smuggling events, the threat of nuclear smuggling has received increased attention in recent years. This attention has resulted in a focused effort to improve the ability to deter or detect smuggling attempts through border crossings, including seaports, airports, and rail and road crossings. These efforts have primarily been focused on installing SNM detectors across vehicle and pedestrian gates entering these ports. However, the effectiveness of this application in deterring or detecting events has not been carefully evaluated. A recent effort was undertaken to evaluate in detail the susceptibility of an international seaport and airport to nuclear smuggling. The evaluation considered a range of adversary profiles to match these against existing and proposed port security measures and equipment. The evaluation was pursued using path analysis methodologies, which were adapted to the port environment. As a result of limited data concerning the effectiveness of patrol, search, and access control procedures at the port, an assessment methodology was developed to estimate these in a standardized fashion. The methodology considers a detailed list of tasks each type of adversary must successfully accomplish for any particular smuggling scenario and path through the port. Within these tasks, locations or times of potential detection are identified. From a look-up table, a detection level (Low, Medium, or High) is assigned to each detection potential based upon the type of detection possible and considering the possible access or authority of each adversary. The overall detection potential in determined as a sum of these individual detection potentials according to the equation: P t ={1-Σ(1-P n ). Where: P t is the total detection potential for an adversary path, and P n is the individual detection at a particular location or time. The evaluation revealed that the current process of installing portals at

  15. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  16. Safety evaluation for packaging (onsite) SERF cask

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1997-01-01

    This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

  17. Safety evaluation for packaging CPC metal boxes

    International Nuclear Information System (INIS)

    Romano, T.

    1995-01-01

    This Safety Evaluation for Packaging (SEP) provides authorization for the use of Container Products Corporation (CPC) metal boxes, as described in this document, for the interarea shipment of radioactive contaminated equipment and debris for storage in the Central Waste Complex (CWC) or T Plant located in the 200 West Area. Authorization is granted until November 30, 1995. The CPC boxes included in this SEP were originally procured as US Department of Transportation (DOT) Specification 7A Type A boxes. A review of the documentation provided by the manufacturer revealed the documentation did not adequately demonstrate compliance to the 4 ft drop test requirement of 49 CFR 173.465(c). Preparation of a SEP is necessary to document the equivalent safety of the onsite shipment in lieu of meeting DOT packaging requirements until adequate documentation is received. The equivalent safety of the shipment is based on the fact that the radioactive contents consist of contaminated equipment and debris which are not dispersible. Each piece is wrapped in two layers of no less than 4 mil plastic prior to being placed in the box which has an additional 10 mil liner. Pointed objects and sharp edges are padded to prevent puncture of the plastic liner and wrapping

  18. 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)

  19. Reservoir shorelines : a methodology for evaluating operational impacts

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, M.; Braund-Read, J.; Musgrave, B. [BC Hydro, Burnaby, BC (Canada)

    2009-07-01

    BC Hydro has been operating hydroelectric facilities for over a century in British Columbia. The integrity and stability of the shorelines and slopes bordering hydroelectric reservoirs is affected by changing water levels in the reservoir, natural processes of flooding, wind and wave action and modification of groundwater levels. Establishing setbacks landward of the shoreline are needed in order to protect useable shoreline property that may be at risk of flooding, erosion or instability due to reservoir operations. Many of the reservoirs in British Columbia are situated in steep, glaciated valleys with diverse geological, geomorphological and climatic conditions and a variety of eroding shorelines. As such, geotechnical studies are needed to determine the operational impacts on reservoir shorelines. Since the 1960s BC Hydro has been developing a methodology for evaluating reservoir impacts and determining the land around the reservoir perimeter that should remain as a right of way for operations while safeguarding waterfront development. The methodology was modified in the 1990s to include geomorphological and geological processes. However, uncertainties in the methodology still exist due to limited understanding of key issues such as rates of erosion and shoreline regression, immaturity of present day reservoir shorelines and impacts of climate change. 11 refs., 1 tab., 7 figs.

  20. The evaluation framework for business process management methodologies

    Directory of Open Access Journals (Sweden)

    Sebastian Lahajnar

    2016-06-01

    Full Text Available In an intense competition in the global market, organisations seek to take advantage of all their internal and external potentials, advantages, and resources. It has been found that, in addition to competitive products and services, a good business also requires an effective management of business processes, which is the discipline of the business process management (BPM. The introduction of the BPM in the organisation requires a thoughtful selection of an appropriate methodological approach, since the latter will formalize activities, products, applications and other efforts of the organisation in this field. Despite many technology-driven solutions of software companies, recommendations of consulting companies, techniques, good practices and tools, the decision on what methodology to choose is anything but simple. The aim of this article is to simplify the adoption of such decisions by building a framework for the evaluation of BPM methodologies according to a qualitative multi-attribute decision-making method. The framework defines a hierarchical decision-making model, formalizes the decision-making process and thus contributes significantly to an independent, credible final decision that is the most appropriate for a specific organisation.

  1. A bench-scale biotreatability methodology to evaluate field bioremediation

    International Nuclear Information System (INIS)

    Saberiyan, A.G.; MacPherson, J.R. Jr.; Moore, R.; Pruess, A.J.; Andrilenas, J.S.

    1995-01-01

    A bench-scale biotreatability methodology was designed to assess field bioremediation of petroleum contaminated soil samples. This methodology was performed successfully on soil samples from more than 40 sites. The methodology is composed of two phases, characterization and experimentation. The first phase is physical, chemical, and biological characterization of the contaminated soil sample. This phase determines soil parameters, contaminant type, presence of indigenous contaminant-degrading bacteria, and bacterial population size. The second phase, experimentation, consists of a respirometry test to measure the growth of microbes indirectly (via generation of CO 2 ) and the consumption of their food source directly (via contaminant loss). Based on a Monod kinetic analysis, the half-life of a contaminant can be calculated. Abiotic losses are accounted for based on a control test. The contaminant molecular structure is used to generate a stoichiometric equation. The stoichiometric equation yields a theoretical ratio for mg of contaminant degraded per mg of CO 2 produced. Data collected from the respirometry test are compared to theoretical values to evaluate bioremediation feasibility

  2. European Healthy Cities evaluation: conceptual framework and methodology.

    Science.gov (United States)

    de Leeuw, Evelyne; Green, Geoff; Dyakova, Mariana; Spanswick, Lucy; Palmer, Nicola

    2015-06-01

    This paper presents the methodology, programme logic and conceptual framework that drove the evaluation of the Fifth Phase of the WHO European Healthy Cities Network. Towards the end of the phase, 99 cities were designated progressively through the life of the phase (2009-14). The paper establishes the values, systems and aspirations that these cities sign up for, as foundations for the selection of methodology. We assert that a realist synthesis methodology, driven by a wide range of qualitative and quantitative methods, is the most appropriate perspective to address the wide geopolitical, demographic, population and health diversities of these cities. The paper outlines the rationale for a structured multiple case study approach, the deployment of a comprehensive questionnaire, data mining through existing databases including Eurostat and analysis of management information generation tools used throughout the period. Response rates were considered extremely high for this type of research. Non-response analyses are described, which show that data are representative for cities across the spectrum of diversity. This paper provides a foundation for further analysis on specific areas of interest presented in this supplement. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  3. Evaluating the Staff at Enterprise: Several Theoretical and Methodological Aspects

    Directory of Open Access Journals (Sweden)

    Girman Alla P.

    2017-03-01

    Full Text Available The article is aimed at generalizing and systematizing various knowledge, related to evaluation of staff, on a common theoretical-methodological basis. Concept, objectives, directions, methods, and indicators for evaluating staff in the contemporary economy were analyzed. The topicality of using the theoretical developments on staff evaluation in actual practice of functioning of enterprises has been substantiated. A new approach to the procedure of evaluation of the total human resource of enterprise, based on the life cycle of organization, has been proposed. On the basis of the proposed scientific algorithmic step-by-step approach to the evaluation of staff, managers of companies can design their own models for staff evaluation, develop its separate elements. Prospects for further researches in this direction involve relation of staff evaluation to the life cycle of employee no less than the life cycle of enterprise. Management of the life cycle of employee represents methods for management of his development that would change the level of the employee’s professional maturity as result of a system impact.

  4. Preliminary safety evaluation for CSR1000 with passive safety system

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

    2014-01-01

    Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

  5. Methodology for safety evaluation of existing timber elements

    OpenAIRE

    Sousa, Hélder S.

    2013-01-01

    Tese de doutoramento em Engenharia Civil The past decades have seen a renewed interest in timber research related topics. As instance, timber engineering is not only confined to the design of new construction with innovative wood based products, but also an increasing awareness for the preservation of existing timber structures is visible in nowadays society. Therefore, arises the need for a better understanding of timber performance as a construction material aiming at a bette...

  6. Safety evaluation of the Dalat research reactor operation

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  7. [Safety evaluation of niuhuang jiedu tablet].

    Science.gov (United States)

    Feng, Yu-Ling; Miao, Jia-wei; Li, Jing; Sung, An-Sheng; Liu, Jie

    2014-09-01

    Realgar-containing Niuhuang Jiedu tablet (NHJD) has been applied in clinic for more than 800 years. However, because realgar contains arsenic (As), it has aroused wide concerns and controversies both at home and abroad. Currently, there are two misunderstandings about realgar-containing Chinese patent medicines. First, some people exaggerated realgar's toxicity as that of arsenic. Second, they recommended to remove realgar from traditional Chinese medicine compounds. In this paper, the authors summarized the advance in studies on NHJD, and proposed different opinions: (1) It is inappropriate to take total As as the index in safety evaluation of NHJD. (2) The toxicity of NHJD is dependent on the dose and duration of administration. (3) Realgar is an active ingredient of NHJD, and shall be deeply studied. Classic realgar-containing traditional Chinese medicine prescriptions, such as Niuhuang Jiedu tablet, shall be evaluated with rigorous modern scientific basis, with the aim to guide rational and safe application.

  8. 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

  9. The Sheep as a Model of Preclinical Safety and Pharmacokinetic Evaluations of Candidate Microbicides

    OpenAIRE

    Holt, Jonathon D. S.; Cameron, David; Dias, Nicola; Holding, Jeremy; Muntendam, Alex; Oostebring, Freddy; Dreier, Peter; Rohan, Lisa; Nuttall, Jeremy

    2015-01-01

    When developing novel microbicide products for the prevention of HIV infection, the preclinical safety program must evaluate not only the active pharmaceutical ingredient but also the product itself. To that end, we applied several relatively standard toxicology study methodologies to female sheep, incorporating an assessment of the pharmacokinetics, safety, tolerability, and local toxicity of a dapivirine-containing human vaginal ring formulation (Dapivirine Vaginal Ring-004). We performed a...

  10. Accidental safety analysis methodology development in decommission of the nuclear facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-15

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

  11. Nuclear data evaluation methodology including estimates of covariances

    Directory of Open Access Journals (Sweden)

    Smith D.L.

    2010-10-01

    Full Text Available Evaluated nuclear data rather than raw experimental and theoretical information are employed in nuclear applications such as the design of nuclear energy systems. Therefore, the process by which such information is produced and ultimately used is of critical interest to the nuclear science community. This paper provides an overview of various contemporary methods employed to generate evaluated cross sections and related physical quantities such as particle emission angular distributions and energy spectra. The emphasis here is on data associated with neutron induced reaction processes, with consideration of the uncertainties in these data, and on the more recent evaluation methods, e.g., those that are based on stochastic (Monte Carlo techniques. There is no unique way to perform such evaluations, nor are nuclear data evaluators united in their opinions as to which methods are superior to the others in various circumstances. In some cases it is not critical which approaches are used as long as there is consistency and proper use is made of the available physical information. However, in other instances there are definite advantages to using particular methods as opposed to other options. Some of these distinctions are discussed in this paper and suggestions are offered regarding fruitful areas for future research in the development of evaluation methodology.

  12. Development of an Evaluation Methodology for Loss of Large Area induced from extreme events

    International Nuclear Information System (INIS)

    Kim, Sok Chul; Park, Jong Seuk; Kim, Byung Soon; Jang, Dong Ju; Lee, Seung Woo

    2015-01-01

    USNRC announced several regulatory requirements and guidance documents regarding the event of loss of large area including 10CFR 50.54(hh), Regulatory Guide 1.214 and SRP 19.4. In Korea, consideration of loss of large area has been limitedly taken into account for newly constructing NPPs as voluntary based. In general, it is hardly possible to find available information on methodology and key assumptions for the assessment of LOLA due to 'need to know based approach'. Urgent needs exists for developing country specific regulatory requirements, guidance and evaluation methodology by themselves with the consideration of their own geographical and nuclear safety and security environments. Currently, Korea Hydro and Nuclear Power Company (KHNP) has developed an Extended Damage Mitigation Guideline (EDMG) for APR1400 under contract with foreign consulting company. The submittal guidance NEI 06-12 related to B.5.b Phase 2 and 3 focused on unit-wise mitigation strategy instead of site level mitigation or response strategy. Phase 1 mitigating strategy and guideline for LOLA (Loss of Large Area) provides emphasis on site level arrangement including cooperative networking outside organizations and agile command and control system. Korea Institute of Nuclear Safety has carried out a pilot in-house research project to develop the methodology and guideline for evaluation of LOLA since 2014. This paper introduces the summary of major results and outcomes of the aforementioned research project. After Fukushima Dai-Ichi accident, the awareness on countering the event of loss of large area induced from extreme man-made hazards or extreme beyond design basis external event. Urgent need exists to develop regulatory guidance for coping with this undesirable situation, which has been out of consideration at existing nuclear safety regulatory framework due to the expectation of rare possibility of occurrence

  13. 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)

  14. 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.

  15. Methodology for fast evaluation of Bacillus thuringiensis crystal protein content

    Directory of Open Access Journals (Sweden)

    Alves Lúcia M. Carareto

    2000-01-01

    Full Text Available The development of the production and use of Bacillus thuringiensis in Brazil at a commercial scale faces certain difficulties, among them the establishment of efficient methodologies for the quantitation of toxic products to be commercialized. Presently, the amount of toxin is given in percentage by analyzing the samples total protein content. Such methodology however, does not measure the actual amount of active protein present in the product, since most strains express different endotoxin genes and might even produce b-toxin. Since the various types of toxins exhibit different antigenic characteristics, this work has as objective the utilization of fast immunological techniques to quantify the level of crystal protein. Crystal protein produced by a subspecies of Bacillus thuringiensis var. israelensis was purified by ultracentrifugation and utilized to immunize rabbits and to produce hiperimmune sera. Such sera were latter used to evaluate the level of proteins on commercial bioinsecticide and on laboratory cultures of B. thuringiensis through the immunodot technique. The results were obtained by comparison of data obtained from reactions with known concentrations of crystal protein permitting to evaluate the level of such protein on various materials.

  16. Strategic alternatives ranking methodology: Multiple RCRA incinerator evaluation test case

    International Nuclear Information System (INIS)

    Baker, G.; Thomson, R.D.; Reece, J.; Springer, L.; Main, D.

    1988-01-01

    This paper presents an important process approach to permit quantification and ranking of multiple alternatives being considered in remedial actions or hazardous waste strategies. This process is a methodology for evaluating programmatic options in support of site selection or environmental analyses. Political or other less tangible motivations for alternatives may be quantified by means of establishing the range of significant variables, weighting their importance, and by establishing specific criteria for scoring individual alternatives. An application of the process to a recent AFLC program permitted ranking incineration alternatives from a list of over 130 options. The process forced participation by the organizations to be effected, allowed a consensus of opinion to be achieved, allowed complete flexibility to evaluate factor sensitivity, and resulted in strong, quantifiable support for any subsequent site-selection action NEPA documents

  17. Development of an Evaluation Methodology for Loss of Large Area Induced from Extreme Events with Malicious Origin

    International Nuclear Information System (INIS)

    Kim, S.C.; Park, J.S.; Chang, D.J.; Kim, D.H.; Lee, S.W.; Lee, Y.J.; Kim, H.W.

    2016-01-01

    Event of loss of large area (LOLA) induced from extreme external event at multi-units nuclear installation has been emerged a new challenges in the realm of nuclear safety and regulation after Fukushima Dai-Ichi accident. The relevant information and experience on evaluation methodology and regulatory requirements are rarely available and negative to share due to the security sensitivity. Most of countries has been prepared their own regulatory requirements and methodologies to evaluate impact of LOLA at nuclear power plant. In Korea, newly amended the Nuclear Safety Acts requires to assess LOLA in terms of EDMG (Extended Damage Mitigation Guideline). Korea Institute of Nuclear Safety (KINS) has performed a pilot research project to develop the methodology and regulatory review guidance on LOLA at multi-units nuclear power plant since 2014. Through this research, we proposed a methodology to identify the strategies for preventive and mitigation of the consequences of LOLA utilizing PSA techniques or its results. The proposed methodology is comprised of 8 steps including policy consideration, threat evaluation, identification of damage path sets, SSCs capacity evaluation and identification of mitigation measures and strategies. The consequence of LOLA due to malevolent aircraft crash may significantly susceptible with analysis assumptions including type of aircraft, amount of residual fuel, and hittable angle and so on, which cannot be shared overtly. This paper introduces a evaluation methodology for LOLA using PSA technique and its results. Also we provide a case study to evaluate hittable access angle using flight simulator for two types of air crafts and to identify potential path sets leading to core damage by affected SSCs within damaged area.(author).

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

    Directory of Open Access Journals (Sweden)

    Олег Богданович ЗАЧКО

    2016-03-01

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

  19. 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)

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

    International Nuclear Information System (INIS)

    Yamauchi, Takamichi; Maki, Akira; Nojiri, Ichiro

    1999-02-01

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

  1. Does lean management improve patient safety culture? An extensive evaluation of safety culture in a radiotherapy institute.

    Science.gov (United States)

    Simons, Pascale A M; Houben, Ruud; Vlayen, Annemie; Hellings, Johan; Pijls-Johannesma, Madelon; Marneffe, Wim; Vandijck, Dominique

    2015-02-01

    The importance of a safety culture to maximize safety is no longer questioned. However, achieving sustainable culture improvements are less evident. Evidence is growing for a multifaceted approach, where multiple safety interventions are combined. Lean management is such an integral approach to improve safety, quality and efficiency and therefore, could be expected to improve the safety culture. This paper presents the effects of lean management activities on the patient safety culture in a radiotherapy institute. Patient safety culture was evaluated over a three year period using triangulation of methodologies. Two surveys were distributed three times, workshops were performed twice, data from an incident reporting system (IRS) was monitored and results were explored using structured interviews with professionals. Averages, chi-square, logistical and multi-level regression were used for analysis. The workshops showed no changes in safety culture, whereas the surveys showed improvements on six out of twelve dimensions of safety climate. The intention to report incidents not reaching patient-level decreased in accordance with the decreasing number of reports in the IRS. However, the intention to take action in order to prevent future incidents improved (factorial survey presented β: 1.19 with p: 0.01). Due to increased problem solving and improvements in equipment, the number of incidents decreased. Although the intention to report incidents not reaching patient-level decreased, employees experienced sustained safety awareness and an increased intention to structurally improve. The patient safety culture improved due to the lean activities combined with an organizational restructure, and actual patient safety outcomes might have improved as well. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. Development of a methodology for the evaluation of the thermomechanical behavior of the TRISO fuel

    International Nuclear Information System (INIS)

    Garcia, Lorena P. Rodríguez; Pérez, Daniel Milian; Hernández, Carlos Rafael García; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Universidad Federal de Pernambuco

    2017-01-01

    The use of the Generation IV Very High Temperature Reactors (VHTR) presents significant perspectives to assume the future nuclear energy and hydrogen production. VHTR has advantages because its low electricity generation costs, short construction periods, high hydrogen production efficiency, safety and reliability, proliferation resistance and inherent safety features of the fuel and reactor. However, it faces substantial challenges to be successfully deployed as a sustainable energy source. One of these key challenges is the nuclear safety which mainly relies on the quality and integrity of the coated fuel particles (TRISO) planned to be used in these reactors taking into consideration the high temperatures (1000°C in normal operation and up to 1800°C in accidents conditions) and burnup degrees (150 - 200 GWd/tonU) achievable in these reactors. In this paper is presented the current state of development of a methodology for the evaluation of the thermomechanical behavior of the TRISO fuel in function of the variation of different parameters in the VHTR. In order to achieve this goal will be used coupled computational modeling using analytical methods and Monte Carlo and CFD codes such as MCNPX version 2.6e and Ansys version 14. The studies performed in this investigation included the evaluation of key parameters in the TRISO such as the release of fission gases and CO, gas pressure, temperature distributions, kernel migration, maximum stress values, and failure probabilities. The results achieved in this investigation contributes to demonstrating the viability of the proposed methodology for the study, the design and safety calculations of VHTR. (author)

  3. Development of a methodology for the evaluation of the thermomechanical behavior of the TRISO fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, Lorena P. Rodríguez; Pérez, Daniel Milian; Hernández, Carlos Rafael García; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira, E-mail: lorenapilar1109@gmail.com, E-mail: milianperez89@gmail.com, E-mail: cgh@instec.cu, E-mail: dmilian@instec.cu, E-mail: cabol@ufpe.br [Higher Institute of Technologies and Applied Sciences (InSTEC), Habana (Cuba); Universidad Federal de Pernambuco (UFPE), Recife (Brazil). Departamento de Energia Nuclear

    2017-11-01

    The use of the Generation IV Very High Temperature Reactors (VHTR) presents significant perspectives to assume the future nuclear energy and hydrogen production. VHTR has advantages because its low electricity generation costs, short construction periods, high hydrogen production efficiency, safety and reliability, proliferation resistance and inherent safety features of the fuel and reactor. However, it faces substantial challenges to be successfully deployed as a sustainable energy source. One of these key challenges is the nuclear safety which mainly relies on the quality and integrity of the coated fuel particles (TRISO) planned to be used in these reactors taking into consideration the high temperatures (1000°C in normal operation and up to 1800°C in accidents conditions) and burnup degrees (150 - 200 GWd/tonU) achievable in these reactors. In this paper is presented the current state of development of a methodology for the evaluation of the thermomechanical behavior of the TRISO fuel in function of the variation of different parameters in the VHTR. In order to achieve this goal will be used coupled computational modeling using analytical methods and Monte Carlo and CFD codes such as MCNPX version 2.6e and Ansys version 14. The studies performed in this investigation included the evaluation of key parameters in the TRISO such as the release of fission gases and CO, gas pressure, temperature distributions, kernel migration, maximum stress values, and failure probabilities. The results achieved in this investigation contributes to demonstrating the viability of the proposed methodology for the study, the design and safety calculations of VHTR. (author)

  4. Safety evaluation of small samples for isotope production

    International Nuclear Information System (INIS)

    Sharma, Archana; Singh, Tej; Varde, P.V.

    2015-09-01

    Radioactive isotopes are widely used in basic and applied science and engineering, most notably as environmental and industrial tracers, and for medical imaging procedures. Production of radioisotope constitutes important activity of Indian nuclear program. Since its initial criticality DHRUVA reactor has been facilitating the regular supply of most of the radioisotopes required in the country for application in the fields of medicine, industry and agriculture. In-pile irradiation of the samples requires a prior estimation of the sample reactivity load, heating rate, activity developed and shielding thickness required for post irradiation handling. This report is an attempt to highlight the contributions of DHRUVA reactor, as well as to explain in detail the methodologies used in safety evaluation of the in pile irradiation samples. (author)

  5. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.M.; Ketcham, D.R.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table testing which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its ''Generic Safety Evaluation Report'' approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the United States and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluating program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  6. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.; Ketcham, D.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table tested which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its Generic Safety Evaluation Report approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the US and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective approach developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluation program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

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

    International Nuclear Information System (INIS)

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

    2008-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-12-15

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

  9. Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

    2010-01-01

    To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

  10. Design Methodology and Performance Evaluation of New Generation Sounding Rockets

    Directory of Open Access Journals (Sweden)

    Marco Pallone

    2018-01-01

    Full Text Available Sounding rockets are currently deployed for the purpose of providing experimental data of the upper atmosphere, as well as for microgravity experiments. This work provides a methodology in order to design, model, and evaluate the performance of new sounding rockets. A general configuration composed of a rocket with four canards and four tail wings is sized and optimized, assuming different payload masses and microgravity durations. The aerodynamic forces are modeled with high fidelity using the interpolation of available data. Three different guidance algorithms are used for the trajectory integration: constant attitude, near radial, and sun-pointing. The sun-pointing guidance is used to obtain the best microgravity performance while maintaining a specified attitude with respect to the sun, allowing for experiments which are temperature sensitive. Near radial guidance has instead the main purpose of reaching high altitudes, thus maximizing the microgravity duration. The results prove that the methodology at hand is straightforward to implement and capable of providing satisfactory performance in term of microgravity duration.

  11. Safety Evaluation of Roundabouts in Georgia

    Science.gov (United States)

    2018-02-28

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

  12. Methodology to evaluation of the density in radiographic image

    International Nuclear Information System (INIS)

    Louzada, M.J.Q.; Pela, C.A.; Belangero, W.D.; Santos-Pinto, R.

    1998-01-01

    This study was designed in order to optimize the optical densitometry technique in radiographic images by the setorization of the characteristic curves of the radiographic films. We used 24 radiographs of a stepped aluminium wedge that were taken without rigorous control development and manually revealed. The densitometric values of the steps images and its thickness, for each radiographic, was utilized to generate its particular mathematics expressions that represent its characteristic densitometric curves and then it were used for setorization. The densitometric values were obtained by a Macbeth TD528 densitometer. The study showed an optimization in the representation of the relationship between the optical density of the steps images of the wedge and its correspondent thickness, provided by the setorization, with mean square error around 10 -5 . This optimization will allow the use of this methodology in quantitative evaluations of bone mass, by radiographic images. (author)

  13. Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

    International Nuclear Information System (INIS)

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

    2010-07-01

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

  14. Development and evaluation of clicker methodology for introductory physics courses

    Science.gov (United States)

    Lee, Albert H.

    Many educators understand that lectures are cost effective but not learning efficient, so continue to search for ways to increase active student participation in this traditionally passive learning environment. In-class polling systems, or "clickers", are inexpensive and reliable tools allowing students to actively participate in lectures by answering multiple-choice questions. Students assess their learning in real time by observing instant polling summaries displayed in front of them. This in turn motivates additional discussions which increase the opportunity for active learning. We wanted to develop a comprehensive clicker methodology that creates an active lecture environment for a broad spectrum of students taking introductory physics courses. We wanted our methodology to incorporate many findings of contemporary learning science. It is recognized that learning requires active construction; students need to be actively involved in their own learning process. Learning also depends on preexisting knowledge; students construct new knowledge and understandings based on what they already know and believe. Learning is context dependent; students who have learned to apply a concept in one context may not be able to recognize and apply the same concept in a different context, even when both contexts are considered to be isomorphic by experts. On this basis, we developed question sequences, each involving the same concept but having different contexts. Answer choices are designed to address students preexisting knowledge. These sequences are used with the clickers to promote active discussions and multiple assessments. We have created, validated, and evaluated sequences sufficient in number to populate all of introductory physics courses. Our research has found that using clickers with our question sequences significantly improved student conceptual understanding. Our research has also found how to best measure student conceptual gain using research-based instruments

  15. A methodology for the evaluation of fuel rod failures under transportation accidents

    International Nuclear Information System (INIS)

    Rashid, J.Y.R.; Machiels, A.J.

    2004-01-01

    Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

  16. EMCAS: An evaluation methodology for safeguards and security systems

    International Nuclear Information System (INIS)

    Eggers, R.F.; Giese, E.W.; Bichl, F.J.

    1987-01-01

    EMCAS is an evaluation methodology for safeguards and security systems. It provides a score card of projected or actual system performance for several areas of system operation. In one area, the performance of material control and accounting and security systems, which jointly defend against the insider threat to divert or steal special nuclear material (SNM) using stealth and deceit, is evaluated. Time-dependent and time-independent risk equations are used for both diversion and theft risk calculations. In the case of loss detection by material accounting, a detailed timeliness model is provided to determine the combined effects of loss detection sensitivity and timeliness on the overall effectiveness of the material accounting detection procedure. Calculated risks take into account the capabilities of process area containment/surveillance, material accounting mass balance tests, and physical protection barriers and procedures. In addition, EMCAS evaluates the Material Control and Accounting (MC and A) System in the following areas: (1) system capability to detect errors in the official book inventory of SNM, using mass balance accounting methods, (2) system capability to prevent errors from entering the nuclear material data base during periods of operation between mass balance tests, (3) time to conduct inventories and resolve alarms, and (4) time lost from production to carry out material control and accounting loss detection activities. 3 figs., 5 tabs

  17. EMCAS, an evaluation methodology for safeguards and security systems

    International Nuclear Information System (INIS)

    Eggers, R.F.; Giese, E.W.; Bichl, F.J.

    1987-01-01

    EMCAS is an evaluation methodology for safeguards and security systems. It provides a score card of projected or actual system performance for several areas of system operation. In one area, the performance of material control and accounting and security systems, which jointly defend against the insider threat to divert or steal special nuclear material (SNM) using stealth and deceit, is evaluated. Time-dependent and time-independent risk equations are used for both diversion and theft risk calculations. In the case of loss detection by material accounting, a detailed timeliness model is provided to determine the combined effects of loss detection sensitivity and timeliness on the overall effectiveness of the material accounting detection procedure. Calculated risks take into account the capabilities of process area containment/surveillance, material accounting mass balance tests, and physical protection barriers and procedures. In addition, EMCAS evaluates the Material Control and Accounting (MCandA) System in the following areas: (1) system capability to detect errors in the official book inventory of SNM, using mass balance accounting methods, (2) system capability to prevent errors from entering the nuclear material data base during periods of operation between mass balance tests, (3) time to conduct inventories and resolve alarms, and (4) time lost from production to carry out material control and accounting loss detection activities

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  19. A Proposal for a Methodology to Develop a Cyber-Attack Penetration Test Scenario Including NPPs Safety

    Energy Technology Data Exchange (ETDEWEB)

    Lee, In Hyo [KAIST, Daejeon (Korea, Republic of); Son, Han Seong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Si Won [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of); Kang, Hyun Gook [Rensselaer Polytechnic Institute, Troy (United States)

    2016-10-15

    Penetration test is a method to evaluate the cyber security of NPPs; so, this approach was performed in some studies. Because they focused on vulnerability finding or test bed construction, scenario based approach was not performed. However, to test the cyber security of NPPs, a proper test scenario should be needed. Ahn et al. developed cyber-attack scenarios but those scenarios couldn't be applied in penetration test because they developed the scenarios based on past incidents of NPPs induced by cyber-attack. That is, those scenarios only covered scenarios which were happened before; so, they couldn't cover other various scenarios and couldn't reflect them into a penetration test. In this study, a method to develop a cyber-attack penetration test scenario of NPPs especially focused on safety point of view is suggested. To evaluate the cyber security of NPPs, penetration test can be a possible way. In this study, a method to develop a penetration test scenario was explained. Especially, the goal of hacker was focused on nuclear fuel integrity deterioration. So, in the methodology, Level 1 PSA results were utilized to reflect plant safety into the security. From the PSA results, basic event was post processed and possible cyber-attacks were reviewed with vulnerabilities of digital control system.

  20. A Proposal for a Methodology to Develop a Cyber-Attack Penetration Test Scenario Including NPPs Safety

    International Nuclear Information System (INIS)

    Lee, In Hyo; Son, Han Seong; Kim, Si Won; Kang, Hyun Gook

    2016-01-01

    Penetration test is a method to evaluate the cyber security of NPPs; so, this approach was performed in some studies. Because they focused on vulnerability finding or test bed construction, scenario based approach was not performed. However, to test the cyber security of NPPs, a proper test scenario should be needed. Ahn et al. developed cyber-attack scenarios but those scenarios couldn't be applied in penetration test because they developed the scenarios based on past incidents of NPPs induced by cyber-attack. That is, those scenarios only covered scenarios which were happened before; so, they couldn't cover other various scenarios and couldn't reflect them into a penetration test. In this study, a method to develop a cyber-attack penetration test scenario of NPPs especially focused on safety point of view is suggested. To evaluate the cyber security of NPPs, penetration test can be a possible way. In this study, a method to develop a penetration test scenario was explained. Especially, the goal of hacker was focused on nuclear fuel integrity deterioration. So, in the methodology, Level 1 PSA results were utilized to reflect plant safety into the security. From the PSA results, basic event was post processed and possible cyber-attacks were reviewed with vulnerabilities of digital control system

  1. 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?

  2. Method of safety evaluation in nuclear power plants

    International Nuclear Information System (INIS)

    Kuraszkiewicz, P.; Zahn, P.

    1988-01-01

    A novel quantitative technique for evaluating safety of subsystems of nuclear power plants based on expert estimations is presented. It includes methods of mathematical psychology recognizing the effect of subjective factors in the expert estimates and, consequently, contributes to further objectification of evaluation. It may be applied to complementing probabilistic safety assessment. As a result of such evaluations a characteristic 'safety of nuclear power plants' is obtained. (author)

  3. Economic evaluation of occupational safety preventive measures in a hospital.

    Science.gov (United States)

    Ramos, Delfina G; Arezes, Pedro M; Afonso, Paulo

    2015-01-01

    When an organization performs an integrated analysis of risks through its Occupational Health and Safety Management System, several steps are suggested to address the implications of the identified risks. Namely, the organization should make a detailed analysis of the monetary impact for the organization of each of the preventive measures considered. However, it is also important to perform an analysis of the impact of each measure on society (externalities). The aim of this paper is to present a case study related to the application of the proposed economic evaluation methodology. An analysis of the work accidents in a hospital has been made. Three of the major types of accidents have been selected: needle stings, falls and excessive strain. Following the risk assessment, some preventive measures have been designed. Subsequently, the Benefit/Cost ratio (B/C) of these measures has been calculated, both in financial terms (from the organization's perspective) and in economic terms (including the benefits for the worker and for the Society). While the financial ratio is only advantageous in some cases, when the externalities are taken into account, the B/C ratio increases significantly. It is important to consider external benefits to make decisions concerning the implementation of preventive measures in Occupational Health and Safety projects.

  4. 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

  5. APPETITE CONTROL: METHODOLOGICAL ASPECTS OF THE EVALUATION OF FOODS

    Science.gov (United States)

    Blundell, John; de Graaf, Cees; Hulshof, Toine; Jebb, Susan; Livingstone, Barbara; Lluch, Anne; Mela, David; Salah, Samir; Schuring, Ewoud; van der Knaap, Henk; Westerterp, Margriet

    2013-01-01

    This report describes a set of scientific procedures used to assess the impact of foods and food ingredients on the expression of appetite (psychological and behavioural). An overarching priority has been to enable potential evaluators of health claims about foods to identify justified claims, and to exclude claims that are not supported by scientific evidence for the effect cited. This priority follows precisely from the principles set down in the PASSCLAIM report. (4) The report allows the evaluation of the strength of health claims, about the effects of foods on appetite, which can be sustained on the basis of the commonly used scientific designs and experimental procedures. The report includes different designs for assessing effects on satiation as opposed to satiety,detailed coverage of the extent to which a change in hunger can stand-alone as a measure of appetite control, and an extensive discussion of the statistical procedures appropriate for handling data in this field of research. Since research in this area is continually evolving, new improved methodologies may emerge over time and will need to be incorporated into the framework. One main objective of the report has been to produce guidance on good practice in carrying out appetite research, and not to set down a series of commandments that must be followed. PMID:20122136

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

    International Nuclear Information System (INIS)

    Subki, Hadid M.

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    2011-01-01

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

  8. Evaluation in health: participatory methodology and involvement of municipal managers.

    Science.gov (United States)

    Almeida, Cristiane Andrea Locatelli de; Tanaka, Oswaldo Yoshimi

    2016-08-04

    To analyze scopes and limits of the use of participatory methodology of evaluation with municipal health managers and administrators. Qualitative research with health policymakers and managers of the Comissão Intergestores Regional (CIR - Regional Interagency Commission) of a health region of the state of Sao Paulo in Brazil. Representatives from seven member cities participated in seven workshops facilitated by the researchers, with the aim of assessing a specific problem of the care line, which would be used as a tracer of the system integrality. The analysis of the collected empirical material was based on the hermeneutic-dialectic methodology and aimed at the evaluation of the applied participatory methodology, according to its capacity of promoting a process of assessment capable to be used as a support for municipal management. With the participatory approach of evaluation, we were able to promote in-depth discussions with the group, especially related to the construction of integral care and to the inclusion of the user's perspective in decision-making, linked to the search for solution to concrete problems of managers. By joint exploration, the possibility of using data from electronic information systems was opened, as well as information coming directly from the users of the services, to enhance discussions and negotiations between partners. The participants were disbelievers of the replication potential of this type of evaluation without the direct monitoring of the academy, given the difficulty of organizing the process in everyday life, already taken by emergency and political issues. Evaluations of programs and services carried out within the Regional Interagency Commission, starting from the local interest and facilitating the involvement of its members by the use of participatory methodologies, can contribute to the construction of integral care. To the extent that the act of evaluating stay invested with greater significance to the local actors

  9. Evaluation methodology based on physical security assessment results: a utility theory approach

    International Nuclear Information System (INIS)

    Bennett, H.A.; Olascoaga, M.T.

    1978-03-01

    This report describes an evaluation methodology which aggregates physical security assessment results for nuclear facilities into an overall measure of adequacy. This methodology utilizes utility theory and conforms to a hierarchical structure developed by the NRC. Implementation of the methodology is illustrated by several examples. Recommendations for improvements in the evaluation process are given

  10. Nuclear safety. How is it evaluated?

    International Nuclear Information System (INIS)

    Andersson, Kjell; Andersson, Johan; Carlsson, Lennart; Olsson, Richard; Ericsson, A.M.; Gunsell, L.; Wene, C.O.

    1996-09-01

    A working group with representatives for the three subject areas reactor safety, disposal of spent fuels and transport of radioactive materials has performed a project aiming to clarify similarities and differences of the three areas concerning methods for safety analysis, criteria, risks etc; and to develop contacts between experts in the areas in order to facilitate transfer of methods. Some of the more precise objectives were: To identify common problems that could be solved jointly, to discuss prospects for a 'meta-method' that can support safety analysis in the entire field of nuclear safety, and to discuss possibilities for a homogeneous attitude towards risk management

  11. Evaluation of operating experience with safety values

    International Nuclear Information System (INIS)

    Bung, W.; Hoemke, P.; Oberender, W.; Paul, H.; Rueter, W.

    1985-01-01

    This report describes statistical investigations of 2076 functional tests carried out on power operated safety valves in conventional power plants in 1972 until 1983 with special regard to Common Mode-Failures. The results clearly show that Common Mode-Failures play an important part of non-availability for the controlled safety valves, especially in the control system. The 'Deutsche Risikostudie' does not consider any Common Mode-Failures of the primary safety valves. However there is no significant increase of the risk resulted by the primary safety valves in the 'Referenzanlage' if the calculated Common Mode-Failures probabilities are considered. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    2008-03-01

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

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

    International Nuclear Information System (INIS)

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

    2006-06-01

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

  14. Evaluation methodology and prospective introduction scenarios of FR cycle systems

    International Nuclear Information System (INIS)

    Fujii, Sumio; Katoh, Atsushi; Sato, Kazujiro

    2003-01-01

    The 21st century society will be facing growing demand of energy, global environmental issues and concerns about fossil fuel exhaustion. In this society, it is essential to seek for a sustainable energy system to solve these energy-related concerns. In order to find ways for solving these problems, 'Feasibility Study on Commercialized FR Cycle System' was launched in 1999 as a joint research project of JNC, electric utilities, JAERI, CRIEPI etc. This project aims to build promising FR cycle technologies for the 21st century. Now, we are in the second phase (JFY 2001-2005) of this project. At the end of the second phase, we will propose promising concepts through applying innovative technologies. We started this Feasibility Study with defining the development target, which ended in five items; safety, economy, environmental burden, resource utilization and nuclear non-proliferation. These items should also serve as basic viewpoints to evaluate achievements of the study. This paper describes how we evaluate FR cycle options to come up with the final promising candidates. This paper also describes a prospective scenario to introduce FR cycle system, which shows how the FR cycle will be replacing existing LWRs by using limited natural uranium resource and spent fuels. (author)

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

    International Nuclear Information System (INIS)

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

    2000-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  17. Radionuclide sorption from the safety evaluation perspective

    International Nuclear Information System (INIS)

    1992-01-01

    Research and development directed towards the assessment of the long-term performance of radioactive waste disposal systems has been recognised as a priority area with a strong need for international co-operation and co-ordination. The ultimate aims is to promote the quality and credibility of safety assessment techniques for radioactive waste disposal. Sorption in the geosphere is one of the key processes for retarding the transport of radionuclide from the underground disposal facility to the biosphere. In many cases, sorption in the near field and in the biosphere is also important. A workshop, organised to favor discussion around a small number of invited papers, was held in October 1991: - to evaluate critically the way sorption processes are incorporated in performance assessment models; - to identify open issues of high priority, and; - to propose future activities to resolve these issues. These proceedings reproduce the invited papers and the conclusions and recommendations adopted by the workshop. Eight papers are in the INIS SCOPE. The main subjects studied are: sorption database comparison, sorption database development and three case studies, experimental techniques, adsorption models

  18. Development of inspection safety evaluation technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Seok Chul; Yoon, Yeo Chang; Kim, Jong Soo; Lee, Tae Young; Kim, Chang Ryol; Lee, Hyung Sub; Kim, Jong Soo

    1995-12-01

    The purpose of this project is to protection nation inspector`s over exposure from radiation that can be occurred by inspection activity at nuclear facilities and its environment, and to ensure the safety of inspection activity at the nuclear facilities. To effectively carry out the domestic inspection task to be enforced from 1996, the evaluation for special radiation exposure rate of nuclear facilities, air and surface contamination level, and measurement and monitoring of water contamination level were made to determine whether these measured values exceeded permissible limitations, and to protect the inspector`s over exposure from radiation at domestic nuclear facilities. Management of inspector`s exposure was carried out under assistance of the Department of Health Physics. Performance tests of two gamma detectors, one neutron detector, alpha and beta detector, and gamma spectroscopy analyzer were carried out to control dose on extremity, the characteristic test for extremity dosimeter was carried out and the theoretical calculation of gamma dose conversion factors based on ANSI N13.32 standard was performed. Under the 93+2 program, IAEA began to recognize the necessity of environmental observation technology development of air-borne particulates travelled from long distance location. Associated with the necessity of this technology development, a proposal of international joint research for development of the special radiation measurement and analysis has been prepared. (author). 21 tabs., 24 figs., 20 refs.

  19. Evaluation of Safety Culture Implementation and Socialization Results

    International Nuclear Information System (INIS)

    Situmorang, Johnny

    2003-01-01

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

  20. Methodology for evaluation of environmental impact of radioactive waste storage systems

    International Nuclear Information System (INIS)

    Peres, Sueli da Silva; Rochedo, Elaine R.R.

    2005-01-01

    The Biosphere has an important role in the assessment of the long-term environmental impact of radioactive waste disposal systems. This is because the biosphere is dynamic and its evolution over time can significantly affect the dose estimates and potential environmental impacts of a repository. Future events that may occur in the biosphere, such as climate change and the human actions, are the main sources of uncertainty in the modeling of the biosphere, and consequently, in anticipation of the scenarios of human exposure to radiation. In this context, the use of an alternative methodology more detailed and systematic for the development of conceptual models and prediction of uncertainty has been shown to be a useful tool to improve the quality of the evaluation. This methodology indicates the components and phenomena inherent to waste, design and location of the storage installation that need to be identified during the development of the conceptual model and the selection of the computer code to be used to represent the model. This methodology has been applied in assessing the long-term safety of radioactive waste storage systems. This paper presents the advantages of using this approach in the development of conceptual models and in the treatment of uncertainties

  1. 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.

  2. 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.

  3. 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

  4. THE STUDY OF ORGANIZATIONAL CULTURE: METHODOLOGY FOR QUANTITATIVE EVALUATION AND ANALYSIS

    Directory of Open Access Journals (Sweden)

    To Thu Trang

    2014-01-01

    Full Text Available This article discusses the concept: methodof evaluation organizational culture, qualitative and quantitative assessment methodology and lists the basic methodologyfor assessing organizational culture. Fullydescribe professor Denison’s methodology for assessing organizational culture.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  7. 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)

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

    International Nuclear Information System (INIS)

    Hashim Muhammad; Yoshikawa, Hidekazu; Yang Ming

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  10. 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)

  11. 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

  12. Review of effectiveness-evaluation methodologies for safeguards and security systems

    International Nuclear Information System (INIS)

    Dowdy, E.J.; Mangan, D.L.

    1982-06-01

    We discuss the factors that influence the effectiveness of safeguards and security measures and the characteristics required of effectiveness evaluation methodologies. Within this context and from a utility standpoint, we review those effectiveness evaluation methodologies that have been developed. Our principal recommendation concerns the application and concomitant validation of existing methodologies. This recommendation derives from our conclusion that there has been a gross imbalance between the effort spent on methodology development and the application of those methodologies. Only for those safeguards measures that do not seem to be covered by existing methodologies or that seem to be inadequately covered do we suggest development. 44 references

  13. Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

    International Nuclear Information System (INIS)

    Wetzel, N.; Scharfe, A.

    1998-01-01

    The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

  14. Russian Сonsumer Sector: Methodology of Evaluation

    Directory of Open Access Journals (Sweden)

    Gavriil Aleksandrovich Agarkov

    2015-12-01

    Full Text Available The article describes the methodology for estimating the Russia’s consumer sector and the effect of its application. The monitoring procedure of the Russian consumer sector groups indicators into two units: the unit of the estimation of consumer goods and the services market estimation unit. The estimation unit of consumer goods is composed of two modules: food products and non-food products. This module offers two components that provide an estimation of the consumer sector: marketing (estimates the accessibility of retail trade and services for end users and production (estimates the domestic manufacture. The results of the estimation show general improvements in the consumer sector in the period of 2000–2014, but overall development is evaluated as low. The analysis revealed that the financing is growing faster than the quality indices of development. As an example, the financing of agriculture has increased by 1.5 times over 15 years (against comparable prices from 2000, while agricultural production has not changed. Another most pressing challenge is the weak differentiation of the Russian economy, as evidenced by the low rates of non-food production (availability of non-foods of own production remains at a low level and averages 20 %. The results of the estimation suggest the need to reform the regulation of the sector primarily concerning priorities for its development and improvement of financial and economic mechanisms to achieve them.

  15. Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  16. A dynamic systems engineering methodology research study. Phase 2: Evaluating methodologies, tools, and techniques for applicability to NASA's systems projects

    Science.gov (United States)

    Paul, Arthur S.; Gill, Tepper L.; Maclin, Arlene P.

    1989-01-01

    A study of NASA's Systems Management Policy (SMP) concluded that the primary methodology being used by the Mission Operations and Data Systems Directorate and its subordinate, the Networks Division, is very effective. Still some unmet needs were identified. This study involved evaluating methodologies, tools, and techniques with the potential for resolving the previously identified deficiencies. Six preselected methodologies being used by other organizations with similar development problems were studied. The study revealed a wide range of significant differences in structure. Each system had some strengths but none will satisfy all of the needs of the Networks Division. Areas for improvement of the methodology being used by the Networks Division are listed with recommendations for specific action.

  17. Risk allocation approach to reactor safety design and evaluation

    International Nuclear Information System (INIS)

    Gokcek, O.; Temme, M.I.; Derby, S.L.

    1978-01-01

    This paper describes a risk allocation technique used for determining nuclear power plant design reliability requirements. The concept of risk allocation-optimum choice of safety function reliabilities under a maximum risk constraint - is described. An example of risk allocation is presented to demonstrate the application of the methodology

  18. Guide for understanding and evaluation of safety culture

    International Nuclear Information System (INIS)

    2008-01-01

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

  19. Safety Evaluation of Elderly Laparoscopic Cholecystectomy

    Directory of Open Access Journals (Sweden)

    Bijan Khorasani

    2008-10-01

    Full Text Available Objectives: The most common indication for abdominal surgery in elderly (65 & older is gallstone disease.The elderly are more prone to complication of such surgery due to their other co-morbidity and thus they may benefit mostly from a safer method of surgical procedure. The purpose of this study was to evaluate the safety and outcome of laparoscopic cholecystectomy in elderly compare to the conventional method. Methods & Materials: Via prospective study from June 2005 to March 2008 included all patient older than 60 years of age who underwent cholecystectomy by open (Group A and laparoscopic (Group B method in Milad Hospital.The demographic data (sex age, American Society of Anesthesiologists’ (ASA score, postoperative mortality morbidity, length of stay and operative time were recorded for each patient and were compared between two methods. Data of patient’s analysis by SPSS with chi-square and t test. Results: The study included 311 elderlies (34% men and 66% women. Hundred fifty-seven patients underwent open cholecystectomy; 154 patients underwent laparoscopic cholecystectomy. The mean age was 7141 years. The outcome in group B (laparoscopic included: morality 0%, morbidity 2%, postoperative hospital stay 1 days, mean operation time was 40 minutes. In group A(open: mortality and morbidity rate were 21% and 12% respectively with postoperative hospital stay 331 days and similar operation time as group A. Conclusion: Laparoscopic cholecystectomy is the gold standard treatment and safe procedure in elderly patient and aging is not considered to be a contraindication laparoscopic surgery in such patients.

  20. PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

  1. Influenza vaccines: Evaluation of the safety profile

    Science.gov (United States)

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

    2018-01-01

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

  2. Methodology for the application of the probabilistic safety analysis to the cobalto therapy units in Cuba

    International Nuclear Information System (INIS)

    Vilaragut Llanes, Juan Jose; Ferro Fernandez, Ruben; Troncoso Fleitas, Mayra; Lozano Lima, Berta; De la Fuente Puch, Andres; Perez Reyes, Yolanda; Dumenigo Gonzalez, Cruz

    2001-01-01

    Presently work the main elements are discussed kept in mind for the use of the Analyses Probabilistas of Security in the evaluation of the security of the units of cobalto therapy of Cuba and it is presented, like part of the results of the first stage of the Study, the Methodological Guide that is being used in a Contract of Investigation of the OIEA that at the moment carries out the community of authors of the CNSN, of group with other specialists of the Ministry of Public Health (MINSAP)

  3. The impact of WASH-1400 on reactor safety evaluation

    International Nuclear Information System (INIS)

    Tanguy, P.Y.

    1976-01-01

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

  4. Systematic safety evaluation of old nuclear power plants

    International Nuclear Information System (INIS)

    Dredemis, G.; Fourest, B.

    1984-01-01

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

  5. An integrated methodology to evaluate a spent nuclear fuel storage system

    International Nuclear Information System (INIS)

    Yoon, Jeong Hyoun

    2008-02-01

    This study introduced a methodology that can be applied for development of a dry storage system for spent nuclear fuels. It consisted of several design activities that includes development of a simplified program to analyze the amount of spent nuclear fuels from reflecting the practical situation in spent nuclear fuel management and a simplified program to evaluate the cost of 4 types of representing storage system to choose the most competitive option considering economic factor. As verification of the implementation of the reference module to practical purpose, a simplified thermal analysis code was suggested that can see fulfillment of limitation of temperature in long term storage and oxidation analysis. From the thermal related results, the reference module can accommodate full range of PHWR spent nuclear fuels and significant portion of PWR ones too. From the results, the reference storage system can be concluded that has fulfilled the important requirements in terms of long term integrity and radiological safety. Also for the purpose of solving scattered radiation along with deep penetration problems in cooling storage system, small but efficient design alternation was suggested together with its efficiency that can reduce scattered radiation by 1/3 from the original design. Along with the countermeasure for the shielding problem, in consideration of PWR spent nuclear fuels, simplified criticality analysis methodology retaining conservativeness was proposed. The results show the reference module is efficient low enrichment PWR spent nuclear fuel and even relatively high enrichment fuels too if burnup credit is taken. As conclusive remark, the methodology is simple but efficient to plan a concept design of convective cooling type of spent nuclear fuels storage. It can be also concluded that the methodology derived in this study and the reference module has feasibility in practical implementation to mitigate the current complex situation in spent fuel

  6. 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

  7. 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

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

    International Nuclear Information System (INIS)

    Perez, S.

    2002-01-01

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

  9. 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

  10. Code coverage measurement methodology for MMI software of safety-class I and C system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eun Hyung; Jung, Beom Young; Choi, Seok Joo [Suresofttech, Seoul (Korea, Republic of)

    2016-10-15

    MMI (Man-Machine Interface) software of the safety instrumentation and control system used in nuclear power plants carry out an important functions, such as displaying and transmitting the commend to another system, and change setpoints the safety-related information. Yet, this has been recognized reliability of the MMI software plays an important role in enhancing nuclear power plants are operating, regulatory standards have been strengthened with it. Strengthening of regulatory standards has affected even perform software testing soon, and accordingly, the current regulatory require the measurement of code coverage with legal standard. In this paper, it poses a problem of the conventional method used for measuring the above-mentioned code coverage, presents a new coverage measuring method for solving the exposed problems. In this paper, we checked the problems such as limit and the low efficiency of the existing test coverage measuring method on the MMI software using in nuclear power instrumentation and control systems, and it proposed a new test coverage measuring method as a solution for this. If you apply a new method of Top-Down approach, can mitigate all of the problems of existing test coverage measurement methods and possible coverage achievement of the desired objectives. Of course, it is still necessary to secure more cases, and the methodology should be systematization based on the cases. Thus, if later the efficient and reliable are ensured through the application in many cases, as well as nuclear power instrumentation and control, may be used to ensure code coverage of software of the many areas where the GUI is utilized.

  11. Deepening om Safety culture Auto evaluation

    International Nuclear Information System (INIS)

    Lopez Churruca, I.; Buedo Jimenez, J. L.

    2009-01-01

    The concept of safety culture used in nuclear forums refers to the series of actions aimed at guaranteeing that safety issues in nuclear power plants are adequately addressed. The activities to which this concept refers have been gradually extended over the years so that, today, they encompass the whole organizations structure. In other words, the safety culture implies that all positions in the organizational structure perform their tasks with a level of attention such that all their senses are focused on them. And this performance is what leads us to excellence in plant operation. In addition, the implementation of a self-assessment system in the Cofrentes Nuclear Power Plant of its activities and processes has resulted in the identification of expectations on the basis of which we can identify strengths and weaknesses, enabling us to leverage the former and correct the latter. (Author)

  12. Evaluation of safety management in an Appliances manufacturing company

    Directory of Open Access Journals (Sweden)

    F. Golbabaei

    2015-01-01

    Full Text Available Introduction: Prevention of accidents and work related diseases, are not allowed regardless of the safety of employees, customers, contractors and other persons. Assessment of individual safety management activities could reduce many losses. Present study aimed to evaluate the safety management of a household appliance manufacturing company.  .Material and Method: This study has done in a household appliance manufacturing company in Damavand city. Two questionnaires were firstly designed based on the weighted scores. The questionnaire 1 consisted of 4 indicators: Safety of machinery, Electrical safety, Risk assessment and Fire safety. Questionnaire 2 consisted of 11 sub indicators. Both questionnaires were completed by 30 HSE experts and supervisors. Reliability of questionnaires was based on cronbachs alpha coefficient. the safety status of each unit was determined and scored using information acquired by the questionnaires. Lastly, the safety of the entire company was determined.  .Result: Results showed that in safety management: the pressing and store house were in a good range of 66.66 and 60.12 points. Powder painting, enameling, laboratory were in a average range of 56.25, 55.92 and 54.15 points. Assembling and door storage were in a week range of 46.06 points.  .Conclusion: The findings showed that the safety status in the studied appliances company is in average range with 55.45 points. Therefore, it is recommended that the safety indicators should be improved for the betterment of the safety management in the company.

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

    NARCIS (Netherlands)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Vladescu, G.; Banciu, O.

    1999-01-01

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

  15. Application of the methodology of safety probabilistic analysis to the modelling the emergency feedwater system of Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Troncoso, M.; Oliva, G.

    1993-01-01

    The application of the methodology developed in the framework of the national plan of safety probabilistic analysis (APS) to the emergency feed water system for the failures of small LOCAS and external electrical supply loss in the nuclear power plant is illustrated in this work. The facilities created by the ARCON code to model the systems and its documentation are also expounded

  16. Criticality safety benchmark evaluation project: Recovering the past

    Energy Technology Data Exchange (ETDEWEB)

    Trumble, E.F.

    1997-06-01

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

  17. Philosophy of safety evaluation on fast breeder reactor

    International Nuclear Information System (INIS)

    1981-01-01

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

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

    Science.gov (United States)

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

    2009-01-01

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

  19. A paradigm shift in organisational safety culture evaluation and training

    OpenAIRE

    Cram, Robert; Sime, Julie-Ann

    2015-01-01

    The focus of this research is to explore the issues surrounding traditional approaches towards understanding the safety culture of an organisation operating in a high risk environment and to identify an effective technique to educate corporate management in how to measure and evaluate the underlying safety culture of their own organisations. The results of the first part of the research highlight the concerns being expressed by both academic and industrial communities that current safety cult...

  20. 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)

  1. Outcomes from the regional Co-operation in the Area of the Safety Analysis Methodology

    International Nuclear Information System (INIS)

    D'Auria, F.; Mavko, B.; Prosek, A.; Debrecin, N.; Bajs, T.

    2000-01-01

    International Atomic Energy Agency (IAEA) carried out the Co-ordinated Research Program (CRP) ON V alidation of Accident and Safety Analysis Methodology'' in the period between 1995 and 1998. Three areas of interest identified by the participants referred to the pressurised water reactors of Western and Eastern type (PWR and WWER type). The specific areas of attention were: system behaviour of the primary and secondary loops (PS area), the containment response (CO area) and the severe accidents (SA area). During the CRP it became clear that the technology advancements, the available tools (i.e. codes) and the experimental databases in the above areas are quite different. At the conclusion of the CRP, all objectives of the program have been reached. This paper presents the summary of the regional co-operation in this framework. The CRP activities focused on the codes and expertise available at the participating organisations. This overview therefore summarises their experience related to the state-of-the-art in the field of computational accident analysis. In addition, the paper proposes the recommendations for future activities related to the code usage, the user effects and code development. In pursuing of these goals special attention is given to the importance of the international co-operation. (author)

  2. Safety-specific benefit of the probabilistic evaluation of older nuclear power plants

    International Nuclear Information System (INIS)

    Hoertner, H.; Koeberlein, K.

    1991-01-01

    The report summarizes the experience of the GRS obtained within the framework of a probabilistic evaluation of older nuclear power plants and the German risk study. The applied methodology and the problems involved are explained first. After a brief summary of probabilistic analyses carried out for German nuclear power plants, reliability analyses for older systems are discussed in detail. The findings from the probabilistic safety analyses and the conclusions drawn are presented. (orig.) [de

  3. The Making of discussion groups in a combined process of internal evaluation of safety culture

    International Nuclear Information System (INIS)

    German, S.; Buedo, J. L.; La Salabarnada, E.; Navajas, J.; Silla, I.

    2012-01-01

    The purpose of this paper is to show the design and evaluation of safety culture conducted in the Cofrentes nuclear plant. The process has combined the use of different methodologies and techniques and has allowed the participation of different internal and external stake holders. For internal assessment discussion groups were conducted. These groups, which were designed and analyzed by the CIEMAT, were led by employees from different levels of Cofrentes.

  4. 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

  5. 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

  6. Methodology for economic evaluation of software development projects

    International Nuclear Information System (INIS)

    Witte, D.M.

    1990-01-01

    Many oil and gas exploration and production companies develop computer software in-house or with contract programmers to support their exploration activities. Software development projects compete for funding with exploration and development projects, though most companies lack valid comparison measures for the two types of projects. This paper presents a methodology of pro form a cash flow analysis for software development proposals intended for internal use. This methodology, based on estimates of development and support costs, exploration benefits, and probability of successful development and implementation, can be used to compare proposed software development projects directly with competing exploration proposals

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

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

    2013-01-01

    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

  8. Criticality Safety Evaluation of Hanford Tank Farms Facility

    Energy Technology Data Exchange (ETDEWEB)

    WEISS, E.V.

    2000-12-15

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

  9. Criticality Safety Evaluation of Hanford Tank Farms Facility

    International Nuclear Information System (INIS)

    WEISS, E.V.

    2000-01-01

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

  10. A reliability evaluation method for NPP safety DCS application software

    International Nuclear Information System (INIS)

    Li Yunjian; Zhang Lei; Liu Yuan

    2014-01-01

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

  11. Study on scenario evaluation methodology for decommissioning nuclear facilities using fuzzy logic

    International Nuclear Information System (INIS)

    Matsuhashi, Kazuya; Yanagihara, Satoshi

    2015-01-01

    Since there are many scenarios of the process from start to completion of a decommissioning project, it is important to study scenarios of decommissioning by evaluating such properties as safety, cost, and technology. An optimum scenario with the highest feasibility in accordance with the facility and environmental conditions should be selected on the basis of the results of the study. For analyzing a scenario of decommissioning, we prepared structured work packages by using the work breakdown structures (WBS) method together with qualitative evaluation of the technologies being applied to work packages located at the bottom (the third level) of the WBS. A calculation model was constructed to evaluate the feasibility of a scenario where fuzzy logic is applied to derive a score of technology performance and TOPSIS is applied for getting a feasibility grade of the scenario from technical performance scoring. As a case study, the model was applied to the debris removal scenario of Fukushima Daiichi Nuclear Power Plant to confirm its applicability. Two scenarios, underwater and in-air debris removal cases, were characterized by extracting the work packages with the lowest feasibility and by obtaining total average scores of the scenarios. It is confirmed that the methodology developed is useful for the scenario evaluation of decommissioning nuclear facilities. (author)

  12. Research on the Evaluation System for Rural Public Safety Planning

    Institute of Scientific and Technical Information of China (English)

    Ming; SUN; Jianxin; YAN

    2014-01-01

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

  13. Integrated methodology for the evaluation of the vulnerability of archaeological sites: the Roman Theater of Verona

    International Nuclear Information System (INIS)

    Cescatti, Elvis; Lorenzoni, Filippo; Caldon, Mauro; Modena, Claudio; Da Porto, Francesca

    2016-01-01

    The knowledge-based methodologies for assessing the structural safety and vulnerability of historical buildings are based on the application of an operational approach that is divided into several phases including historical research, the execution of non-destructive or moderately destructive investigations, monitoring and structural analysis. Monitoring is a useful tool to be used throughout the whole process of knowledge not only in the preliminary diagnosis, but also for the control of the quality and effectiveness of the interventions, in the context of appropriate conservation programs e protection of the object being studied. From this point of view, monitoring becomes essential for: (i) the evaluation of the real structural behavior and the identification of vulnerabilities, prior to the execution of any intervention; (ii) minimization and optimization of interventions, firstly by providing indications regarding those that are not necessary and / or invasive. This methodology has been recently applied and valid from the authors to an archaeological site of particular relevance: the Roman Theater of Verona. The deepening of the knowledge phase (historical analysis, survey, study of the subsoil, properties of materials), combined with the installation of a permanent monitoring system and the structural modeling, allowed to obtain a reliable assessment of the site's vulnerabilities with the aim of guaranteeing in an effective and careful protection and enhancement.

  14. Evaluation of safety practices and performance in a brewery industry ...

    African Journals Online (AJOL)

    Evaluation of safety practices and performance in a brewery industry in Nigeria between 2000 – 2007. ... Journal of Applied Sciences and Environmental Management ... The study revealed that a total of 156 accidents were prevented in the period of the safety programme which translates to an average of 19.45 per year.

  15. Second Meeting for Evaluation of the Nuclear Safety Convention

    International Nuclear Information System (INIS)

    2002-01-01

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

  16. Contribution of maintainability and maintenance to problems of safety evaluation

    International Nuclear Information System (INIS)

    Adnot, Serge; Meriaux, Pierre.

    1977-10-01

    A method has been developed for defining the contribution of Maintainability and the Maintenance Studies to Safety evaluation problems. The efficiency of this method is shown and results obtained are given for two theoretical examples approximating reality. For repairable systems, the risk defined according to such given safety criterion, becomes a characteristic of the systems in operation [fr

  17. Destination brands and website evaluation: a research methodology

    Directory of Open Access Journals (Sweden)

    J Fernández-Cavia

    2013-10-01

    Full Text Available Introduction:The World Wide Web has become the primary instrument used by tourists in order to search for information. As a result, tourism websites pertaining to destinations need to be appealing and must convey their brand image in an appropriate, effective manner. However, there is no methodology in place to assess the quality and communicative effectiveness of destination websites that is scientifically sound and universally accepted. The development of such a methodology is one of the tasks we have proposed within the framework of the research project: “New strategies for advertising and promoting Spanish tourism brands online” (CSO2008-02627, funded by the Spanish Ministry of Science and Innovation. Method: The project team have developed an interdisciplinary, all-embracing analysis template combining certain automated analyses with other qualitative and quantitative ones. The template comprises a total of 12 subject areas and 154 indicators prepared on the basis of contributions from prominent experts in each of the fields of work. This article sets out the analysis methodology drawn up and possible applications are given. Results: The primary aim of the project is to provide an assessment methodology that would make it possible to optimise destination brand websites, thus providing a tool to support the work of public tourism destination managers.

  18. Methodological Considerations in Designing and Evaluating Animal-Assisted Interventions.

    Science.gov (United States)

    Stern, Cindy; Chur-Hansen, Anna

    2013-02-27

    This paper presents a discussion of the literature on animal-assisted interventions and describes limitations surrounding current methodological quality. Benefits to human physical, psychological and social health cannot be empirically confirmed due to the methodological limitations of the existing body of research, and comparisons cannot validly be made across different studies. Without a solid research base animal-assisted interventions will not receive recognition and acceptance as a credible alternative health care treatment. The paper draws on the work of four systematic reviews conducted over April-May 2009, with no date restrictions, focusing exclusively on the use of canine-assisted interventions for older people residing in long-term care. The reviews revealed a lack of good quality studies. Although the literature base has grown in volume since its inception, it predominantly consists of anecdotal accounts and reports. Experimental studies undertaken are often flawed in aspects of design, conduct and reporting. There are few qualitative studies available leading to the inability to draw definitive conclusions. It is clear that due to the complexities associated with these interventions not all weaknesses can be eliminated. However, there are basic methodological weaknesses that can be addressed in future studies in the area. Checklists for quantitative and qualitative research designs to guide future research are offered to help address methodological rigour.

  19. Structural evaluation of safety class components to natural phenomena loadings

    International Nuclear Information System (INIS)

    Conrads, T.J.

    1989-01-01

    This paper addresses the efforts completed at the US Department of Energy Hanford Site near Richland, Washington, to qualify structurally a number of existing safety class components in the Plutonium Finishing Plant complex. Design, fabrication, and installation of the facility occurred in the 1950s and 1960s and were based on the Uniform Building Code criteria for wind and earthquake loads. Recently the buildings were qualified to site-specific wind and seismic hazards. The methodology employed to qualify seismically the safety class components is discussed

  20. 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

  1. Evaluation of reliability assurance approaches to operational nuclear safety

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  2. 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.

  3. Criteria for the Research Institute for Fragrance Materials, Inc. (RIFM) safety evaluation process for fragrance ingredients.

    Science.gov (United States)

    Api, A M; Belsito, D; Bruze, M; Cadby, P; Calow, P; Dagli, M L; Dekant, W; Ellis, G; Fryer, A D; Fukayama, M; Griem, P; Hickey, C; Kromidas, L; Lalko, J F; Liebler, D C; Miyachi, Y; Politano, V T; Renskers, K; Ritacco, G; Salvito, D; Schultz, T W; Sipes, I G; Smith, B; Vitale, D; Wilcox, D K

    2015-08-01

    The Research Institute for Fragrance Materials, Inc. (RIFM) has been engaged in the generation and evaluation of safety data for fragrance materials since its inception over 45 years ago. Over time, RIFM's approach to gathering data, estimating exposure and assessing safety has evolved as the tools for risk assessment evolved. This publication is designed to update the RIFM safety assessment process, which follows a series of decision trees, reflecting advances in approaches in risk assessment and new and classical toxicological methodologies employed by RIFM over the past ten years. These changes include incorporating 1) new scientific information including a framework for choosing structural analogs, 2) consideration of the Threshold of Toxicological Concern (TTC), 3) the Quantitative Risk Assessment (QRA) for dermal sensitization, 4) the respiratory route of exposure, 5) aggregate exposure assessment methodology, 6) the latest methodology and approaches to risk assessments, 7) the latest alternatives to animal testing methodology and 8) environmental risk assessment. The assessment begins with a thorough analysis of existing data followed by in silico analysis, identification of 'read across' analogs, generation of additional data through in vitro testing as well as consideration of the TTC approach. If necessary, risk management may be considered. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. Risk-informed analysis of the large break loss of coolant accident and PCT margin evaluation with the RISMC methodology

    International Nuclear Information System (INIS)

    Liang, T.H.; Liang, K.S.; Cheng, C.K.; Pei, B.S.; Patelli, E.

    2016-01-01

    Highlights: • With RISMC methodology, both aleatory and epistemic uncertainties have been considered. • 14 probabilistically significant sequences have been identified and quantified. • A load spectrum for LBLOCA has been conducted with CPCT and SP of each dominant sequence. • Comparing to deterministic methodologies, the risk-informed PCT margin can be greater by 44–62 K. • The SP of the referred sequence to cover 99% in the load spectrum is only 5.07 * 10 −3 . • The occurrence probability of the deterministic licensing sequence is 5.46 * 10 −5 . - Abstract: For general design basis accidents, such as SBLOCA and LBLOCA, the traditional deterministic safety analysis methodologies are always applied to analyze events based on a so called surrogate or licensing sequence, without considering how low this sequence occurrence probability is. In the to-be-issued 10 CFR 50.46a, the LBLOCA will be categorized as accidents beyond design basis and the PCT margin shall be evaluated in a risk-informed manner. According to the risk-informed safety margin characterization (RISMC) methodology, a process has been suggested to evaluate the risk-informed PCT margin. Following the RISMC methodology, a load spectrum of PCT for LBLOCA has been generated for the Taiwan’s Maanshan Nuclear Power plant and 14 probabilistic significant sequences have been identified. It was observed in the load spectrum that the conditional PCT generally ascends with the descending sequence occurrence probability. With the load spectrum covering both aleatory and epistemic uncertainties, the risk-informed PCT margin can be evaluated by either expecting value estimation method or sequence probability coverage method. It was found that by comparing with the traditional deterministic methodology, the PCT margin evaluated by the RISMC methodology can be greater by 44–62 K. Besides, to have a cumulated occurrence probability over 99% in the load spectrum, the occurrence probability of the

  5. Risk-informed analysis of the large break loss of coolant accident and PCT margin evaluation with the RISMC methodology

    Energy Technology Data Exchange (ETDEWEB)

    Liang, T.H. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Liang, K.S., E-mail: ksliang@alum.mit.edu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Cheng, C.K.; Pei, B.S. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Patelli, E. [Institute of Risk and Uncertainty, University of Liverpool, Room 610, Brodie Tower, L69 3GQ (United Kingdom)

    2016-11-15

    Highlights: • With RISMC methodology, both aleatory and epistemic uncertainties have been considered. • 14 probabilistically significant sequences have been identified and quantified. • A load spectrum for LBLOCA has been conducted with CPCT and SP of each dominant sequence. • Comparing to deterministic methodologies, the risk-informed PCT margin can be greater by 44–62 K. • The SP of the referred sequence to cover 99% in the load spectrum is only 5.07 * 10{sup −3}. • The occurrence probability of the deterministic licensing sequence is 5.46 * 10{sup −5}. - Abstract: For general design basis accidents, such as SBLOCA and LBLOCA, the traditional deterministic safety analysis methodologies are always applied to analyze events based on a so called surrogate or licensing sequence, without considering how low this sequence occurrence probability is. In the to-be-issued 10 CFR 50.46a, the LBLOCA will be categorized as accidents beyond design basis and the PCT margin shall be evaluated in a risk-informed manner. According to the risk-informed safety margin characterization (RISMC) methodology, a process has been suggested to evaluate the risk-informed PCT margin. Following the RISMC methodology, a load spectrum of PCT for LBLOCA has been generated for the Taiwan’s Maanshan Nuclear Power plant and 14 probabilistic significant sequences have been identified. It was observed in the load spectrum that the conditional PCT generally ascends with the descending sequence occurrence probability. With the load spectrum covering both aleatory and epistemic uncertainties, the risk-informed PCT margin can be evaluated by either expecting value estimation method or sequence probability coverage method. It was found that by comparing with the traditional deterministic methodology, the PCT margin evaluated by the RISMC methodology can be greater by 44–62 K. Besides, to have a cumulated occurrence probability over 99% in the load spectrum, the occurrence probability

  6. An evaluation of the uranium mine radiation safety course

    International Nuclear Information System (INIS)

    1984-07-01

    The report evaluates the Uranium Mine Radiation Safety Course focussing on the following areas: effectivenss of the course; course content; instructional quality; course administration. It notes strengths and weaknesses in these areas and offers preliminary recommendations for future action

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

    International Nuclear Information System (INIS)

    TOFFER, H.

    2006-01-01

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

  8. Nuclear criticality safety parameter evaluation for uranium metallic alloy

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  9. Application of a structural model for advanced analysis in the evaluation of nuclear safety

    International Nuclear Information System (INIS)

    Landesmann, Alexandre; Barros, Francisco Claudio Pereira de; Batista, Eduardo de Miranda

    2003-01-01

    The Advanced Analysis concept, which means the direct consideration of both physical and geometric nonlinear effects in the analysis and design of steel buildings structures, represents the state-of-art in the field of structural analysis by this beginning of the 21 st century. In this context, the present paper presents an Advanced Analysis methodology applied to the Safety Evaluation of high hazardous civil structures. This Safety Evaluation plays an important part in the regulators position as a step in the licensing process performed by CNEN - Brazilian Nuclear Energy Commission. The proposed Advance Analysis procedure is implemented by a refined second-order plastic hinge model. The application of this model allows to carry out: the description of the inelastic structural behavior; the identification of the collapse mechanism; the ultimate load level; structural safety's level and the service ability limit. (author)

  10. Implementation of Recommendations from the One System Comparative Evaluation of the Hanford Tank Farms and Waste Treatment Plant Safety Bases

    International Nuclear Information System (INIS)

    Garrett, Richard L.; Niemi, Belinda J.; Paik, Ingle K.; Buczek, Jeffrey A.; Lietzow, J.; McCoy, F.; Beranek, F.; Gupta, M.

    2013-01-01

    A Comparative Evaluation was conducted for One System Integrated Project Team to compare the safety bases for the Hanford Waste Treatment and Immobilization Plant Project (WTP) and Tank Operations Contract (TOC) (i.e., Tank Farms) by an Expert Review Team. The evaluation had an overarching purpose to facilitate effective integration between WTP and TOC safety bases. It was to provide One System management with an objective evaluation of identified differences in safety basis process requirements, guidance, direction, procedures, and products (including safety controls, key safety basis inputs and assumptions, and consequence calculation methodologies) between WTP and TOC. The evaluation identified 25 recommendations (Opportunities for Integration). The resolution of these recommendations resulted in 16 implementation plans. The completion of these implementation plans will help ensure consistent safety bases for WTP and TOC along with consistent safety basis processes. procedures, and analyses. and should increase the likelihood of a successful startup of the WTP. This early integration will result in long-term cost savings and significant operational improvements. In addition, the implementation plans lead to the development of eight new safety analysis methodologies that can be used at other U.S. Department of Energy (US DOE) complex sites where URS Corporation is involved

  11. 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)

  12. Economic evaluation studies in reproductive medicine: a systematic review of methodologic quality

    NARCIS (Netherlands)

    Moolenaar, Lobke M.; Vijgen, Sylvia M. C.; Hompes, Peter; van der Veen, Fulco; Mol, Ben Willem J.; Opmeer, Brent C.

    2013-01-01

    To evaluate the methodologic quality of economic analyses published in the field of reproductive medicine. Systematic review. Centers for reproductive care. Infertility patients. We performed a Medline search to identify economic evaluation studies in reproductive medicine. We included studies that

  13. 77 FR 26292 - Risk Evaluation and Mitigation Strategy Assessments: Social Science Methodologies to Assess Goals...

    Science.gov (United States)

    2012-05-03

    ...] Risk Evaluation and Mitigation Strategy Assessments: Social Science Methodologies to Assess Goals... announcing a public workshop entitled ``Risk Evaluation and Mitigation Strategy Assessments: Social Science... constructive dialogue and information-sharing among regulators, researchers, the pharmaceutical industry...

  14. A tool for safety evaluations of road improvements.

    Science.gov (United States)

    Peltola, Harri; Rajamäki, Riikka; Luoma, Juha

    2013-11-01

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

  15. Packaging Evaluation Approach to Improve Cosmetic Product Safety

    OpenAIRE

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

    2016-01-01

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

  16. 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

  17. Seismic re-evaluation of Kozloduy NPP criteria, methodology, implementation

    International Nuclear Information System (INIS)

    Kostov, M.

    2003-01-01

    The paper describes some features of the methodology applied for seismic upgrading of civil structures at the site of the Kozloduy NPP. The essence of the methodology is the use of as-build data, realistic damping and inelastic reduction factors. As an example of seismic upgrading the analyses of units 3 and 4 are presented. The analyses are showing that for effective seismic upgrading detailed investigations are needed in order to understand the significant response modes of the structures. In the presented case this is the rotation of the attached flexible structures to the stiff reactor building. Based on this an upgrading approach is applied to increase the seismic resistance for the predominant motion. The second significant approach applied is the strengthening of the prefabricated element joints. Although it is very simple it allows use of the available element capacity. (author)

  18. Evaluating a federated medical search engine: tailoring the methodology and reporting the evaluation outcomes.

    Science.gov (United States)

    Saparova, D; Belden, J; Williams, J; Richardson, B; Schuster, K

    2014-01-01

    Federated medical search engines are health information systems that provide a single access point to different types of information. Their efficiency as clinical decision support tools has been demonstrated through numerous evaluations. Despite their rigor, very few of these studies report holistic evaluations of medical search engines and even fewer base their evaluations on existing evaluation frameworks. To evaluate a federated medical search engine, MedSocket, for its potential net benefits in an established clinical setting. This study applied the Human, Organization, and Technology (HOT-fit) evaluation framework in order to evaluate MedSocket. The hierarchical structure of the HOT-factors allowed for identification of a combination of efficiency metrics. Human fit was evaluated through user satisfaction and patterns of system use; technology fit was evaluated through the measurements of time-on-task and the accuracy of the found answers; and organization fit was evaluated from the perspective of system fit to the existing organizational structure. Evaluations produced mixed results and suggested several opportunities for system improvement. On average, participants were satisfied with MedSocket searches and confident in the accuracy of retrieved answers. However, MedSocket did not meet participants' expectations in terms of download speed, access to information, and relevance of the search results. These mixed results made it necessary to conclude that in the case of MedSocket, technology fit had a significant influence on the human and organization fit. Hence, improving technological capabilities of the system is critical before its net benefits can become noticeable. The HOT-fit evaluation framework was instrumental in tailoring the methodology for conducting a comprehensive evaluation of the search engine. Such multidimensional evaluation of the search engine resulted in recommendations for system improvement.

  19. CONDOS methodology for evaluation of radiation exposure from consumer products

    International Nuclear Information System (INIS)

    O'Donnell, F.R.

    1979-01-01

    The CONDOS methodology is a tool for estimating radiation doses to man from exposures to radionuclides incorporated in consumer products. It consists of two parts: (1) an outline, checklist, and selected data for modeling the life span of a product or the material from which it is made; and (2) a computer code that uses the life-span model to calculate radiation doses to exposed individuals and population groups

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, D.; Larue, J.; Fischer- Appelt, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koln (Germany)

    2006-07-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  3. [Implementation of "5S" methodology in laboratory safety and its effect on employee satisfaction].

    Science.gov (United States)

    Dogan, Yavuz; Ozkutuk, Aydan; Dogan, Ozlem

    2014-04-01

    Health institutions use the accreditation process to achieve improvement across the organization and management of the health care system. An ISO 15189 quality and efficiency standard is the recommended standard for medical laboratories qualification. The "safety and accommodation conditions" of this standard covers the requirement to improve working conditions and maintain the necessary safety precautions. The most inevitable precaution for ensuring a safe environment is the creation of a clean and orderly environment to maintain a potentially safe surroundings. In this context, the 5S application which is a superior improvement tool that has been used by the industry, includes some advantages such as encouraging employees to participate in and to help increase the productivity. The main target of this study was to implement 5S methods in a clinical laboratory of a university hospital for evaluating its effect on employees' satisfaction, and correction of non-compliance in terms of the working environment. To start with, first, 5S education was given to management and employees. Secondly, a 5S team was formed and then the main steps of 5S (Seiri: Sort, Seiton: Set in order, Seiso: Shine, Seiketsu: Standardize, and Shitsuke: Systematize) were implemented for a duration of 3 months. A five-point likert scale questionnaire was used in order to determine and assess the impact of 5S on employees' satisfaction considering the areas such as facilitating the job, the job satisfaction, setting up a safe environment, and the effect of participation in management. Questionnaire form was given to 114 employees who actively worked during the 5S implementation period, and the data obtained from 63 (52.3%) participants (16 male, 47 female) were evaluated. The reliability of the questionnaire's Cronbach's alpha value was determined as 0.858 (p5S it was observed and determined that facilitating the job and setting up a safe environment created a statistically significant effect on

  4. Evaluating safety-critical organizations - emphasis on the nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-15

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

  5. Evaluating safety-critical organizations - emphasis on the nuclear industry

    International Nuclear Information System (INIS)

    Reiman, Teemu; Oedewald, Pia

    2009-04-01

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

  6. Evaluation for nuclear safety-critical software reliability of DCS

    International Nuclear Information System (INIS)

    Liu Ying

    2015-01-01

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

  7. Assessment of compliance with regulatory requirements for a best estimate methodology for evaluation of ECCS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Un Chul; Jang, Jin Wook; Lim, Ho Gon; Jeong, Ik [Seoul National Univ., Seoul (Korea, Republic of); Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    Best estimate methodolog