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Sample records for probability safety assessment

  1. Bounding probabilistic safety assessment probabilities by reality

    International Nuclear Information System (INIS)

    Fragola, J.R.; Shooman, M.L.

    1991-01-01

    The investigation of the failure in systems where failure is a rare event makes the continual comparisons between the developed probabilities and empirical evidence difficult. The comparison of the predictions of rare event risk assessments with historical reality is essential to prevent probabilistic safety assessment (PSA) predictions from drifting into fantasy. One approach to performing such comparisons is to search out and assign probabilities to natural events which, while extremely rare, have a basis in the history of natural phenomena or human activities. For example the Segovian aqueduct and some of the Roman fortresses in Spain have existed for several millennia and in many cases show no physical signs of earthquake damage. This evidence could be used to bound the probability of earthquakes above a certain magnitude to less than 10 -3 per year. On the other hand, there is evidence that some repetitive actions can be performed with extremely low historical probabilities when operators are properly trained and motivated, and sufficient warning indicators are provided. The point is not that low probability estimates are impossible, but continual reassessment of the analysis assumptions, and a bounding of the analysis predictions by historical reality. This paper reviews the probabilistic predictions of PSA in this light, attempts to develop, in a general way, the limits which can be historically established and the consequent bounds that these limits place upon the predictions, and illustrates the methodology used in computing such limits. Further, the paper discusses the use of empirical evidence and the requirement for disciplined systematic approaches within the bounds of reality and the associated impact on PSA probabilistic estimates

  2. The probability safety assessment impact on the BR2 refurbishment

    International Nuclear Information System (INIS)

    Pouleur, Yvan

    1995-01-01

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

  3. Probability safety assessment of LOOP accident to molten salt reactor

    International Nuclear Information System (INIS)

    Mei Mudan; Shao Shiwei; Yu Zhizhen; Chen Kun; Zuo Jiaxu

    2013-01-01

    Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10 -11 /(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

  4. On the meaning of probability in the context of probabilistic safety assessment

    International Nuclear Information System (INIS)

    Oestberg, G.

    1988-01-01

    Assessments of reliability and safety in technology require the application not only of frequentistic statistics but also of subjective estimates of probabilities. This is true in particular for decision-making about complex systems made up of both 'hard' and 'soft'elements. Problems occur when objective and subjective considerations have to be integrated and accommodated to the decision-makers' mental models. (author)

  5. Probability safety assessment of the Kozloduy-5 and Kozloduy-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, A; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    A probability safety assessment (PSA) of Level 1 (assessment of plant failures leading to the determination of core damage frequency) has been carried out for the NPP Kozloduy Units 5 and 6 (reactors WWER-1000). The scope of the study includes all significant accident initiators including seismic (earthquake) and fire initiators. Event trees for all initiators and fault trees for front line systems, support systems and major safety systems have been built. A distribution of the different initiators has been established as follows: internal initiators - 85%, seismic initiators - 5%, fire initiators- 10%. The loss of offsite power was identified as main contributor from the internal initiators with frequency 1,1.10{sup -4}/y. It is concluded that the safety functions of WWER-1000 are adequately covered by the safety systems. 4 refs., 2 tabs.

  6. Trends in development of probability assessment of nuclear power plant safety

    International Nuclear Information System (INIS)

    Dach, K.

    1989-01-01

    A complete study of probability safety assessment (PSA) of nuclear power plants is a multidisciplinary endeavor, requiring a qualified decision-making team composed of experienced professionals in individual disciplines and requiring good coordination of effort. The main concerns for the execution of a PSA study and related tasks are schematically presented. Also shown is a summary of the main steps for a PSA study at all three levels, with the incorporation of analysis of external events and the reliability of humans, including the necessary uncertainty analyses. 1 ref., 2 figs., 3 tabs

  7. 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

  8. 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)

  9. Joint probability safety assessment for NPP defense infrastructure against extreme external natural hazards

    International Nuclear Information System (INIS)

    Guilin, L.; Defu, L.; Huajun, L.; Fengqing, W.; Tao, Z.

    2012-01-01

    With the increasing tendency of natural hazards, the typhoon, hurricane and tropical Cyclone induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of planned, designed And constructed NPP the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane /Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compare with IAEA 2006-2009 recommended safety regulation design criteria for some NPP defense infrastructures along China coast. (authors)

  10. The philosophical basis for the use of probabilities in safety assessments

    International Nuclear Information System (INIS)

    Abramson, L.R.

    1988-01-01

    The axiomatic theory of probability is analogous to the theory of moving objects based on relations such as Newton's second law, F = ma. Each theory purports to describe the behavior of physical objects, and each has been validated by countless observations and experiments. In this sense, the probability of coming up heads is just as intrinsic a property of a real coin as is its mass. In contrast to the established validity of the axiomatic approach, the major weakness of the subjectivistic (Bayesian) approach to probability is the questionable connection between its conclusions and the real world. (author)

  11. Probability safety assessment activities in India for new and advanced reactors

    International Nuclear Information System (INIS)

    Guptan, R.; Ghagde, S.G.; Nama, R.; Varde, P.V.; Vinod, G.; Arul, J.; Solanki, R.B.

    2012-01-01

    This paper discusses, in brief, the salient features of the Level 1 PSA for New and Advanced reactors in India. The features of Level 1 PSA for new reactors are being discussed through a case study of 540 MWe twin unit (comprises of Unit 3 and 4) PHWRs at TAPS. The reactors uses Heavy water moderator and pressurized heavy water coolant, natural uranium fuel and horizontal pressure tubes. The major feature of PSA of advanced reactors is also discussed through the specific issues that were encountered during PSA modeling of AHWR (Advanced Heavy Water Reactor) and 700 MWe PHWR. The results of the PSA indicate that a fairly high level of redundancies exists in TAPS-3 and -4 design. It is recommended that staggered testing philosophy should be adopted especially for Emergency Core Cooling System, to reduce the probability of common cause failure among the motorized valves. It is also recommended to emphasize the importance of Small Break LOCA in general and their consequences in the licensing process of the plant operators

  12. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

    2008-10-15

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

  13. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    International Nuclear Information System (INIS)

    Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

    2008-01-01

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

  14. Application and problems of probability methods in technical safety assessment in the field of nuclear engineering and other technologies

    International Nuclear Information System (INIS)

    Mathiak, E.; Schuetz, B.

    1980-01-01

    The authors explain purpose, latest developments and application of probabilistic methods in safety assessments of nuclear facilities, and of non-nuclear installations. Their findings show that the methods of probabilistic systems analysis and of structural reliability analysis proved to be successful, above all with regard to systematics and reproducibility. Above all probabilistic systems analyses have been applied to a large extent in the Rasmussen study. Although this study has been intended to present objective information on the risks to be expected from nuclear power plant operation, the results of the study have not been accepted by the public as an unbiased presentation. It is worth mentioning that in the opinion of a number of social scientists, solutions accepted by the whole of society cannot be reached by defining and adhering to risk standards, but rather by entering into discussions with those groups directly affected, working out compromises meeting all interests. Risk analyses supply information that facilitates practical planning of emergency measures. A description of probable accidents allows conclusions to be drawn in terms of quality and quantity as to how and to what extent appropriate precautionary measures can be taken and planned. Risk analyses offer the possibility of preventing damage hitherto known only by experience (e.g. through accident analyses) by precalculating possible events, and then initiating the required improvements. It is these positive effects that make up the importance of such analyses. (orig./HSCH) [de

  15. Application and problems of probability methods in technical safety assessment in the field of nuclear engineering and other technologies

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1980-01-01

    On the basis of a deterministic safety concept that has been developed in nuclear engineering, approaches for a probabilistic interpretation of existing safety requirements and for a further risk assessment are described. The procedures in technical reliability analysis and its application in nuclear engineering are discussed. By the example of a reliability analysis for a reactor protection system the author discusses the question as to what extent methods of reliability analysis can be used to interpret deterministically derived safety requirements. The the author gives a survey of the current value and application of probabilistic reliability assessments in non-nuclear technology. The last part of this report deals with methods of risk analysis and its use for safety assessment in nuclear engineering. On the basis of WASH 1,400 the most important phases and tasks of research work in risk assessment are explained, showing the basic criteria and the methods to be applied in risk analysis. (orig./HSCH) [de

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  17. Probability and uncertainty in nuclear safety decisions

    International Nuclear Information System (INIS)

    Pate-Cornell, M.E.

    1986-01-01

    In this paper, we examine some problems posed by the use of probabilities in Nuclear Safety decisions. We discuss some of the theoretical difficulties due to the collective nature of regulatory decisions, and, in particular, the calibration and the aggregation of risk information (e.g., experts opinions). We argue that, if one chooses numerical safety goals as a regulatory basis, one can reduce the constraints to an individual safety goal and a cost-benefit criterion. We show the relevance of risk uncertainties in this kind of regulatory framework. We conclude that, whereas expected values of future failure frequencies are adequate to show compliance with economic constraints, the use of a fractile (e.g., 95%) to be specified by the regulatory agency is justified to treat hazard uncertainties for the individual safety goal. (orig.)

  18. PROBABILITY SURVEYS , CONDITIONAL PROBABILITIES AND ECOLOGICAL RISK ASSESSMENT

    Science.gov (United States)

    We show that probability-based environmental resource monitoring programs, such as the U.S. Environmental Protection Agency's (U.S. EPA) Environmental Monitoring and Assessment Program, and conditional probability analysis can serve as a basis for estimating ecological risk over ...

  19. Dependent Human Error Probability Assessment

    International Nuclear Information System (INIS)

    Simic, Z.; Mikulicic, V.; Vukovic, I.

    2006-01-01

    This paper presents an assessment of the dependence between dynamic operator actions modeled in a Nuclear Power Plant (NPP) PRA and estimate the associated impact on Core damage frequency (CDF). This assessment was done improve HEP dependencies implementation inside existing PRA. All of the dynamic operator actions modeled in the NPP PRA are included in this assessment. Determining the level of HEP dependence and the associated influence on CDF are the major steps of this assessment. A decision on how to apply the results, i.e., should permanent HEP model changes be made, is based on the resulting relative CDF increase. Some CDF increase was selected as a threshold based on the NPP base CDF value and acceptance guidelines from the Regulatory Guide 1.174. HEP dependence resulting in a CDF increase of > 5E-07 would be considered potential candidates for specific incorporation into the baseline model. The approach used to judge the level of dependence between operator actions is based on dependency level categories and conditional probabilities developed in the Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications NUREG/CR-1278. To simplify the process, NUREG/CR-1278 identifies five levels of dependence: ZD (zero dependence), LD (low dependence), MD (moderate dependence), HD (high dependence), and CD (complete dependence). NUREG/CR-1278 also identifies several qualitative factors that could be involved in determining the level of dependence. Based on the NUREG/CR-1278 information, Time, Function, and Spatial attributes were judged to be the most important considerations when determining the level of dependence between operator actions within an accident sequence. These attributes were used to develop qualitative criteria (rules) that were used to judge the level of dependence (CD, HD, MD, LD, ZD) between the operator actions. After the level of dependence between the various HEPs is judged, quantitative values associated with the

  20. Defining Probability in Sex Offender Risk Assessment.

    Science.gov (United States)

    Elwood, Richard W

    2016-12-01

    There is ongoing debate and confusion over using actuarial scales to predict individuals' risk of sexual recidivism. Much of the debate comes from not distinguishing Frequentist from Bayesian definitions of probability. Much of the confusion comes from applying Frequentist probability to individuals' risk. By definition, only Bayesian probability can be applied to the single case. The Bayesian concept of probability resolves most of the confusion and much of the debate in sex offender risk assessment. Although Bayesian probability is well accepted in risk assessment generally, it has not been widely used to assess the risk of sex offenders. I review the two concepts of probability and show how the Bayesian view alone provides a coherent scheme to conceptualize individuals' risk of sexual recidivism.

  1. Safety and reliability assessment

    International Nuclear Information System (INIS)

    1979-01-01

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

  2. Lessons Learnt in the Development of Level 1 PUSPATI TRIGA Reactor Probability Safety Assessment: A Collaboration Project under the Norwegian Extra Budgetary Fund

    International Nuclear Information System (INIS)

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

    2014-01-01

    This article reports about the lessons learnt from the development of level 1 probabilistic safety assessment (PSA) project that was implemented under the IAEA mentoring program for TRIGA MARK II PUSPATI research reactor (RTP). As a project that involved more than 3 organizations, a strategic planning of the management and implementation of individual assignment is truly a hectic task. This report compiles all related activities from the forming of the Malaysian PSA team up to the final report submitted to the IAEA. (author)

  3. Assessing the clinical probability of pulmonary embolism

    International Nuclear Information System (INIS)

    Miniati, M.; Pistolesi, M.

    2001-01-01

    Clinical assessment is a cornerstone of the recently validated diagnostic strategies for pulmonary embolism (PE). Although the diagnostic yield of individual symptoms, signs, and common laboratory tests is limited, the combination of these variables, either by empirical assessment or by a prediction rule, can be used to express a clinical probability of PE. The latter may serve as pretest probability to predict the probability of PE after further objective testing (posterior or post-test probability). Over the last few years, attempts have been made to develop structured prediction models for PE. In a Canadian multicenter prospective study, the clinical probability of PE was rated as low, intermediate, or high according to a model which included assessment of presenting symptoms and signs, risk factors, and presence or absence of an alternative diagnosis at least as likely as PE. Recently, a simple clinical score was developed to stratify outpatients with suspected PE into groups with low, intermediate, or high clinical probability. Logistic regression was used to predict parameters associated with PE. A score ≤ 4 identified patients with low probability of whom 10% had PE. The prevalence of PE in patients with intermediate (score 5-8) and high probability (score ≥ 9) was 38 and 81%, respectively. As opposed to the Canadian model, this clinical score is standardized. The predictor variables identified in the model, however, were derived from a database of emergency ward patients. This model may, therefore, not be valid in assessing the clinical probability of PE in inpatients. In the PISA-PED study, a clinical diagnostic algorithm was developed which rests on the identification of three relevant clinical symptoms and on their association with electrocardiographic and/or radiographic abnormalities specific for PE. Among patients who, according to the model, had been rated as having a high clinical probability, the prevalence of proven PE was 97%, while it was 3

  4. Heart sounds analysis using probability assessment

    Czech Academy of Sciences Publication Activity Database

    Plešinger, Filip; Viščor, Ivo; Halámek, Josef; Jurčo, Juraj; Jurák, Pavel

    2017-01-01

    Roč. 38, č. 8 (2017), s. 1685-1700 ISSN 0967-3334 R&D Projects: GA ČR GAP102/12/2034; GA MŠk(CZ) LO1212; GA MŠk ED0017/01/01 Institutional support: RVO:68081731 Keywords : heart sounds * FFT * machine learning * signal averaging * probability assessment Subject RIV: FS - Medical Facilities ; Equipment OBOR OECD: Medical engineering Impact factor: 2.058, year: 2016

  5. VOLCANIC RISK ASSESSMENT - PROBABILITY AND CONSEQUENCES

    International Nuclear Information System (INIS)

    G.A. Valentine; F.V. Perry; S. Dartevelle

    2005-01-01

    Risk is the product of the probability and consequences of an event. Both of these must be based upon sound science that integrates field data, experiments, and modeling, but must also be useful to decision makers who likely do not understand all aspects of the underlying science. We review a decision framework used in many fields such as performance assessment for hazardous and/or radioactive waste disposal sites that can serve to guide the volcanological community towards integrated risk assessment. In this framework the underlying scientific understanding of processes that affect probability and consequences drive the decision-level results, but in turn these results can drive focused research in areas that cause the greatest level of uncertainty at the decision level. We review two examples of the determination of volcanic event probability: (1) probability of a new volcano forming at the proposed Yucca Mountain radioactive waste repository, and (2) probability that a subsurface repository in Japan would be affected by the nearby formation of a new stratovolcano. We also provide examples of work on consequences of explosive eruptions, within the framework mentioned above. These include field-based studies aimed at providing data for ''closure'' of wall rock erosion terms in a conduit flow model, predictions of dynamic pressure and other variables related to damage by pyroclastic flow into underground structures, and vulnerability criteria for structures subjected to conditions of explosive eruption. Process models (e.g., multiphase flow) are important for testing the validity or relative importance of possible scenarios in a volcanic risk assessment. We show how time-dependent multiphase modeling of explosive ''eruption'' of basaltic magma into an open tunnel (drift) at the Yucca Mountain repository provides insight into proposed scenarios that include the development of secondary pathways to the Earth's surface. Addressing volcanic risk within a decision

  6. Comment on 'The meaning of probability in probabilistic safety analysis'

    International Nuclear Information System (INIS)

    Yellman, Ted W.; Murray, Thomas M.

    1995-01-01

    A recent article in Reliability Engineering and System Safety argues that there is 'fundamental confusion over how to interpret the numbers which emerge from a Probabilistic Safety Analysis [PSA]', [Watson, S. R., The meaning of probability in probabilistic safety analysis. Reliab. Engng and System Safety, 45 (1994) 261-269.] As a standard for comparison, the author employs the 'realist' interpretation that a PSA output probability should be a 'physical property' of the installation being analyzed, 'objectively measurable' without controversy. The author finds all the other theories and philosophies discussed wanting by this standard. Ultimately, he argues that the outputs of a PSA should be considered to be no more than constructs of the computational procedure chosen - just an 'argument' or a 'framework for the debate about safety' rather than a 'representation of truth'. He even suggests that 'competing' PSA's be done - each trying to 'argue' for a different message. The commentors suggest that the position the author arrives at is an overreaction to the subjectivity which is part of any complex PSA, and that that overreaction could in fact easily lead to the belief that PSA's are meaningless. They suggest a broader interpretation, one based strictly on relative frequency--a concept which the commentors believe the author abandoned too quickly. Their interpretation does not require any 'tests' to determine whether a statement of likelihood is qualified to be a 'true' probability and it applies equally well in pure analytical models. It allows anyone's proper numerical statement of the likelihood of an event to be considered a probability. It recognizes that the quality of PSA's and their results will vary. But, unlike the author, the commentors contend that a PSA should always be a search for truth--not a vehicle for adversarial pleadings

  7. Risk prediction, safety analysis and quantitative probability methods - a caveat

    International Nuclear Information System (INIS)

    Critchley, O.H.

    1976-01-01

    Views are expressed on the use of quantitative techniques for the determination of value judgements in nuclear safety assessments, hazard evaluation, and risk prediction. Caution is urged when attempts are made to quantify value judgements in the field of nuclear safety. Criteria are given the meaningful application of reliability methods but doubts are expressed about their application to safety analysis, risk prediction and design guidances for experimental or prototype plant. Doubts are also expressed about some concomitant methods of population dose evaluation. The complexities of new designs of nuclear power plants make the problem of safety assessment more difficult but some possible approaches are suggested as alternatives to the quantitative techniques criticized. (U.K.)

  8. Heart sounds analysis using probability assessment.

    Science.gov (United States)

    Plesinger, F; Viscor, I; Halamek, J; Jurco, J; Jurak, P

    2017-07-31

    This paper describes a method for automated discrimination of heart sounds recordings according to the Physionet Challenge 2016. The goal was to decide if the recording refers to normal or abnormal heart sounds or if it is not possible to decide (i.e. 'unsure' recordings). Heart sounds S1 and S2 are detected using amplitude envelopes in the band 15-90 Hz. The averaged shape of the S1/S2 pair is computed from amplitude envelopes in five different bands (15-90 Hz; 55-150 Hz; 100-250 Hz; 200-450 Hz; 400-800 Hz). A total of 53 features are extracted from the data. The largest group of features is extracted from the statistical properties of the averaged shapes; other features are extracted from the symmetry of averaged shapes, and the last group of features is independent of S1 and S2 detection. Generated features are processed using logical rules and probability assessment, a prototype of a new machine-learning method. The method was trained using 3155 records and tested on 1277 hidden records. It resulted in a training score of 0.903 (sensitivity 0.869, specificity 0.937) and a testing score of 0.841 (sensitivity 0.770, specificity 0.913). The revised method led to a test score of 0.853 in the follow-up phase of the challenge. The presented solution achieved 7th place out of 48 competing entries in the Physionet Challenge 2016 (official phase). In addition, the PROBAfind software for probability assessment was introduced.

  9. Safety Assessment for Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-06-15

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

  10. Assessment of Safety Culture

    International Nuclear Information System (INIS)

    Bilic Zabric, T.; Kavsek, D.

    2006-01-01

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

  11. Safety Auditing and Assessments

    Science.gov (United States)

    Goodin, James Ronald (Ronnie)

    2005-01-01

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

  12. Human factors in safety assessment. Safety culture assessment

    International Nuclear Information System (INIS)

    Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

    1996-01-01

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

  13. Safety assessments for potential exposures

    International Nuclear Information System (INIS)

    Dunn, D.I.

    2012-04-01

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

  14. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. Reactor safety assessment system

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  16. Metrics design for safety assessment

    NARCIS (Netherlands)

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

    2016-01-01

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

  17. Assessment of criticality safety

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Heaberlin, S.W.; Clayton, E.D.; Carter, R.D.

    1979-01-01

    A study was made of 100 violations of criticality safety specifications reported over a 10-y period in the operations of fuel reprocessing plants. The seriousness of each rule violation was evaluated by assigning it a severity index value. The underlying causes or reasons, for the violations were identified. A criticality event tree was constructed using the parameters, causes, and reasons found in the analysis of the infractions. The event tree provides a means for visualizing the paths to an accidental criticality. Some 65% of the violations were caused by misinterpretation on the part of the operator, being attributed to a lack of clarity in the specification and insufficient training; 33% were attributed to lack of care, whereas only 2% were caused by mechanical failure. A fault tree was constructed by assembling the events that could contribute to an accident. With suitable data on the probabilities of contributing events, the probability of the accident's occurrence can be forecast. Estimated probabilities for criticality were made, based on the limited data available, that in this case indicate a minimum time span of 244 y of plant operation per accident ranging up to approx. 3000 y subject to the various underlying assumptions made. Some general suggestions for improvement are formulated based on the cases studied. Although conclusions for other plants may differ in detail, the general method of analysis and the fault tree logic should prove applicable. 4 figures, 8 tables

  18. Probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hoertner, H.; Schuetz, B.

    1982-09-01

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

  19. Probability

    CERN Document Server

    Shiryaev, A N

    1996-01-01

    This book contains a systematic treatment of probability from the ground up, starting with intuitive ideas and gradually developing more sophisticated subjects, such as random walks, martingales, Markov chains, ergodic theory, weak convergence of probability measures, stationary stochastic processes, and the Kalman-Bucy filter Many examples are discussed in detail, and there are a large number of exercises The book is accessible to advanced undergraduates and can be used as a text for self-study This new edition contains substantial revisions and updated references The reader will find a deeper study of topics such as the distance between probability measures, metrization of weak convergence, and contiguity of probability measures Proofs for a number of some important results which were merely stated in the first edition have been added The author included new material on the probability of large deviations, and on the central limit theorem for sums of dependent random variables

  20. The estimation of small probabilities and risk assessment

    International Nuclear Information System (INIS)

    Kalbfleisch, J.D.; Lawless, J.F.; MacKay, R.J.

    1982-01-01

    The primary contribution of statistics to risk assessment is in the estimation of probabilities. Frequently the probabilities in question are small, and their estimation is particularly difficult. The authors consider three examples illustrating some problems inherent in the estimation of small probabilities

  1. Landslide Probability Assessment by the Derived Distributions Technique

    Science.gov (United States)

    Muñoz, E.; Ochoa, A.; Martínez, H.

    2012-12-01

    Landslides are potentially disastrous events that bring along human and economic losses; especially in cities where an accelerated and unorganized growth leads to settlements on steep and potentially unstable areas. Among the main causes of landslides are geological, geomorphological, geotechnical, climatological, hydrological conditions and anthropic intervention. This paper studies landslides detonated by rain, commonly known as "soil-slip", which characterize by having a superficial failure surface (Typically between 1 and 1.5 m deep) parallel to the slope face and being triggered by intense and/or sustained periods of rain. This type of landslides is caused by changes on the pore pressure produced by a decrease in the suction when a humid front enters, as a consequence of the infiltration initiated by rain and ruled by the hydraulic characteristics of the soil. Failure occurs when this front reaches a critical depth and the shear strength of the soil in not enough to guarantee the stability of the mass. Critical rainfall thresholds in combination with a slope stability model are widely used for assessing landslide probability. In this paper we present a model for the estimation of the occurrence of landslides based on the derived distributions technique. Since the works of Eagleson in the 1970s the derived distributions technique has been widely used in hydrology to estimate the probability of occurrence of extreme flows. The model estimates the probability density function (pdf) of the Factor of Safety (FOS) from the statistical behavior of the rainfall process and some slope parameters. The stochastic character of the rainfall is transformed by means of a deterministic failure model into FOS pdf. Exceedance probability and return period estimation is then straightforward. The rainfall process is modeled as a Rectangular Pulses Poisson Process (RPPP) with independent exponential pdf for mean intensity and duration of the storms. The Philip infiltration model

  2. Incidents in nuclear research reactor examined by deterministic probability and probabilistic safety analysis

    International Nuclear Information System (INIS)

    Lopes, Valdir Maciel

    2010-01-01

    This study aims to evaluate the potential risks submitted by the incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency, IAEA, were used, the Incident Report System for Research Reactor and Research Reactor Data Base. For this type of assessment was used the Probabilistic Safety Analysis (PSA), within a confidence level of 90% and the Deterministic Probability Analysis (DPA). To obtain the results of calculations of probabilities for PSA, were used the theory and equations in the paper IAEA TECDOC - 636. The development of the calculations of probabilities for PSA was used the program Scilab version 5.1.1, free access, executable on Windows and Linux platforms. A specific program to get the results of probability was developed within the main program Scilab 5.1.1., for two distributions Fischer and Chi-square, both with the confidence level of 90%. Using the Sordi equations and Origin 6.0 program, were obtained the maximum admissible doses related to satisfy the risk limits established by the International Commission on Radiological Protection, ICRP, and were also obtained these maximum doses graphically (figure 1) resulting from the calculations of probabilities x maximum admissible doses. It was found that the reliability of the results of probability is related to the operational experience (reactor x year and fractions) and that the larger it is, greater the confidence in the outcome. Finally, a suggested list of future work to complement this paper was gathered. (author)

  3. [Agricultural biotechnology safety assessment].

    Science.gov (United States)

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

    2015-01-01

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

  4. Probability, conditional probability and complementary cumulative distribution functions in performance assessment for radioactive waste disposal

    International Nuclear Information System (INIS)

    Helton, J.C.

    1996-03-01

    A formal description of the structure of several recent performance assessments (PAs) for the Waste Isolation Pilot Plant (WIPP) is given in terms of the following three components: a probability space (S st , S st , p st ) for stochastic uncertainty, a probability space (S su , S su , p su ) for subjective uncertainty and a function (i.e., a random variable) defined on the product space associated with (S st , S st , p st ) and (S su , S su , p su ). The explicit recognition of the existence of these three components allows a careful description of the use of probability, conditional probability and complementary cumulative distribution functions within the WIPP PA. This usage is illustrated in the context of the U.S. Environmental Protection Agency's standard for the geologic disposal of radioactive waste (40 CFR 191, Subpart B). The paradigm described in this presentation can also be used to impose a logically consistent structure on PAs for other complex systems

  5. Probability, conditional probability and complementary cumulative distribution functions in performance assessment for radioactive waste disposal

    International Nuclear Information System (INIS)

    Helton, J.C.

    1996-01-01

    A formal description of the structure of several recent performance assessments (PAs) for the Waste Isolation Pilot Plant (WIPP) is given in terms of the following three components: a probability space (S st , L st , P st ) for stochastic uncertainty, a probability space (S su , L su , P su ) for subjective uncertainty and a function (i.e., a random variable) defined on the product space associated with (S st , L st , P st ) and (S su , L su , P su ). The explicit recognition of the existence of these three components allows a careful description of the use of probability, conditional probability and complementary cumulative distribution functions within the WIPP PA. This usage is illustrated in the context of the US Environmental Protection Agency's standard for the geologic disposal of radioactive waste (40 CFR 191, Subpart B). The paradigm described in this presentation can also be used to impose a logically consistent structure on PAs for other complex systems

  6. HSE's safety assessment principles for criticality safety

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  7. Probability calculations for three-part mineral resource assessments

    Science.gov (United States)

    Ellefsen, Karl J.

    2017-06-27

    Three-part mineral resource assessment is a methodology for predicting, in a specified geographic region, both the number of undiscovered mineral deposits and the amount of mineral resources in those deposits. These predictions are based on probability calculations that are performed with computer software that is newly implemented. Compared to the previous implementation, the new implementation includes new features for the probability calculations themselves and for checks of those calculations. The development of the new implementation lead to a new understanding of the probability calculations, namely the assumptions inherent in the probability calculations. Several assumptions strongly affect the mineral resource predictions, so it is crucial that they are checked during an assessment. The evaluation of the new implementation leads to new findings about the probability calculations,namely findings regarding the precision of the computations,the computation time, and the sensitivity of the calculation results to the input.

  8. Impact of proof test interval and coverage on probability of failure of safety instrumented function

    International Nuclear Information System (INIS)

    Jin, Jianghong; Pang, Lei; Hu, Bin; Wang, Xiaodong

    2016-01-01

    Highlights: • Introduction of proof test coverage makes the calculation of the probability of failure for SIF more accurate. • The probability of failure undetected by proof test is independently defined as P TIF and calculated. • P TIF is quantified using reliability block diagram and simple formula of PFD avg . • Improving proof test coverage and adopting reasonable test period can reduce the probability of failure for SIF. - Abstract: Imperfection of proof test can result in the safety function failure of safety instrumented system (SIS) at any time in its life period. IEC61508 and other references ignored or only elementarily analyzed the imperfection of proof test. In order to further study the impact of the imperfection of proof test on the probability of failure for safety instrumented function (SIF), the necessity of proof test and influence of its imperfection on system performance was first analyzed theoretically. The probability of failure for safety instrumented function resulted from the imperfection of proof test was defined as probability of test independent failures (P TIF ), and P TIF was separately calculated by introducing proof test coverage and adopting reliability block diagram, with reference to the simplified calculation formula of average probability of failure on demand (PFD avg ). Research results show that: the shorter proof test period and the higher proof test coverage indicate the smaller probability of failure for safety instrumented function. The probability of failure for safety instrumented function which is calculated by introducing proof test coverage will be more accurate.

  9. Wald Sequential Probability Ratio Test for Space Object Conjunction Assessment

    Science.gov (United States)

    Carpenter, James R.; Markley, F Landis

    2014-01-01

    This paper shows how satellite owner/operators may use sequential estimates of collision probability, along with a prior assessment of the base risk of collision, in a compound hypothesis ratio test to inform decisions concerning collision risk mitigation maneuvers. The compound hypothesis test reduces to a simple probability ratio test, which appears to be a novel result. The test satisfies tolerances related to targeted false alarm and missed detection rates. This result is independent of the method one uses to compute the probability density that one integrates to compute collision probability. A well-established test case from the literature shows that this test yields acceptable results within the constraints of a typical operational conjunction assessment decision timeline. Another example illustrates the use of the test in a practical conjunction assessment scenario based on operations of the International Space Station.

  10. Experiences in assessing safety culture

    International Nuclear Information System (INIS)

    Spitalnik, J.

    2002-01-01

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

  11. Dynamic probability evaluation of safety levels of earth-rockfill dams using Bayesian approach

    Directory of Open Access Journals (Sweden)

    Zi-wu Fan

    2009-06-01

    Full Text Available In order to accurately predict and control the aging process of dams, new information should be collected continuously to renew the quantitative evaluation of dam safety levels. Owing to the complex structural characteristics of dams, it is quite difficult to predict the time-varying factors affecting their safety levels. It is not feasible to employ dynamic reliability indices to evaluate the actual safety levels of dams. Based on the relevant regulations for dam safety classification in China, a dynamic probability description of dam safety levels was developed. Using the Bayesian approach and effective information mining, as well as real-time information, this study achieved more rational evaluation and prediction of dam safety levels. With the Bayesian expression of discrete stochastic variables, the a priori probabilities of the dam safety levels determined by experts were combined with the likelihood probability of the real-time check information, and the probability information for the evaluation of dam safety levels was renewed. The probability index was then applied to dam rehabilitation decision-making. This method helps reduce the difficulty and uncertainty of the evaluation of dam safety levels and complies with the current safe decision-making regulations for dams in China. It also enhances the application of current risk analysis methods for dam safety levels.

  12. Probability, conditional probability and complementary cumulative distribution functions in performance assessment for radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Helton, J.C. [Arizona State Univ., Tempe, AZ (United States)

    1996-03-01

    A formal description of the structure of several recent performance assessments (PAs) for the Waste Isolation Pilot Plant (WIPP) is given in terms of the following three components: a probability space (S{sub st}, S{sub st}, p{sub st}) for stochastic uncertainty, a probability space (S{sub su}, S{sub su}, p{sub su}) for subjective uncertainty and a function (i.e., a random variable) defined on the product space associated with (S{sub st}, S{sub st}, p{sub st}) and (S{sub su}, S{sub su}, p{sub su}). The explicit recognition of the existence of these three components allows a careful description of the use of probability, conditional probability and complementary cumulative distribution functions within the WIPP PA. This usage is illustrated in the context of the U.S. Environmental Protection Agency`s standard for the geologic disposal of radioactive waste (40 CFR 191, Subpart B). The paradigm described in this presentation can also be used to impose a logically consistent structure on PAs for other complex systems.

  13. An extended car-following model considering random safety distance with different probabilities

    Science.gov (United States)

    Wang, Jufeng; Sun, Fengxin; Cheng, Rongjun; Ge, Hongxia; Wei, Qi

    2018-02-01

    Because of the difference in vehicle type or driving skill, the driving strategy is not exactly the same. The driving speeds of the different vehicles may be different for the same headway. Since the optimal velocity function is just determined by the safety distance besides the maximum velocity and headway, an extended car-following model accounting for random safety distance with different probabilities is proposed in this paper. The linear stable condition for this extended traffic model is obtained by using linear stability theory. Numerical simulations are carried out to explore the complex phenomenon resulting from multiple safety distance in the optimal velocity function. The cases of multiple types of safety distances selected with different probabilities are presented. Numerical results show that the traffic flow with multiple safety distances with different probabilities will be more unstable than that with single type of safety distance, and will result in more stop-and-go phenomena.

  14. Human Error Probability Assessment During Maintenance Activities of Marine Systems

    Directory of Open Access Journals (Sweden)

    Rabiul Islam

    2018-03-01

    Full Text Available Background: Maintenance operations on-board ships are highly demanding. Maintenance operations are intensive activities requiring high man–machine interactions in challenging and evolving conditions. The evolving conditions are weather conditions, workplace temperature, ship motion, noise and vibration, and workload and stress. For example, extreme weather condition affects seafarers' performance, increasing the chances of error, and, consequently, can cause injuries or fatalities to personnel. An effective human error probability model is required to better manage maintenance on-board ships. The developed model would assist in developing and maintaining effective risk management protocols. Thus, the objective of this study is to develop a human error probability model considering various internal and external factors affecting seafarers' performance. Methods: The human error probability model is developed using probability theory applied to Bayesian network. The model is tested using the data received through the developed questionnaire survey of >200 experienced seafarers with >5 years of experience. The model developed in this study is used to find out the reliability of human performance on particular maintenance activities. Results: The developed methodology is tested on the maintenance of marine engine's cooling water pump for engine department and anchor windlass for deck department. In the considered case studies, human error probabilities are estimated in various scenarios and the results are compared between the scenarios and the different seafarer categories. The results of the case studies for both departments are also compared. Conclusion: The developed model is effective in assessing human error probabilities. These probabilities would get dynamically updated as and when new information is available on changes in either internal (i.e., training, experience, and fatigue or external (i.e., environmental and operational conditions

  15. Reducing the Probability of Incidents Through Behavior-Based Safety -- An Anomaly or Not?

    International Nuclear Information System (INIS)

    Turek, John A

    2002-01-01

    Reducing the probability of incidents through Behavior-Based Safety--an anomaly or not? Can a Behavior-Based Safety (BBS) process reduce the probability of an employee sustaining a work-related injury or illness? This presentation describes the actions taken to implement a sustainable BBS process and evaluates its effectiveness. The BBS process at the Stanford Linear Accelerator Center used a pilot population of national laboratory employees to: Achieve employee and management support; Reduce the probability of employees' sustaining work-related injuries and illnesses; and Provide support for additional funding to expand within the laboratory

  16. Safety Assessment of Probiotics

    Science.gov (United States)

    Lahtinen, Sampo J.; Boyle, Robert J.; Margolles, Abelardo; Frias, Rafael; Gueimonde, Miguel

    Viable microbes have been a natural part of human diet throughout the history of mankind. Today, different fermented foods and other foods containing live microbes are consumed around the world, including industrialized countries, where the diet has become increasingly sterile during the last decades. By definition, probiotics are viable microbes with documented beneficial effects on host health. Probiotics have an excellent safety record, both in humans and in animals. Despite the wide and continuously increasing consumption of probiotics, adverse events related to probiotic use are extremely rare. Many popular probiotic strains such as lactobacilli and bifidobacteria can be considered as components of normal healthy intestinal microbiota, and thus are not thought to pose a risk for the host health - in contrast, beneficial effects on health are commonly reported. Nevertheless, the safety of probiotics is an important issue, in particular in the case of new potential probiotics which do not have a long history of safe use, and of probiotics belonging to species for which general assumption of safety cannot be made. Furthermore, safety of probiotics in high-risk populations such as critically ill patients and immunocompromized subjects deserves particular attention, as virtually all reported cases of bacteremia and fungemia associated with probiotic use, involve subjects with underlying diseases, compromised immune system or compromised intestinal integrity.

  17. Uncertainty analysis in safety assessment

    International Nuclear Information System (INIS)

    Lemos, Francisco Luiz de; Sullivan, Terry

    1997-01-01

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

  18. Assigning probability distributions to input parameters of performance assessment models

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, Srikanta [INTERA Inc., Austin, TX (United States)

    2002-02-01

    This study presents an overview of various approaches for assigning probability distributions to input parameters and/or future states of performance assessment models. Specifically,three broad approaches are discussed for developing input distributions: (a) fitting continuous distributions to data, (b) subjective assessment of probabilities, and (c) Bayesian updating of prior knowledge based on new information. The report begins with a summary of the nature of data and distributions, followed by a discussion of several common theoretical parametric models for characterizing distributions. Next, various techniques are presented for fitting continuous distributions to data. These include probability plotting, method of moments, maximum likelihood estimation and nonlinear least squares analysis. The techniques are demonstrated using data from a recent performance assessment study for the Yucca Mountain project. Goodness of fit techniques are also discussed, followed by an overview of how distribution fitting is accomplished in commercial software packages. The issue of subjective assessment of probabilities is dealt with in terms of the maximum entropy distribution selection approach, as well as some common rules for codifying informal expert judgment. Formal expert elicitation protocols are discussed next, and are based primarily on the guidance provided by the US NRC. The Bayesian framework for updating prior distributions (beliefs) when new information becomes available is discussed. A simple numerical approach is presented for facilitating practical applications of the Bayes theorem. Finally, a systematic framework for assigning distributions is presented: (a) for the situation where enough data are available to define an empirical CDF or fit a parametric model to the data, and (b) to deal with the situation where only a limited amount of information is available.

  19. Assigning probability distributions to input parameters of performance assessment models

    International Nuclear Information System (INIS)

    Mishra, Srikanta

    2002-02-01

    This study presents an overview of various approaches for assigning probability distributions to input parameters and/or future states of performance assessment models. Specifically,three broad approaches are discussed for developing input distributions: (a) fitting continuous distributions to data, (b) subjective assessment of probabilities, and (c) Bayesian updating of prior knowledge based on new information. The report begins with a summary of the nature of data and distributions, followed by a discussion of several common theoretical parametric models for characterizing distributions. Next, various techniques are presented for fitting continuous distributions to data. These include probability plotting, method of moments, maximum likelihood estimation and nonlinear least squares analysis. The techniques are demonstrated using data from a recent performance assessment study for the Yucca Mountain project. Goodness of fit techniques are also discussed, followed by an overview of how distribution fitting is accomplished in commercial software packages. The issue of subjective assessment of probabilities is dealt with in terms of the maximum entropy distribution selection approach, as well as some common rules for codifying informal expert judgment. Formal expert elicitation protocols are discussed next, and are based primarily on the guidance provided by the US NRC. The Bayesian framework for updating prior distributions (beliefs) when new information becomes available is discussed. A simple numerical approach is presented for facilitating practical applications of the Bayes theorem. Finally, a systematic framework for assigning distributions is presented: (a) for the situation where enough data are available to define an empirical CDF or fit a parametric model to the data, and (b) to deal with the situation where only a limited amount of information is available

  20. 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

  1. Confidence building in safety assessments

    International Nuclear Information System (INIS)

    Grundfelt, Bertil

    1999-01-01

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

  2. Propensity scores-potential outcomes framework to incorporate severity probabilities in the highway safety manual crash prediction algorithm.

    Science.gov (United States)

    Sasidharan, Lekshmi; Donnell, Eric T

    2014-10-01

    Accurate estimation of the expected number of crashes at different severity levels for entities with and without countermeasures plays a vital role in selecting countermeasures in the framework of the safety management process. The current practice is to use the American Association of State Highway and Transportation Officials' Highway Safety Manual crash prediction algorithms, which combine safety performance functions and crash modification factors, to estimate the effects of safety countermeasures on different highway and street facility types. Many of these crash prediction algorithms are based solely on crash frequency, or assume that severity outcomes are unchanged when planning for, or implementing, safety countermeasures. Failing to account for the uncertainty associated with crash severity outcomes, and assuming crash severity distributions remain unchanged in safety performance evaluations, limits the utility of the Highway Safety Manual crash prediction algorithms in assessing the effect of safety countermeasures on crash severity. This study demonstrates the application of a propensity scores-potential outcomes framework to estimate the probability distribution for the occurrence of different crash severity levels by accounting for the uncertainties associated with them. The probability of fatal and severe injury crash occurrence at lighted and unlighted intersections is estimated in this paper using data from Minnesota. The results show that the expected probability of occurrence of fatal and severe injury crashes at a lighted intersection was 1 in 35 crashes and the estimated risk ratio indicates that the respective probabilities at an unlighted intersection was 1.14 times higher compared to lighted intersections. The results from the potential outcomes-propensity scores framework are compared to results obtained from traditional binary logit models, without application of propensity scores matching. Traditional binary logit analysis suggests that

  3. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

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

  4. Additional components of risk assessment and their impact on the probability parameter

    Directory of Open Access Journals (Sweden)

    Piotr Saja

    2017-04-01

    Full Text Available The article raises the issue of risk assessment and its impact on the quality and safety of work. During the assessment of the turning lathe position additional components associated with the jobs personalization were taken into account. Paragraph 2 item 7 of the Regulation of the Minister of Laborr and Social Policy of 26 September 1997 on general safety regulations defines occupational risk as the likelihood of an adverse event. The authors drew attention to the reality of the accident, which sometimes depends on the predisposition of the employee. It turns out that a correct estimation of the probability of occurrence of the accident to be able to react in a timely way seems extremely important.. This parameter will be assessed more accurately if we take into account a number of additional components resulting from the characteristics of the employee. The results of the personalized assessment of risk may allow appropriate planning of corrective and preventive actions.

  5. Uncertainty analysis in safety assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  6. Safety management system needs assessment.

    Science.gov (United States)

    2016-04-01

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

  7. Prospects for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hirschberg, S.

    1992-01-01

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

  8. Living probabilistic safety assessment (LPSA)

    International Nuclear Information System (INIS)

    1999-08-01

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

  9. 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.

  10. Vulnerability assessment: Determining probabilities of neutralization of adversaries

    International Nuclear Information System (INIS)

    Graves, B.R.

    1987-01-01

    The Security Manager charged with the responsibility of designing Safeguards and Security Systems at Department of Energy facilities must take many factors into consideration. There must be a clear understanding, supported by documented guidance, of the level of threat to be addressed; the nature of the facility to be protected, and the funds available to design, implement, and maintain the Safeguards and Security System. Armed with these prerequisites, the Security Manager may then determine the characteristics of the Safeguards measures and security forces necessary to protect the facility. Security forces selection and training programs may then be established based on realistic facility needs. The next step is to attempt to determine the probability of security forces winning in a confrontation with adversaries. To determine the probability of success the Security Manager must consider the characteristics of the facility and surrounding area; the characteristics of the security forces and safeguards system at the facility; the response time and capabilities of the augmentation forces and the characteristics and capabilities of the adversary threat level to be addressed. Obviously, the Safeguards and Security Systems must initially address ''worst case'' scenarios consistent with stated guidelines. Validation of the assessment of the Safeguards and Security Systems must then be determined by simulation testing of the capabilities of the response forces against the capabilities of the adversary

  11. Controlling principles for prior probability assignments in nuclear risk assessment

    International Nuclear Information System (INIS)

    Cook, I.; Unwin, S.D.

    1986-01-01

    As performed conventionally, nuclear probabilistic risk assessment (PRA) may be criticized as utilizing inscrutable and unjustifiably ''precise'' quantitative informed judgment or extrapolation from that judgment. To meet this criticism, controlling principles that govern the formulation of probability densities are proposed, given only the informed input that would be required for a simple bounding analysis. These principles are founded upon information theoretic ideas of maximum uncertainty and cover both cases in which there exists a stochastic model of the phenomenon of interest and cases in which these is no such model. In part, the principles are conventional, and such an approach is justified by appealing to certain analogies in accounting practice and judicial decision making. Examples are given. Appropriate employment of these principles is expected to facilitate substantial progress toward PRA scrutability and transparency

  12. Understanding and assessing safety culture

    International Nuclear Information System (INIS)

    Dalling, Ian

    1997-01-01

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

  13. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  14. Framework of nuclear safety and safety assessment

    International Nuclear Information System (INIS)

    Furuta, Kazuo

    2007-01-01

    Since enormous energy is released by nuclear chain reaction mainly as a form of radiation, a great potential risk accompanies utilization of nuclear energy. Safety has been continuously a critical issue therefore from the very beginning of its development. Though the framework of nuclear safety that has been established at an early developmental stage of nuclear engineering is still valid, more comprehensive approaches are required having experienced several events such as Three Mile Island, Chernobyl, and JCO. This article gives a brief view of the most basic principles how nuclear safety is achieved, which were introduced and sophisticated in nuclear engineering but applicable also to other engineering domains in general. (author)

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

    International Nuclear Information System (INIS)

    Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

    2012-04-01

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

  16. Confidence building in safety assessment

    International Nuclear Information System (INIS)

    Osthols, E.

    1999-01-01

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

  17. Engineering safety assessment. An introduction

    International Nuclear Information System (INIS)

    Thomson, J.R.

    1987-01-01

    There have been several major industrial accidents in recent years, notably the Chernobyl reactor accident, the Bhopal disaster and the Mexico City gas explosions. It is essential that engineers should understand the natures of the hazards and risks of industrial plant. This book is intended to give an introduction to this subject. Determining the risks associated with hazardous plant such as a chemical plant or a nuclear power station requires knowledge of the probability of a given accident, as well as knowledge of the likely consequences in terms of human lives or damaged property. In this book, the author explains these two main themes in assessing overall risk. He introduces basic probability theory, the assessment of systems reliabilities and structural reliabilities, and the means used to estimate the consequences of hypothetical major accidents. In particular, the methodologies for estimating the consequences of accidental explosion and accidents involving the release of airborne toxic materials (including chemical toxins, carcinogens, and radioactive materials) are explained. A final chapter explores the difficult question of 'acceptable' risk, and how much can justifiably be spent on risk reduction. Intended for use by final year students of chemical, mechanical or nuclear engineering, this book will also be suitable for postgraduate students and practising engineers and scientists whose work involves them in risk assessment. (author)

  18. Assessment of radiation safety in well logging

    International Nuclear Information System (INIS)

    Alles, A.; Pérez, Y.; Duménigo, C.

    2015-01-01

    Radiation safety assessments required by current regulations are a means to verify compliance with the requirements. Different methods have been used for this purpose. In the paper the results of applying the method of risk matrices, applied for the first time in the practice of well logging are exposed. For each initiating event frequency of occurrence, the severity of its consequences and the probability of failure of the barriers identified were evaluated. Starting from these assumptions, the risk associated is determined for each of the identified accident sequences, using for this the SEVRRA code 'Risk Assessment System', originally designed for use in radiotherapy. As an result sequences increased risk associated with the practice of well logging were identified, which is the starting point for the further implementation of a coherent program of dose optimization in practice. [es

  19. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

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

  20. Probabilistic risk assessment on maritime spent nuclear fuel transportation (Part II: Ship collision probability)

    International Nuclear Information System (INIS)

    Christian, Robby; Kang, Hyun Gook

    2017-01-01

    This paper proposes a methodology to assess and reduce risks of maritime spent nuclear fuel transportation with a probabilistic approach. Event trees detailing the progression of collisions leading to transport casks’ damage were constructed. Parallel and crossing collision probabilities were formulated based on the Poisson distribution. Automatic Identification System (AIS) data were processed with the Hough Transform algorithm to estimate possible intersections between the shipment route and the marine traffic. Monte Carlo simulations were done to compute collision probabilities and impact energies at each intersection. Possible safety improvement measures through a proper selection of operational transport parameters were investigated. These parameters include shipment routes, ship's cruise velocity, number of transport casks carried in a shipment, the casks’ stowage configuration and loading order on board the ship. A shipment case study is presented. Waters with high collision probabilities were identified. Effective range of cruising velocity to reduce collision risks were discovered. The number of casks in a shipment and their stowage method which gave low cask damage frequencies were obtained. The proposed methodology was successful in quantifying ship collision and cask damage frequency. It was effective in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • Proposes a probabilistic framework on the safety of spent nuclear fuel transportation by sea. • Developed a marine traffic simulation model using Generalized Hough Transform (GHT) algorithm. • A transportation case study on South Korean waters is presented. • Single-vessel risk reduction method is outlined by optimizing transport parameters.

  1. Expert estimation of human error probabilities in nuclear power plant operations: a review of probability assessment and scaling

    International Nuclear Information System (INIS)

    Stillwell, W.G.; Seaver, D.A.; Schwartz, J.P.

    1982-05-01

    This report reviews probability assessment and psychological scaling techniques that could be used to estimate human error probabilities (HEPs) in nuclear power plant operations. The techniques rely on expert opinion and can be used to estimate HEPs where data do not exist or are inadequate. These techniques have been used in various other contexts and have been shown to produce reasonably accurate probabilities. Some problems do exist, and limitations are discussed. Additional topics covered include methods for combining estimates from multiple experts, the effects of training on probability estimates, and some ideas on structuring the relationship between performance shaping factors and HEPs. Preliminary recommendations are provided along with cautions regarding the costs of implementing the recommendations. Additional research is required before definitive recommendations can be made

  2. Fire probability safety analysis in France for 900 MWe nuclear power plants

    International Nuclear Information System (INIS)

    Bertrand, R.; Bonneval, F.; Mattei, J.M.

    2000-01-01

    This paper describes the methodology implemented by the Institute for Nuclear Safety and Protection (IPSN) to carry out the Fire Probabilistic Safety Assessment (Fire PSA) for French 900 MWe pressurised water reactors. The initial results obtained are presented. Additional research and development activities are indicated which IPSN carried out or decided to perform in order to reduce the amount of uncertainty associated with the data or to confirm hypotheses that can impact significantly the study results. (orig.) [de

  3. Aging in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jordan Cizelj, R.; Kozuh, M.

    1995-01-01

    Aging is a phenomenon, which is influencing on unavailability of all components of the plant. The influence of aging on Probabilistic Safety Assessment calculations was estimated for Electrical Power Supply System. The average increase of system unavailability due to aging of system components was estimated and components were prioritized regarding their influence on change of system unavailability and relative increase of their unavailability due to aging. After the analysis of some numerical results, the recommendation for a detailed research of aging phenomena and its influence on system availability is given. (author)

  4. Waste isolation safety assessment program

    International Nuclear Information System (INIS)

    Brandstetter, A.; Harwell, M.A.

    1979-05-01

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

  5. Probabilistic safety assessment - regulatory perspective

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  6. Impact probabilities of meteoroid streams with artificial satellites: An assessment

    International Nuclear Information System (INIS)

    Foschini, L.; Cevolani, G.

    1997-01-01

    Impact probabilities of artificial satellites with meteoroid streams were calculated using data collected with the CNR forward scatter (FS) bistatic radar over the Bologna-Lecce baseline (about 700 km). Results show that impact probabilities are 2 times higher than other previously calculated values. Nevertheless, although catastrophic impacts are still rare even in the case of meteor storm conditions, it is expected that high meteoroid fluxes can erode satellites surfaces and weaken their external structures

  7. Risk-informed approaches to assess ecological safety of facilities with radioactive waste

    International Nuclear Information System (INIS)

    Vashchenko, V.N.; Zlochevskij, V.V.; Skalozubov, V.I.

    2011-01-01

    Ingenious risk-informed methods to assess ecological safety of facilities with radioactive waste are proposed in the paper. Probabilistic norms on lethal outcomes and reliability of safety barriers are used as safety criteria. Based on the probability measures, it is established that ecological safety conditions are met for the standard criterion of lethal outcomes

  8. Input-profile-based software failure probability quantification for safety signal generation systems

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Lim, Ho Gon; Lee, Ho Jung; Kim, Man Cheol; Jang, Seung Cheol

    2009-01-01

    The approaches for software failure probability estimation are mainly based on the results of testing. Test cases represent the inputs, which are encountered in an actual use. The test inputs for the safety-critical application such as a reactor protection system (RPS) of a nuclear power plant are the inputs which cause the activation of protective action such as a reactor trip. A digital system treats inputs from instrumentation sensors as discrete digital values by using an analog-to-digital converter. Input profile must be determined in consideration of these characteristics for effective software failure probability quantification. Another important characteristic of software testing is that we do not have to repeat the test for the same input value since the software response is deterministic for each specific digital input. With these considerations, we propose an effective software testing method for quantifying the failure probability. As an example application, the input profile of the digital RPS is developed based on the typical plant data. The proposed method in this study is expected to provide a simple but realistic mean to quantify the software failure probability based on input profile and system dynamics.

  9. The Human Bathtub: Safety and Risk Predictions Including the Dynamic Probability of Operator Errors

    International Nuclear Information System (INIS)

    Duffey, Romney B.; Saull, John W.

    2006-01-01

    Reactor safety and risk are dominated by the potential and major contribution for human error in the design, operation, control, management, regulation and maintenance of the plant, and hence to all accidents. Given the possibility of accidents and errors, now we need to determine the outcome (error) probability, or the chance of failure. Conventionally, reliability engineering is associated with the failure rate of components, or systems, or mechanisms, not of human beings in and interacting with a technological system. The probability of failure requires a prior knowledge of the total number of outcomes, which for any predictive purposes we do not know or have. Analysis of failure rates due to human error and the rate of learning allow a new determination of the dynamic human error rate in technological systems, consistent with and derived from the available world data. The basis for the analysis is the 'learning hypothesis' that humans learn from experience, and consequently the accumulated experience defines the failure rate. A new 'best' equation has been derived for the human error, outcome or failure rate, which allows for calculation and prediction of the probability of human error. We also provide comparisons to the empirical Weibull parameter fitting used in and by conventional reliability engineering and probabilistic safety analysis methods. These new analyses show that arbitrary Weibull fitting parameters and typical empirical hazard function techniques cannot be used to predict the dynamics of human errors and outcomes in the presence of learning. Comparisons of these new insights show agreement with human error data from the world's commercial airlines, the two shuttle failures, and from nuclear plant operator actions and transient control behavior observed in transients in both plants and simulators. The results demonstrate that the human error probability (HEP) is dynamic, and that it may be predicted using the learning hypothesis and the minimum

  10. OSART Independent Safety Culture Assessment (ISCA) Guidelines

    International Nuclear Information System (INIS)

    2016-01-01

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

  11. Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

    International Nuclear Information System (INIS)

    Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

    2009-01-01

    The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

  12. A framework to assess diagnosis error probabilities in the advanced MCR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ar Ryum; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of); Kim, Jong Hyun [Chosun University, Gwangju (Korea, Republic of); Jang, Inseok; Park, Jinkyun [Korea Atomic Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The Institute of Nuclear Power Operations (INPO)’s operating experience database revealed that about 48% of the total events in world NPPs for 2 years (2010-2011) happened due to human errors. The purposes of human reliability analysis (HRA) method are to evaluate the potential for, and mechanism of, human errors that may affect plant safety. Accordingly, various HRA methods have been developed such as technique for human error rate prediction (THERP), simplified plant analysis risk human reliability assessment (SPAR-H), cognitive reliability and error analysis method (CREAM) and so on. Many researchers have asserted that procedure, alarm, and display are critical factors to affect operators’ generic activities, especially for diagnosis activities. None of various HRA methods was explicitly designed to deal with digital systems. SCHEME (Soft Control Human error Evaluation MEthod) considers only for the probability of soft control execution error in the advanced MCR. The necessity of developing HRA methods in various conditions of NPPs has been raised. In this research, the framework to estimate diagnosis error probabilities in the advanced MCR was suggested. The assessment framework was suggested by three steps. The first step is to investigate diagnosis errors and calculate their probabilities. The second step is to quantitatively estimate PSFs’ weightings in the advanced MCR. The third step is to suggest the updated TRC model to assess the nominal diagnosis error probabilities. Additionally, the proposed framework was applied by using the full-scope simulation. Experiments conducted in domestic full-scope simulator and HAMMLAB were used as data-source. Total eighteen tasks were analyzed and twenty-three crews participated in.

  13. METHOD OF FOREST FIRES PROBABILITY ASSESSMENT WITH POISSON LAW

    Directory of Open Access Journals (Sweden)

    A. S. Plotnikova

    2016-01-01

    Full Text Available The article describes the method for the forest fire burn probability estimation on a base of Poisson distribution. The λ parameter is assumed to be a mean daily number of fires detected for each Forest Fire Danger Index class within specific period of time. Thus, λ was calculated for spring, summer and autumn seasons separately. Multi-annual daily Forest Fire Danger Index values together with EO-derived hot spot map were input data for the statistical analysis. The major result of the study is generation of the database on forest fire burn probability. Results were validated against EO daily data on forest fires detected over Irkutsk oblast in 2013. Daily weighted average probability was shown to be linked with the daily number of detected forest fires. Meanwhile, there was found a number of fires which were developed when estimated probability was low. The possible explanation of this phenomenon was provided.

  14. Probabilistic safety assessment for seismic events

    International Nuclear Information System (INIS)

    1993-10-01

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

  15. Consideration of aging in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Titina, B.; Cepin, M.

    2007-01-01

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

  16. Safety assessment principles for nuclear plants

    International Nuclear Information System (INIS)

    1992-01-01

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

  17. Radiological safety and risk assessment

    International Nuclear Information System (INIS)

    Hunter, P.H.; Barg, D.C.; Baird, R.D.; Card, D.H.; de Souza, F.; Elder, J.; Felthauser, K.; Jensen, C.; Winkler, V.

    1982-02-01

    A brief radiological safety and risk assessment of a nuclear power generation center with an adjacent on-site waste disposal facility at a specific site in the State of Utah is presented. The assessment was conducted to assist in determining the feasibility and practicality of developing a nuclear energy center (NEC) in Utah consisting of nine 1250 MWe nuclear pressurized water reactor (PWR) electrical generating units arranged in 3 clusters of 3 units each known as triads. The site selected for this conceptual study is in the Horse Bench area about 15 miles directly south of the town of Green River, Utah. The radiological issues included direct radiation exposures to on-site workers and the off-site population, release of radioactive material, and effects of these releases for both normal operations and accidental occurrences. The basic finding of this study is that the concept of an NEC in the Green River area, specifically at the Horse Bench site, is radiologically feasible

  18. Model uncertainty in safety assessment

    International Nuclear Information System (INIS)

    Pulkkinen, U.; Huovinen, T.

    1996-01-01

    The uncertainty analyses are an essential part of any risk assessment. Usually the uncertainties of reliability model parameter values are described by probability distributions and the uncertainty is propagated through the whole risk model. In addition to the parameter uncertainties, the assumptions behind the risk models may be based on insufficient experimental observations and the models themselves may not be exact descriptions of the phenomena under analysis. The description and quantification of this type of uncertainty, model uncertainty, is the topic of this report. The model uncertainty is characterized and some approaches to model and quantify it are discussed. The emphasis is on so called mixture models, which have been applied in PSAs. Some of the possible disadvantages of the mixture model are addressed. In addition to quantitative analyses, also qualitative analysis is discussed shortly. To illustrate the models, two simple case studies on failure intensity and human error modeling are described. In both examples, the analysis is based on simple mixture models, which are observed to apply in PSA analyses. (orig.) (36 refs., 6 figs., 2 tabs.)

  19. Model uncertainty in safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Pulkkinen, U; Huovinen, T [VTT Automation, Espoo (Finland). Industrial Automation

    1996-01-01

    The uncertainty analyses are an essential part of any risk assessment. Usually the uncertainties of reliability model parameter values are described by probability distributions and the uncertainty is propagated through the whole risk model. In addition to the parameter uncertainties, the assumptions behind the risk models may be based on insufficient experimental observations and the models themselves may not be exact descriptions of the phenomena under analysis. The description and quantification of this type of uncertainty, model uncertainty, is the topic of this report. The model uncertainty is characterized and some approaches to model and quantify it are discussed. The emphasis is on so called mixture models, which have been applied in PSAs. Some of the possible disadvantages of the mixture model are addressed. In addition to quantitative analyses, also qualitative analysis is discussed shortly. To illustrate the models, two simple case studies on failure intensity and human error modeling are described. In both examples, the analysis is based on simple mixture models, which are observed to apply in PSA analyses. (orig.) (36 refs., 6 figs., 2 tabs.).

  20. Development of safety related technology and infrastructure for safety assessment

    International Nuclear Information System (INIS)

    Venkat Raj, V.

    1997-01-01

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

  1. Thinking of the safety assessment of HLW disposal

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Efficacy, safety and tolerability of rivastigmine capsules in patients with probable vascular dementia: the VantagE study.

    Science.gov (United States)

    Ballard, C; Sauter, M; Scheltens, P; He, Y; Barkhof, F; van Straaten, E C W; van der Flier, W M; Hsu, C; Wu, S; Lane, R

    2008-09-01

    The aim was to evaluate the efficacy, safety and tolerability of rivastigmine capsules in patients diagnosed with probable vascular dementia (VaD). VantagE (Vascular Dementia trial studying Exelon) was a 24-week, multicentre, double-blind study. VaD patients aged 50-85 years were randomized to rivastigmine capsules (3-12 mg/day) or placebo. Efficacy assessments included global and cognitive performances, activities of daily living and neuropsychiatric symptoms. Adverse events were recorded. Additional exploratory analyses determined whether heterogeneity in pathologies and symptoms extended to differential treatment effects. NCT00099216. 710 patients were randomized. Rivastigmine demonstrated superiority over placebo on three measures of cognitive performance (Vascular Dementia Assessment Scale, Alzheimer's Disease Assessment Scale cognitive subscale, Mini-Mental State Examination; all p or =75 years old), assumed more likely to also have Alzheimer's disease (AD) pathology, demonstrated significant cognitive responses to rivastigmine and a safety profile similar to that seen in AD patients. Younger patients, assumed less likely to have concomitant AD pathology, showed no efficacy response and were associated with slight elevations of blood pressure, cerebrovascular accidents and mortality. Rivastigmine-placebo differences in patients with, versus those without, medial temporal atrophy (also suggestive of concomitant AD) showed a numerical difference similar to that seen between the older versus younger patients, but did not attain statistical significance. Consistent with trials evaluating other cholinesterase inhibitors, rivastigmine did not provide consistent efficacy in probable VaD. The efficacy apparent on cognitive outcomes was derived from effects in older patients likely to have concomitant Alzheimer pathology. This is supportive of an existing argument that the putative cholinergic deficit in VaD reflects the presence of concomitant Alzheimer pathology.

  3. Regulatory review of safety cases and safety assessments - associated challenges

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  4. Intermediate probabilistic safety assessment approach for safety critical digital systems

    International Nuclear Information System (INIS)

    Taeyong, Sung; Hyun Gook, Kang

    2001-01-01

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

  5. Nirex Safety Assessment Research Programme bibliography, 1990

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1990-10-01

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

  6. Study on probability distribution of fire scenarios in risk assessment to emergency evacuation

    International Nuclear Information System (INIS)

    Chu Guanquan; Wang Jinhui

    2012-01-01

    Event tree analysis (ETA) is a frequently-used technique to analyze the probability of probable fire scenario. The event probability is usually characterized by definite value. It is not appropriate to use definite value as these estimates may be the result of poor quality statistics and limited knowledge. Without addressing uncertainties, ETA will give imprecise results. The credibility of risk assessment will be undermined. This paper presents an approach to address event probability uncertainties and analyze probability distribution of probable fire scenario. ETA is performed to construct probable fire scenarios. The activation time of every event is characterized as stochastic variable by considering uncertainties of fire growth rate and other input variables. To obtain probability distribution of probable fire scenario, Markov Chain is proposed to combine with ETA. To demonstrate the approach, a case study is presented.

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

    OpenAIRE

    Nganje, William E.

    1999-01-01

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

  8. Impact of the specialization from failures data in probability safety analysis for process plants

    International Nuclear Information System (INIS)

    Ribeiro, Antonio C.O.; Melo, P.F. Frutuoso e

    2005-01-01

    Full text: The aim of this paper is to show the Bayesian inference in reliability studies, which are used to failures, rates updating in safety analyses. It is developed the impact of its using in quantitative risk assessments (QRA) for industrial process plants. With this approach we find a structured and auditable way of showing the difference between an industrial installation with a good project and maintenance structure from another one that shows a low level of quality in these areas. In general the evidence from failures rates and as follow the frequency of occurrence from scenarios, which the risks taken in account in ERA, are taken from generics data banks, instead of, the installation in analysis. The use of this methodology in probabilistic safety analysis (PSA) for nuclear plants is commonly used when you need to find the final fault tree event evaluation applied to a scenario, but it is not showed in a PSA level III. (author)

  9. Safety culture assessment developed by JANTI

    International Nuclear Information System (INIS)

    Hamada, Jun

    2009-01-01

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

  10. 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.

  11. 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

  12. Safety assessment for Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Leahy, T.J.

    2012-01-01

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

  13. NPP Krsko periodic safety review. Safety assessment and analyses

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  14. ALARP considerations in criticality safety assessments

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  15. Preliminary safety assessment of the WIPP facility

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  16. Safety assessments for deep geological disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Lyon, R.B.

    1984-01-01

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

  17. Safety Assessment - Swedish Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B. [Luleaa Univ. of Technology (Sweden)

    1996-12-31

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs.

  18. Safety Assessment - Swedish Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kjellstroem, B.

    1996-01-01

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs

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

    International Nuclear Information System (INIS)

    1992-01-01

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

  20. Application of Probability Calculations to the Study of the Permissible Step and Touch Potentials to Ensure Personnel Safety

    International Nuclear Information System (INIS)

    Eisawy, E.A.

    2011-01-01

    The aim of this paper is to develop a practical method to evaluate the actual step and touch potential distributions in order to determine the risk of failure of the grounding system. The failure probability, indicating the safety level of the grounding system, is related to both applied (stress) and withstand (strength) step or touch potentials. The probability distributions of the applied step and touch potentials as well as the corresponding withstand step and touch potentials which represent the capability of the human body to resist stress potentials are presented. These two distributions are used to evaluate the failure probability of the grounding system which denotes the probability that the applied potential exceeds the withstand potential. The method is accomplished in considering the resistance of the human body, the foot contact resistance and the fault clearing time as an independent random variables, rather than fixed values as treated in the previous analysis in determining the safety requirements for a given grounding system

  1. Complementary safety assessments - Report by the French Nuclear Safety Authority

    International Nuclear Information System (INIS)

    2011-12-01

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

  2. Food and feed safety assessment

    NARCIS (Netherlands)

    Kuiper, H.A.; Paoletti, Claudia

    2015-01-01

    The general principles for safety and nutritional evaluation of foods and feed and the potential health risks associated with hazardous compounds are described as developed by the Food and Agriculture Organization (FAO) and the World Health Organization (WHO) and further elaborated in the

  3. Assessment of basic safety issues

    International Nuclear Information System (INIS)

    Queniart, D.

    1996-01-01

    Work on the French-German common safety approach for future nuclear power plants continued in 1994 to allow for more detailed discussion of some major issues, taking into account the options provided by the industry for the EPR (European Pressurized water Reactor) project, as described in the document entitled 'Conceptual Safety Features Review File'. Seven meetings of a GPR/RSK advisory experts subgroup, six GPR/RSK plenary sessions and six meetings of the safety authorities (DFD) dealt with the following topics: design of the systems and use of probabilistic approaches, application of a 'break preclusion' approach to the main primary pipings, protection against external hazards (aircraft crashes, explosions, earthquakes), provisions with respect to accidents involving core melt and to containment design, radiological consequences of reference accidents and accidents involving core melt at low pressure. The important aspects of the joint policy are recalled in the presentation. The whole set of GPR/RSK recommendations were agreed by the French and German safety authorities during the DFD meetings of 1994 and early 1995. The utilities decided to begin the basic design phase in February, 1995. Work is now continuing to develop the common French-German approach for future nuclear power plants, in the same way as before. In 1995, this mainly covers the design of the containment and of the systems, but also new issues such as the protection against secondary side overpressurization, radiological protection of workers and radioactive wastes. (J.S.). 3 figs., 1 tab

  4. Independent assessment for new nuclear reactor safety

    Directory of Open Access Journals (Sweden)

    D'Auria Francesco

    2017-01-01

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

  5. Independent assessment for new nuclear reactor safety

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  6. Developing IAM for Life Cycle Safety Assessment

    NARCIS (Netherlands)

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

    2013-01-01

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

  7. Safety factors for neutron fluences in NPP safety assessment

    International Nuclear Information System (INIS)

    Demekhin, V.L.; Bukanov, V.N.; Il'kovich, V.V.; Pugach, A.M.

    2016-01-01

    In accordance with global practice and a number of existing regulations, the use of conservative approach is required for the calculations related to nuclear safety assessment of NPP. It implies the need to consider the determination of neutron fluence errors that is rather complicated. It is proposed to carry out the consideration by the way of multiplying the neutron fluences obtained with transport calculations by safety factors. The safety factor values are calculated by the developed technique based on the theory of errors, features of the neutron transport calculation code and the results obtained with the code. It is shown that the safety factor value is equal 1.18 with the confidence level of not less than 0.95 for the majority of VVER-1000 reactor places where neutron fluences are determined by MCPV code, and its maximum value is 1.25

  8. Assessment of safety culture at INPP

    International Nuclear Information System (INIS)

    Lesin, S.

    2002-01-01

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

  9. Assessment of elementary school safety restraint programs.

    Science.gov (United States)

    1985-06-01

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

  10. Probability-based assessment and maintenance of welded joints damaged by fatigue

    International Nuclear Information System (INIS)

    Cremona, C.; Lukic, M.

    1998-01-01

    This paper presents a probabilistic reliability assessment procedure for steel components damaged by fatigue. The study combines the structural reliability theory with a maintenance strategy. The fatigue assessment model is based on a modelisation of the fatigue phenomenon issued from the principles of fracture mechanics theory. The safety margin includes the crack growth propagation and allows to treat fatigue damage in a general manner. Damaging cycles and non damaging cycles are distinguished. The sensitivity study of the different parameters shows that some variables can be taken as deterministic. Applications are made on a welded joint 'stiffener/bottom-plate' of a typical steel bridge. The model is then used for taking into account inspection results. Non destructive inspection (NDI) techniques are also used for updating failure probabilities. The results show their ability to be inserted in a maintenance strategy for optimizing the next inspection time. This has led to define cost functions related to the total maintenance cost; this cost is then minimized for determining the optimal next inspection time. An example of welded joint cracked by fatigue highlights the different concepts. The approach presented in the paper is not only restrained to fatigue problems, but can be applied to a wide variety of degrading phenomena. (orig.)

  11. Assessment of the probability of an aircraft accidentally crashing on a nuclear power station

    International Nuclear Information System (INIS)

    Gravero; Lezer; Lucenet

    1975-01-01

    The probability of an accidental aircraft crash on a power station not situated near a commercial airport is assessed. Three major points in the general analysis of the problem are developed: analysis of accidents as a function of the phase of the flight and in particular during a flight in transit and examin-ation of aircraft crash conditions to determine the angle of impact on the reactor building for example; determination of the apparent surface of buildings allowing for several parameters: geometry of the building and of the aircraft, geography of the site, relative position of the buildings; assessment of air traffic above the region for the year under consideration distinguishing the weight of the aircraft which implies an investigation of the problem for commercial aviation on the one hand (regular or irregular flights, inter-national or internal) and for general aviation on the other hand. The analysis is determined for the years 1980 - 2000 so that ir will be necessary to extrapolate some of the parameters (development of air traffic, safety of transport, etc). (author)

  12. Estimation of Partial Safety Factors and Target Failure Probability Based on Cost Optimization of Rubble Mound Breakwaters

    DEFF Research Database (Denmark)

    Kim, Seung-Woo; Suh, Kyung-Duck; Burcharth, Hans F.

    2010-01-01

    The breakwaters are designed by considering the cost optimization because a human risk is seldom considered. Most breakwaters, however, were constructed without considering the cost optimization. In this study, the optimum return period, target failure probability and the partial safety factors...

  13. Developing a Questionnaire to Assess the Probability Content Knowledge of Prospective Primary School Teachers

    Science.gov (United States)

    Gómez-Torres, Emilse; Batanero, Carmen; Díaz, Carmen; Contreras, José Miguel

    2016-01-01

    In this paper we describe the development of a questionnaire designed to assess the probability content knowledge of prospective primary school teachers. Three components of mathematical knowledge for teaching and three different meanings of probability (classical, frequentist and subjective) are considered. The questionnaire content is based on…

  14. Assessment procedure and probability determination methods of aircraft crash events in siting for nuclear power plants

    International Nuclear Information System (INIS)

    Zheng Qiyan; Zhang Lijun; Huang Weiqi; Yin Qingliao

    2010-01-01

    Assessment procedure of aircraft crash events in siting for nuclear power plants, and the methods of probability determination in two different stages of prelimi- nary screening and detailed evaluation are introduced in this paper. Except for general air traffic, airport operations and aircraft in the corridor, the probability of aircraft crash by military operation in the military airspaces is considered here. (authors)

  15. Building a Model Using Bayesian Network for Assessment of Posterior Probabilities of Falling From Height at Workplaces

    Directory of Open Access Journals (Sweden)

    Seyed Shamseddin Alizadeh

    2014-12-01

    Full Text Available Background: Falls from height are one of the main causes of fatal occupational injuries. The objective of this study was to present a model for estimating occurrence probability of falling from height. Methods: In order to make a list of factors affecting falls, we used four expert group's judgment, literature review and an available database. Then the validity and reliability of designed questionnaire were determined and Bayesian networks were built. The built network, nodes and curves were quantified. For network sensitivity analysis, four types of analysis carried out. Results: A Bayesian network for assessment of posterior probabilities of falling from height proposed. The presented Bayesian network model shows the interrelationships among 37 causes affecting the falling from height and can calculate its posterior probabilities. The most important factors affecting falling were Non-compliance with safety instructions for work at height (0.127, Lack of safety equipment for work at height (0.094 and Lack of safety instructions for work at height (0.071 respectively. Conclusion: The proposed Bayesian network used to determine how different causes could affect the falling from height at work. The findings of this study can be used to decide on the falling accident prevention programs.

  16. HSE assessment of explosion risk analysis in offshore safety cases

    Energy Technology Data Exchange (ETDEWEB)

    Brighton, P.W.M.; Fearnley, P.J.; Brearley, I.G. [Health and Safety Executive, Bootle (United Kingdom). Offshore Safety Div.

    1995-12-31

    In the past two years HSE has assessed around 250 Safety Cases for offshore oil and gas installations, building up a unique overview of the current state of the art on fire and explosion risk assessment. This paper reviews the explosion risk methods employed, focusing on the aspects causing most difficulty for assessment and acceptance of Safety Cases. Prediction of overpressures in offshore explosions has been intensively researched in recent years but the justification of the means of prevention, control and mitigation of explosions often depends on much additional analysis of the frequency and damage potential of explosions. This involves a number of factors, the five usually considered being: leak sizes; gas dispersion; ignition probabilities; the frequency distribution of explosion strength; and the prediction of explosion damage. Sources of major uncertainty in these factors and their implications for practical risk management decisions are discussed. (author)

  17. Probabilistic safety assessment in radioactive waste disposal

    International Nuclear Information System (INIS)

    Robinson, P.C.

    1987-07-01

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

  18. Safety assessment for radioactive waste disposal facility

    International Nuclear Information System (INIS)

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

    2008-08-01

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

  19. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

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

  20. Fire safety assessment of tunnel structures

    DEFF Research Database (Denmark)

    Gkoumas, Konstantinos; Giuliani, Luisa; Petrini, Francesco

    2011-01-01

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

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    EVANS, C B

    2004-12-21

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

  3. Probabilistic safety assessment goals in Canada

    International Nuclear Information System (INIS)

    Snell, V.G.

    1986-01-01

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

  4. Probabilistic safety assessment of the Fugen NPS

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  5. Type I error probability spending for post-market drug and vaccine safety surveillance with binomial data.

    Science.gov (United States)

    Silva, Ivair R

    2018-01-15

    Type I error probability spending functions are commonly used for designing sequential analysis of binomial data in clinical trials, but it is also quickly emerging for near-continuous sequential analysis of post-market drug and vaccine safety surveillance. It is well known that, for clinical trials, when the null hypothesis is not rejected, it is still important to minimize the sample size. Unlike in post-market drug and vaccine safety surveillance, that is not important. In post-market safety surveillance, specially when the surveillance involves identification of potential signals, the meaningful statistical performance measure to be minimized is the expected sample size when the null hypothesis is rejected. The present paper shows that, instead of the convex Type I error spending shape conventionally used in clinical trials, a concave shape is more indicated for post-market drug and vaccine safety surveillance. This is shown for both, continuous and group sequential analysis. Copyright © 2017 John Wiley & Sons, Ltd.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  7. Stockholm Safety Conference. Analysis of the sessions on radiological protection, licensing and risk assessment

    International Nuclear Information System (INIS)

    Gea, A.

    1981-01-01

    A summary of the sessions on radiological protection, licensing and risk assessment in the safety conference of Stockholm is presented. It is considered the new point of view of the nuclear safety, probabilistic analysis, components failures probability and accident analysis. They are included conclusions applicable in many cases to development countries. (author)

  8. Economic aspects of risk assessment in chemical safety

    Energy Technology Data Exchange (ETDEWEB)

    Drummond, M F; Shannon, H S

    1986-05-01

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

  9. Probabilistic Safety Assessment of Waste from PyroGreen Processes

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. Probabilistic Safety Assessment of Waste from PyroGreen Processes

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  11. Quantitative risk assessment of digitalized safety systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  12. Environment, safety and health progress assessment manual

    International Nuclear Information System (INIS)

    1992-12-01

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

  13. RSAS: a Reactor Safety Assessment System

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  14. 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

  15. Safety assessment for the passive system of the nuclear power plants (NPPs) using safety margin estimation

    International Nuclear Information System (INIS)

    Woo, Tae-Ho; Lee, Un-Chul

    2010-01-01

    The probabilistic safety assessment (PSA) for gas-cooled nuclear power plants has been investigated where the operational data are deficient, because there is not any commercial gas-cooled nuclear power plant. Therefore, it is necessary to use the statistical data for the basic event constructions. Several estimations for the safety margin are introduced for the quantification of the failure frequency in the basic event, which is made by the concept of the impact and affordability. Trend of probability of failure (TPF) and fuzzy converter (FC) are introduced using the safety margin, which shows the simplified and easy configurations for the event characteristics. The mass flow rate in the natural circulation is studied for the modeling. The potential energy in the gravity, the temperature and pressure in the heat conduction, and the heat transfer rate in the internal stored energy are also investigated. The values in the probability set are compared with those of the fuzzy set modeling. Non-linearity of the safety margin is expressed by the fuzziness of the membership function. This artificial intelligence analysis of the fuzzy set could enhance the reliability of the system comparing to the probabilistic analysis.

  16. Safety assessment of HLW geological disposal system

    International Nuclear Information System (INIS)

    Naito, Morimasa

    2006-01-01

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

  17. The DYLAM approach to systems safety and reliability assessment

    International Nuclear Information System (INIS)

    Amendola, A.

    1988-01-01

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

  18. Environment, safety and health progress assessment manual

    International Nuclear Information System (INIS)

    1992-12-01

    On June 27, 1989, the Secretary of Energy announced a 10-Point Initiative to strengthen environment, safety, and health (ES ampersand H) programs, and waste management activities at DOE production, research, and testing facilities. One of the points involved conducting dent Tiger Team Assessments of DOE operating facilities. The Office of Special independent Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This manual documents the processes to be used to perform the ES ampersand H Progress Assessments. It was developed based upon the lessons learned from Tiger Team Assessments, the two pilot Progress Assessments, and Progress Assessments that have been completed. The manual will be updated periodically to reflect lessons learned or changes in policy

  19. Probabilistic safety assessment for food irradiation facility

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  20. Earthquake Probability Assessment for the Active Faults in Central Taiwan: A Case Study

    Directory of Open Access Journals (Sweden)

    Yi-Rui Lee

    2016-06-01

    Full Text Available Frequent high seismic activities occur in Taiwan due to fast plate motions. According to the historical records the most destructive earthquakes in Taiwan were caused mainly by inland active faults. The Central Geological Survey (CGS of Taiwan has published active fault maps in Taiwan since 1998. There are 33 active faults noted in the 2012 active fault map. After the Chi-Chi earthquake, CGS launched a series of projects to investigate the details to better understand each active fault in Taiwan. This article collected this data to develop active fault parameters and referred to certain experiences from Japan and the United States to establish a methodology for earthquake probability assessment via active faults. We consider the active faults in Central Taiwan as a good example to present the earthquake probability assessment process and results. The appropriate “probability model” was used to estimate the conditional probability where M ≥ 6.5 and M ≥ 7.0 earthquakes. Our result shows that the highest earthquake probability for M ≥ 6.5 earthquake occurring in 30, 50, and 100 years in Central Taiwan is the Tachia-Changhua fault system. Conversely, the lowest earthquake probability is the Chelungpu fault. The goal of our research is to calculate the earthquake probability of the 33 active faults in Taiwan. The active fault parameters are important information that can be applied in the following seismic hazard analysis and seismic simulation.

  1. Algorithms for the extension of precise and imprecise conditional probability assessments: an implementation with maple V

    Directory of Open Access Journals (Sweden)

    Veronica Biazzo

    2000-05-01

    Full Text Available In this paper, we illustrate an implementation with Maple V of some procedures which allow to exactly propagate precise and imprecise probability assessments. The extension of imprecise assessments is based on a suitable generalization of the concept of coherence of de Finetti. The procedures described are supported by some examples and relevant cases.

  2. On the assessment of extremely low breakdown probabilities by an inverse sampling procedure [gaseous insulation

    DEFF Research Database (Denmark)

    Thyregod, Poul; Vibholm, Svend

    1991-01-01

    the flashover probability function and the corresponding distribution of first breakdown voltages under the inverse sampling procedure, and show how this relation may be utilized to assess the single-shot flashover probability corresponding to the observed average first breakdown voltage. Since the procedure......First breakdown voltages obtained under the inverse sampling procedure assuming a double exponential flashover probability function are discussed. An inverse sampling procedure commences the voltage application at a very low level, followed by applications at stepwise increased levels until...... is based on voltage applications in the neighbourhood of the quantile under investigation, the procedure is found to be insensitive to the underlying distributional assumptions...

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. Human reliability analysis methods for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-11-01

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

  5. Safety analysis and risk assessment handbook

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  6. Safety assessment of plant food supplements (PFS)

    NARCIS (Netherlands)

    Berg, van den S.J.P.L.; Serra-Majem, L.; Coppens, P.; Rietjens, I.

    2011-01-01

    Botanicals and botanical preparations, including plant food supplements (PFS), are widely used in Western diets. The growing use of PFS is accompanied by an increasing concern because the safety of these PFS is not generally assessed before they enter the market. Regulatory bodies have become more

  7. Safety assessment for Dragon fuel element production

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1963-11-01

    This report shall be the Safety Assessment covering the manufacture of the First Charge of Fuel and Fuel Elements for the Dragon Reactor Experiment. It is issued in two parts, of which Part I is descriptive and Part II gives the Hazards Analysis, the Operating Limitations, the Standing Orders and the Emergency Drill. (author)

  8. Safety assessment of genetically modified foods

    NARCIS (Netherlands)

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

    2016-01-01

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

  9. Safety assessment of high consequence robotics system

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

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

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

    International Nuclear Information System (INIS)

    2011-01-01

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

  12. Assessing the present and future probability of Hurricane Harvey’s rainfall

    OpenAIRE

    Emanuel, Kerry

    2017-01-01

    Significance Natural disasters such as the recent Hurricanes Harvey, Irma, and Maria highlight the need for quantitative estimates of the risk of such disasters. Statistically based risk assessment suffers from short records of often poor quality, and in the case of meteorological hazards, from the fact that the underlying climate is changing. This study shows how a recently developed physics-based risk assessment method can be applied to assessing the probabilities of extreme hurricane rainf...

  13. Medicine in words and numbers: a cross-sectional survey comparing probability assessment scales

    Directory of Open Access Journals (Sweden)

    Koele Pieter

    2007-06-01

    Full Text Available Abstract Background In the complex domain of medical decision making, reasoning under uncertainty can benefit from supporting tools. Automated decision support tools often build upon mathematical models, such as Bayesian networks. These networks require probabilities which often have to be assessed by experts in the domain of application. Probability response scales can be used to support the assessment process. We compare assessments obtained with different types of response scale. Methods General practitioners (GPs gave assessments on and preferences for three different probability response scales: a numerical scale, a scale with only verbal labels, and a combined verbal-numerical scale we had designed ourselves. Standard analyses of variance were performed. Results No differences in assessments over the three response scales were found. Preferences for type of scale differed: the less experienced GPs preferred the verbal scale, the most experienced preferred the numerical scale, with the groups in between having a preference for the combined verbal-numerical scale. Conclusion We conclude that all three response scales are equally suitable for supporting probability assessment. The combined verbal-numerical scale is a good choice for aiding the process, since it offers numerical labels to those who prefer numbers and verbal labels to those who prefer words, and accommodates both more and less experienced professionals.

  14. Safety assessment of genetically modified crops

    International Nuclear Information System (INIS)

    Atherton, Keith T.

    2002-01-01

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

  15. MAPLE-X10 reactor safety assessment

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. Safety assessment of a lithium target

    International Nuclear Information System (INIS)

    Burgazzi, Luciano; Roberta, Ferri; Barbara, Giannone

    2006-01-01

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

  17. Using the Reliability Theory for Assessing the Decision Confidence Probability for Comparative Life Cycle Assessments.

    Science.gov (United States)

    Wei, Wei; Larrey-Lassalle, Pyrène; Faure, Thierry; Dumoulin, Nicolas; Roux, Philippe; Mathias, Jean-Denis

    2016-03-01

    Comparative decision making process is widely used to identify which option (system, product, service, etc.) has smaller environmental footprints and for providing recommendations that help stakeholders take future decisions. However, the uncertainty problem complicates the comparison and the decision making. Probability-based decision support in LCA is a way to help stakeholders in their decision-making process. It calculates the decision confidence probability which expresses the probability of a option to have a smaller environmental impact than the one of another option. Here we apply the reliability theory to approximate the decision confidence probability. We compare the traditional Monte Carlo method with a reliability method called FORM method. The Monte Carlo method needs high computational time to calculate the decision confidence probability. The FORM method enables us to approximate the decision confidence probability with fewer simulations than the Monte Carlo method by approximating the response surface. Moreover, the FORM method calculates the associated importance factors that correspond to a sensitivity analysis in relation to the probability. The importance factors allow stakeholders to determine which factors influence their decision. Our results clearly show that the reliability method provides additional useful information to stakeholders as well as it reduces the computational time.

  18. Human reliability in probabilistic safety assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

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

  19. Human Reliability in Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

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

  20. Probability Assessment and Risk Management of Progressive Collapse in Strategic Buildings Facing Blast Loads

    Directory of Open Access Journals (Sweden)

    Gholamreza Abdollahzadeh

    2016-12-01

    Full Text Available Nowadays, as a result of increased terrorist and bomb attacks throughout the globe in the vicinity of strategic buildings, designing these structures against impact loads, particularly the blast-related ones, has been taken into more consideration. The current procedure for designing the structure against an explosion is a design against the local failure of the current elements in the first step and then, in the next step, against local damage as well as tactful thinking to prevent this damage from spreading to other parts of the structure. The present research investigates the impacts of explosives, derived from probable terror–stricken scenarios inside and outside a strategic four-story steel building with a special moment frame system. Then, the resistive capacity of the damaged building (due to blast has been evaluated against the progressive collapse, and finally, the rate of the collapse risk and the reliability of the structure have been obtained by presenting a probable method. Thus, the vulnerable parts inside and outside the building are identified and safety measures have been determined, so that in case of no safety or excessive collapse risk- access to dangerous parts of the building could be reinforced or limited. Results show that progressive collapse probability and reliability of the building are 57% and 43% respectively.

  1. Probabilistic assessment of nuclear safety and safeguards

    International Nuclear Information System (INIS)

    Higson, D.J.

    1987-01-01

    Nuclear reactor accidents and diversions of materials from the nuclear fuel cycle are perceived by many people as particularly serious threats to society. Probabilistic assessment is a rational approach to the evaluation of both threats, and may provide a basis for decisions on appropriate actions to control them. Probabilistic method have become standard tools used in the analysis of safety, but there are disagreements on the criteria to be applied when assessing the results of analysis. Probabilistic analysis and assessment of the effectiveness of nuclear material safeguards are still at an early stage of development. (author)

  2. Using driving simulators to assess driving safety.

    Science.gov (United States)

    Boyle, Linda Ng; Lee, John D

    2010-05-01

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

  3. Ultraviolet safety assessments of insect light traps.

    Science.gov (United States)

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

    2016-01-01

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

  4. Electronuclear's safety culture assessment and enhancement program

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. Reliability assessment and probability based design of reinforced concrete containments and shear walls

    International Nuclear Information System (INIS)

    Hwang, H.; Reich, M.; Ellingwood, B.; Shinozuka, M.

    1986-03-01

    This report summarizes work completed under the program entitled, ''Probability-Based Load Combinations for Design of Category I Structures.'' Under this program, the probabilistic models for various static and dynamic loads were formulated. The randomness and uncertainties in material strengths and structural resistance were established. Several limit states of concrete containments and shear walls were identified and analytically formulated. Furthermore, the reliability analysis methods for estimating limit state probabilities were established. These reliability analysis methods can be used to evaluate the safety levels of nuclear structures under various combinations of static and dynamic loads. They can also be used to generate analytically the fragility data for PRA studies. In addition to the development of reliability analysis methods, probability-based design criteria for concrete containments and shear wall structures have also been developed. The proposed design criteria are in the load and resistance factor design (LRFD) format. The load and resistance factors are determined for several limit states and target limit state probabilities. Thus, the proposed design criteria are risk-consistent and have a well-established rationale. 73 refs., 18 figs., 16 tabs

  6. 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)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. Assessing the probability of carbon and greenhouse gas benefit from the management of peat soils

    International Nuclear Information System (INIS)

    Worrall, F.; Bell, M.J.; Bhogal, A.

    2010-01-01

    This study proposes a method for assessing the probability that land management interventions will lead to an improvement in the carbon sink represented by peat soils. The method is able to: combine studies of different carbon uptake and release pathways in order to assess changes on the overall carbon or greenhouse gas budget; calculate the probability of the management or restoration leading to an improvement in the budget; calculate the uncertainty in that probability estimate; estimate the equivalent number of complete budgets available from the combination of the literature; test the difference in the outcome of different land management interventions; and provide a method for updating the predicted probabilities as new studies become available. Using this methodology, this study considered the impact of: afforestation, managed burning, drainage, drain-blocking, grazing removal; and revegetation, on the carbon budget of peat soils in the UK. The study showed that afforestation, drain-blocking, revegetation, grazing removal and cessation of managed burning would bring a carbon benefit, whereas deforestation, managed burning and drainage would bring a disbenefit. The predicted probabilities of a benefit are often equivocal as each management type or restoration often leads to increase in uptake in one pathway while increasing losses in another.

  9. Assessing the Probability that a Finding Is Genuine for Large-Scale Genetic Association Studies.

    Science.gov (United States)

    Kuo, Chia-Ling; Vsevolozhskaya, Olga A; Zaykin, Dmitri V

    2015-01-01

    Genetic association studies routinely involve massive numbers of statistical tests accompanied by P-values. Whole genome sequencing technologies increased the potential number of tested variants to tens of millions. The more tests are performed, the smaller P-value is required to be deemed significant. However, a small P-value is not equivalent to small chances of a spurious finding and significance thresholds may fail to serve as efficient filters against false results. While the Bayesian approach can provide a direct assessment of the probability that a finding is spurious, its adoption in association studies has been slow, due in part to the ubiquity of P-values and the automated way they are, as a rule, produced by software packages. Attempts to design simple ways to convert an association P-value into the probability that a finding is spurious have been met with difficulties. The False Positive Report Probability (FPRP) method has gained increasing popularity. However, FPRP is not designed to estimate the probability for a particular finding, because it is defined for an entire region of hypothetical findings with P-values at least as small as the one observed for that finding. Here we propose a method that lets researchers extract probability that a finding is spurious directly from a P-value. Considering the counterpart of that probability, we term this method POFIG: the Probability that a Finding is Genuine. Our approach shares FPRP's simplicity, but gives a valid probability that a finding is spurious given a P-value. In addition to straightforward interpretation, POFIG has desirable statistical properties. The POFIG average across a set of tentative associations provides an estimated proportion of false discoveries in that set. POFIGs are easily combined across studies and are immune to multiple testing and selection bias. We illustrate an application of POFIG method via analysis of GWAS associations with Crohn's disease.

  10. Probability evaluation in problems of structural safety and quality assurance through oriented simulation

    International Nuclear Information System (INIS)

    Puppo, A.H.

    1990-01-01

    The procedure described in this work may be considered to improve the Harbitz method by a limitation of the directions corresponding to the sampling points. By the introduction of polar coordinates n-dimensional system, these directions may be limited to those contained in an hypercone, whose axis contains the design point. Some problems with strongly convex failure domain presented in some structural systems (Igusa and Der Kiureghian, 1988; Nadim, 1990) and also in the acceptance probability of analysis and rejection of quality assurance methods may be efficiently treated. (Author) [es

  11. Animal-Free Chemical Safety Assessment

    Directory of Open Access Journals (Sweden)

    George D Loizou

    2016-07-01

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

  12. The radiation safety self-assessment program of Ontario Hydro

    International Nuclear Information System (INIS)

    Armitage, G.; Chase, W.J.

    1987-01-01

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

  13. Assessment of safety culture: Changing regulatory approach in Hungary

    International Nuclear Information System (INIS)

    Ronaky, Jozsef; Toth, Andras

    2002-01-01

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

  14. Comparision of the different probability distributions for earthquake hazard assessment in the North Anatolian Fault Zone

    Energy Technology Data Exchange (ETDEWEB)

    Yilmaz, Şeyda, E-mail: seydayilmaz@ktu.edu.tr; Bayrak, Erdem, E-mail: erdmbyrk@gmail.com [Karadeniz Technical University, Trabzon (Turkey); Bayrak, Yusuf, E-mail: bayrak@ktu.edu.tr [Ağrı İbrahim Çeçen University, Ağrı (Turkey)

    2016-04-18

    In this study we examined and compared the three different probabilistic distribution methods for determining the best suitable model in probabilistic assessment of earthquake hazards. We analyzed a reliable homogeneous earthquake catalogue between a time period 1900-2015 for magnitude M ≥ 6.0 and estimated the probabilistic seismic hazard in the North Anatolian Fault zone (39°-41° N 30°-40° E) using three distribution methods namely Weibull distribution, Frechet distribution and three-parameter Weibull distribution. The distribution parameters suitability was evaluated Kolmogorov-Smirnov (K-S) goodness-of-fit test. We also compared the estimated cumulative probability and the conditional probabilities of occurrence of earthquakes for different elapsed time using these three distribution methods. We used Easyfit and Matlab software to calculate these distribution parameters and plotted the conditional probability curves. We concluded that the Weibull distribution method was the most suitable than other distribution methods in this region.

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

    International Nuclear Information System (INIS)

    Song, Tae Young

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  17. Assessment of aircraft impact probabilities at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Lee, L.G.; Mines, J.M.; Webb, B.B.

    1996-04-01

    The purpose of this study is to evaluate the possibility of an aircraft crash into a facility at the Idaho Chemical Processing Plant (ICPP). The ICPP is part of the Idaho National Engineering Laboratory (INEL). Based on the data used in this study, an air crash into any single facility at the ICPP is incredible, An air crash into aggregate areas incorporating the following is extremely unlikely: (1) ICPP radiological materials storage facilities, (2) ICPP major processing facilities, and (3) the ICPP total surface area. The radiological materials storage facilities aggregate areas are areas of concern usually requiring safety analyses, According to Department of Energy guidance, if the probability of a radiological release event is determined to be incredible, no further review is required. No individual facility in this analysis has a crash potential large enough to be credible. Therefore, an aircraft crash scenario is not required in the safety analysis for a single facility, but should be discussed relative to the ICPP aggregate areas, The highest probability of concern in the study was for aircraft to crash into the aggregate area for radiological materials storage facilities at the ICPP during Federal Aviation Administration (FAA) test flights

  18. 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

  19. Assessing safety culture using RADAR matrix

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. Safety assessment for TA-48 radiochemical operations

    International Nuclear Information System (INIS)

    1994-08-01

    The purpose of this report is to document an assessment performed to evaluate the safety of the radiochemical operations conducted at the Los Alamos National Laboratory operations area designated as TA-48. This Safety Assessment for the TA-48 radiochemical operations was prepared to fulfill the requirements of US Department of Energy (DOE) Order 5481.1B, ''Safety Analysis and Review System.'' The area designated as TA-48 is operated by the Chemical Science and Technology (CST) Division and is involved with radiochemical operations associated with nuclear weapons testing, evaluation of samples collected from a variety of environmental sources, and nuclear medicine activities. This report documents a systematic evaluation of the hazards associated with the radiochemical operations that are conducted at TA-48. The accident analyses are limited to evaluation of the expected consequences associated with a few bounding accident scenarios that are selected as part of the hazard analysis. Section 2 of this report presents an executive summary and conclusions, Section 3 presents pertinent information concerning the TA-48 site and surrounding area, Section 4 presents a description of the TA-48 radiochemical operations, and Section 5 presents a description of the individual facilities. Section 6 of the report presents an evaluation of the hazards that are associated with the TA-48 operations and Section 7 presents a detailed analysis of selected accident scenarios

  1. Grit-mediated frictional ignition of a polymer-bonded explosive during oblique impacts: Probability calculations for safety engineering

    International Nuclear Information System (INIS)

    Heatwole, Eric; Parker, Gary; Holmes, Matt; Dickson, Peter

    2015-01-01

    Frictional heating of high-melting-point grit particles during oblique impacts of consolidated explosives is considered to be the major source of ignition in accidents involving dropped explosives. It has been shown in other work that the lower temperature melting point of two frictionally interacting surfaces will cap the maximum temperature reached, which provides a simple way to mitigate the danger in facilities by implementing surfaces with melting points below the ignition temperature of the explosive. However, a recent series of skid testing experiments has shown that ignition can occur on low-melting-point surfaces with a high concentration of grit particles, most likely due to a grit–grit collision mechanism. For risk-based safety engineering purposes, the authors present a method to estimate the probability of grit contact and/or grit–grit collision during an oblique impact. These expressions are applied to potentially high-consequence oblique impact scenarios in order to give the probability of striking one or more grit particles (for high-melting-point surfaces), or the probability of one or more grit–grit collisions occurring (for low-melting-point surfaces). The probability is dependent on a variety of factors, many of which can be controlled for mitigation to achieve acceptable risk levels for safe explosives handling operations. - Highlights: • Unexpectedly, grit-mediated ignition of a PBX occurred on low-melting point surfaces. • On high-melting surfaces frictional heating is due to a grit–surface interaction. • For low-melting point surfaces the heating mechanism is grit–grit collisions. • A method for estimating the probability of ignition is presented for both surfaces

  2. Management of safety, safety culture and self assessment

    International Nuclear Information System (INIS)

    Carnino, A.

    2000-01-01

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

  3. Quantification of human reliability in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hirschberg, S.; Dankg, Vinh N.

    1996-01-01

    Human performance may substantially influence the reliability and safety of complex technical systems. For this reason, Human Reliability Analysis (HRA) constitutes an important part of Probabilistic Safety Assessment (PSAs) or Quantitative Risk Analyses (QRAs). The results of these studies as well as analyses of past accidents and incidents clearly demonstrate the importance of human interactions. The contribution of human errors to the core damage frequency (CDF), as estimated in the Swedish nuclear PSAs, are between 15 and 88%. A survey of the FRAs in the Swiss PSAs shows that also for the Swiss nuclear power plants the estimated HE contributions are substantial (49% of the CDF due to internal events in the case of Beznau and 70% in the case of Muehleberg; for the total CDF, including external events, 25% respectively 20%). Similar results can be extracted from the PSAs carried out for French, German, and US plants. In PSAs or QRAs, the adequate treatment of the human interactions with the system is a key to the understanding of accident sequences and their relative importance to overall risk. The main objectives of HRA are: first, to ensure that the key human interactions are systematically identified and incorporated into the safety analysis in a traceable manner, and second, to quantify the probabilities of their success and failure. Adopting a structured and systematic approach to the assessment of human performance makes it possible to provide greater confidence that the safety and availability of human-machine systems is not unduly jeopardized by human performance problems. Section 2 discusses the different types of human interactions analysed in PSAs. More generally, the section presents how HRA fits in the overall safety analysis, that is, how the human interactions to be quantified are identified. Section 3 addresses the methods for quantification. Section 4 concludes the paper by presenting some recommendations and pointing out the limitations of the

  4. Assessing wildfire occurrence probability in Pinus pinaster Ait. stands in Portugal

    Energy Technology Data Exchange (ETDEWEB)

    Marques, S.; Garcia-Gonzalo, J.; Botequim, B.; Ricardo, A.; Borges, J. G.; Tome, M.; Oliveira, M. M.

    2012-11-01

    Maritime pine (Pinus pinaster Ait.) is an important conifer from the western Mediterranean Basin extending over 22% of the forest area in Portugal. In the last three decades nearly 4% of Maritime pine area has been burned by wildfires. Yet no wildfire occurrence probability models are available and forest and fire management planning activities are thus carried out mostly independently of each other. This paper presents research to address this gap. Specifically, it presents a model to assess wildfire occurrence probability in regular and pure Maritime pine stands in Portugal. Emphasis was in developing a model based on easily available inventory data so that it might be useful to forest managers. For that purpose, data from the last two Portuguese National Forest Inventories (NFI) and data from wildfire perimeters in the years from 1998 to 2004 and from 2006 to 2007 were used. A binary logistic regression model was build using biometrics data from the NFI. Biometric data included indicators that might be changed by operations prescribed in forest planning. Results showed that the probability of wildfire occurrence in a stand increases in stand located at steeper slopes and with high shrubs load while it decreases with precipitation and with stand basal area. These results are instrumental for assessing the impact of forest management options on wildfire probability thus helping forest managers to reduce the risk of wildfires. (Author) 57 refs.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-05

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

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

    OpenAIRE

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

    2016-01-01

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

  8. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  9. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  10. Safety assessment of foods derived from genetically modified crops

    NARCIS (Netherlands)

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

    2003-01-01

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

  11. Protein single-model quality assessment by feature-based probability density functions.

    Science.gov (United States)

    Cao, Renzhi; Cheng, Jianlin

    2016-04-04

    Protein quality assessment (QA) has played an important role in protein structure prediction. We developed a novel single-model quality assessment method-Qprob. Qprob calculates the absolute error for each protein feature value against the true quality scores (i.e. GDT-TS scores) of protein structural models, and uses them to estimate its probability density distribution for quality assessment. Qprob has been blindly tested on the 11th Critical Assessment of Techniques for Protein Structure Prediction (CASP11) as MULTICOM-NOVEL server. The official CASP result shows that Qprob ranks as one of the top single-model QA methods. In addition, Qprob makes contributions to our protein tertiary structure predictor MULTICOM, which is officially ranked 3rd out of 143 predictors. The good performance shows that Qprob is good at assessing the quality of models of hard targets. These results demonstrate that this new probability density distribution based method is effective for protein single-model quality assessment and is useful for protein structure prediction. The webserver of Qprob is available at: http://calla.rnet.missouri.edu/qprob/. The software is now freely available in the web server of Qprob.

  12. A microcomputer program for energy assessment and aggregation using the triangular probability distribution

    Science.gov (United States)

    Crovelli, R.A.; Balay, R.H.

    1991-01-01

    A general risk-analysis method was developed for petroleum-resource assessment and other applications. The triangular probability distribution is used as a model with an analytic aggregation methodology based on probability theory rather than Monte-Carlo simulation. Among the advantages of the analytic method are its computational speed and flexibility, and the saving of time and cost on a microcomputer. The input into the model consists of a set of components (e.g. geologic provinces) and, for each component, three potential resource estimates: minimum, most likely (mode), and maximum. Assuming a triangular probability distribution, the mean, standard deviation, and seven fractiles (F100, F95, F75, F50, F25, F5, and F0) are computed for each component, where for example, the probability of more than F95 is equal to 0.95. The components are aggregated by combining the means, standard deviations, and respective fractiles under three possible siutations (1) perfect positive correlation, (2) complete independence, and (3) any degree of dependence between these two polar situations. A package of computer programs named the TRIAGG system was written in the Turbo Pascal 4.0 language for performing the analytic probabilistic methodology. The system consists of a program for processing triangular probability distribution assessments and aggregations, and a separate aggregation routine for aggregating aggregations. The user's documentation and program diskette of the TRIAGG system are available from USGS Open File Services. TRIAGG requires an IBM-PC/XT/AT compatible microcomputer with 256kbyte of main memory, MS-DOS 3.1 or later, either two diskette drives or a fixed disk, and a 132 column printer. A graphics adapter and color display are optional. ?? 1991.

  13. Retained gas sampler interim safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Pasamehmetoglu, K.O.; Miller, W.O.; Unal, C.; Fujita, R.K.

    1995-01-13

    This safety assessment addresses the proposed action to install, operate, and remove a Retained Gas Sampler (RGS) in Tank 101-SY at Hanford. Purpose of the RGS is to help characterize the gas species retained in the tank waste; the information will be used to refine models that predict the gas-producing behavior of the waste tank. The RGS will take samples of the tank from top to bottom; these samples will be analyzed for gas constituents. The proposed action is required as part of an evaluation of mitigation concepts for eliminating episodic gas releases that result in high hydrogen concentrations in the tank dome space.

  14. Safety assessment of outdoor live fire range

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-05-01

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

  15. Retained gas sampler interim safety assessment

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Miller, W.O.; Unal, C.; Fujita, R.K.

    1995-01-01

    This safety assessment addresses the proposed action to install, operate, and remove a Retained Gas Sampler (RGS) in Tank 101-SY at Hanford. Purpose of the RGS is to help characterize the gas species retained in the tank waste; the information will be used to refine models that predict the gas-producing behavior of the waste tank. The RGS will take samples of the tank from top to bottom; these samples will be analyzed for gas constituents. The proposed action is required as part of an evaluation of mitigation concepts for eliminating episodic gas releases that result in high hydrogen concentrations in the tank dome space

  16. The waste isolation safety assessment programme

    International Nuclear Information System (INIS)

    Brandstetter, A.; Harwell, M.A.

    1980-01-01

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

  17. Optimization of radiation therapy, III: a method of assessing complication probabilities from dose-volume histograms

    International Nuclear Information System (INIS)

    Lyman, J.T.; Wolbarst, A.B.

    1987-01-01

    To predict the likelihood of success of a therapeutic strategy, one must be able to assess the effects of the treatment upon both diseased and healthy tissues. This paper proposes a method for determining the probability that a healthy organ that receives a non-uniform distribution of X-irradiation, heat, chemotherapy, or other agent will escape complications. Starting with any given dose distribution, a dose-cumulative-volume histogram for the organ is generated. This is then reduced by an interpolation scheme (involving the volume-weighting of complication probabilities) to a slightly different histogram that corresponds to the same overall likelihood of complications, but which contains one less step. The procedure is repeated, one step at a time, until there remains a final, single-step histogram, for which the complication probability can be determined. The formalism makes use of a complication response function C(D, V) which, for the given treatment schedule, represents the probability of complications arising when the fraction V of the organ receives dose D and the rest of the organ gets none. Although the data required to generate this function are sparse at present, it should be possible to obtain the necessary information from in vivo and clinical studies. Volume effects are taken explicitly into account in two ways: the precise shape of the patient's histogram is employed in the calculation, and the complication response function is a function of the volume

  18. NUSS safety standards: A critical assessment

    International Nuclear Information System (INIS)

    Minogue, R.B.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  20. Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

    Energy Technology Data Exchange (ETDEWEB)

    R.M. Dolin

    2003-03-03

    This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

  1. Nirex Safety Assessment Research Programme bibliography, 1988

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1988-05-01

    This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research Programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW) and associated radiological assessments. All work referred to has been funded, or partly funded, by UK Nirex Limited, previously known as the Nuclear Industry Radioactive Waste Executive (NIREX). The bibliography has been divided into two sections, a List of Publications in roughly chronological order and an Author Index. The topics involved include near-field and far-field studies. The near-field includes the waste, its immobilising medium, its container, the engineered structure in which the containers are emplaced, and the immediately adjacent geological formation disturbed by the construction of the repository. The far-field is the undisturbed geological formation between the near-field and the biosphere. (author)

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

    NARCIS (Netherlands)

    Karanikas, Nektarios

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-09-15

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

  4. Failure probability assessment of wall-thinned nuclear pipes using probabilistic fracture mechanics

    International Nuclear Information System (INIS)

    Lee, Sang-Min; Chang, Yoon-Suk; Choi, Jae-Boong; Kim, Young-Jin

    2006-01-01

    The integrity of nuclear piping system has to be maintained during operation. In order to maintain the integrity, reliable assessment procedures including fracture mechanics analysis, etc., are required. Up to now, this has been performed using conventional deterministic approaches even though there are many uncertainties to hinder a rational evaluation. In this respect, probabilistic approaches are considered as an appropriate method for piping system evaluation. The objectives of this paper are to estimate the failure probabilities of wall-thinned pipes in nuclear secondary systems and to propose limited operating conditions under different types of loadings. To do this, a probabilistic assessment program using reliability index and simulation techniques was developed and applied to evaluate failure probabilities of wall-thinned pipes subjected to internal pressure, bending moment and combined loading of them. The sensitivity analysis results as well as prototypal integrity assessment results showed a promising applicability of the probabilistic assessment program, necessity of practical evaluation reflecting combined loading condition and operation considering limited condition

  5. Evaluation of fire probabilistic safety assessment for a PWR plant

    International Nuclear Information System (INIS)

    Wu, C.H.; Lin, T.J.; Kao, T.M.

    2001-01-01

    The internal fire analysis of the level 1 power operation probability safety assessment (PSA) for Maanshan (PWR) Nuclear Power Plant (MNPP) was updated. The fire analysis adopted a scenario-based PSA approach to systematically evaluate fire and smoke hazards and their associated risk impact to MNPP. The result shows that the core damage frequency (CDF) due to fire is about six times lower than the previous one analyzed by the Atomic Energy Council (AEC), Republic of China in 1987. The plant model was modified to reflect the impact of human events and recovery actions during fire. Many tabulated EXCEL spread-sheets were used for evaluation of the fire risk. The fire-induced CDF for MNPP is found to be 2.1 E-6 per year in this study. The relative results of the fire analysis will provide the bases for further risk-informed fire protection evaluation in the near future. (author)

  6. Regulatory review of safety cases and safety assessments for near surface

    International Nuclear Information System (INIS)

    Nys, V.

    2003-01-01

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

  7. Probability and Confidence Trade-space (PACT) Evaluation: Accounting for Uncertainty in Sparing Assessments

    Science.gov (United States)

    Anderson, Leif; Box, Neil; Carter, Katrina; DiFilippo, Denise; Harrington, Sean; Jackson, David; Lutomski, Michael

    2012-01-01

    There are two general shortcomings to the current annual sparing assessment: 1. The vehicle functions are currently assessed according to confidence targets, which can be misleading- overly conservative or optimistic. 2. The current confidence levels are arbitrarily determined and do not account for epistemic uncertainty (lack of knowledge) in the ORU failure rate. There are two major categories of uncertainty that impact Sparing Assessment: (a) Aleatory Uncertainty: Natural variability in distribution of actual failures around an Mean Time Between Failure (MTBF) (b) Epistemic Uncertainty : Lack of knowledge about the true value of an Orbital Replacement Unit's (ORU) MTBF We propose an approach to revise confidence targets and account for both categories of uncertainty, an approach we call Probability and Confidence Trade-space (PACT) evaluation.

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

    International Nuclear Information System (INIS)

    Jones, G.

    2013-01-01

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

  9. The relevance of the early history of probability theory to current risk assessment practices in mental health care.

    Science.gov (United States)

    Large, Matthew

    2013-12-01

    Probability theory is at the base of modern concepts of risk assessment in mental health. The aim of the current paper is to review the key developments in the early history of probability theory in order to enrich our understanding of current risk assessment practices.

  10. Power probability density function control and performance assessment of a nuclear research reactor

    International Nuclear Information System (INIS)

    Abharian, Amir Esmaeili; Fadaei, Amir Hosein

    2014-01-01

    Highlights: • In this paper, the performance assessment of static PDF control system is discussed. • The reactor PDF model is set up based on the B-spline functions. • Acquaints of Nu, and Th-h. equations solve concurrently by reformed Hansen’s method. • A principle of performance assessment is put forward for the PDF of the NR control. - Abstract: One of the main issues in controlling a system is to keep track of the conditions of the system function. The performance condition of the system should be inspected continuously, to keep the system in reliable working condition. In this study, the nuclear reactor is considered as a complicated system and a principle of performance assessment is used for analyzing the performance of the power probability density function (PDF) of the nuclear research reactor control. First, the model of the power PDF is set up, then the controller is designed to make the power PDF for tracing the given shape, that make the reactor to be a closed-loop system. The operating data of the closed-loop reactor are used to assess the control performance with the performance assessment criteria. The modeling, controller design and the performance assessment of the power PDF are all applied to the control of Tehran Research Reactor (TRR) power in a nuclear process. In this paper, the performance assessment of the static PDF control system is discussed, the efficacy and efficiency of the proposed method are investigated, and finally its reliability is proven

  11. Probabilistic safety assessment activities at Ignalina NPP

    International Nuclear Information System (INIS)

    Bagdonas, A.

    1999-01-01

    The Barselina Project was initiated in the summer 1991. The project was a multilateral co-operation between Lithuania, Russia and Sweden up until phase 3, and phase 4 has been performed as a bilateral between Lithuania and Sweden. The long-range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. During phase 3, from 1993 to 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. During phase 4, from 1994 to 1996, the PSA was further developed, taking into account plant changes, improved modelling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The model reflected the plant status before the outage 1996. During phase 4+, 1998 to 1999 the PSA model was upgraded taking into account the newest plant modifications. The new PSA model of CPS/AZRT was developed. Modelling was based on the Single Failure Analysis

  12. A formalism to generate probability distributions for performance-assessment modeling

    International Nuclear Information System (INIS)

    Kaplan, P.G.

    1990-01-01

    A formalism is presented for generating probability distributions of parameters used in performance-assessment modeling. The formalism is used when data are either sparse or nonexistent. The appropriate distribution is a function of the known or estimated constraints and is chosen to maximize a quantity known as Shannon's informational entropy. The formalism is applied to a parameter used in performance-assessment modeling. The functional form of the model that defines the parameter, data from the actual field site, and natural analog data are analyzed to estimate the constraints. A beta probability distribution of the example parameter is generated after finding four constraints. As an example of how the formalism is applied to the site characterization studies of Yucca Mountain, the distribution is generated for an input parameter in a performance-assessment model currently used to estimate compliance with disposal of high-level radioactive waste in geologic repositories, 10 CFR 60.113(a)(2), commonly known as the ground water travel time criterion. 8 refs., 2 figs

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

    International Nuclear Information System (INIS)

    Chung, Dae Wook

    2011-01-01

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

  14. Assessment of Safety Standards for Automotive Electronic Control Systems

    Science.gov (United States)

    2016-06-01

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

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

    International Nuclear Information System (INIS)

    Lange, B. A.

    1997-01-01

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

  16. Safety assessment of OPG's used fuel for dry storage

    International Nuclear Information System (INIS)

    Roman, H.; Khan, A.

    2005-01-01

    'Full text:' Ontario Power Generation (OPG) operates the Pickering Waste Management Facility (PWMF) and Western Waste Management Facility (WWMF) where OPG has been storing 10-year or older used fuel in the Dry Storage Containers (DSCs) since 1996 and 2003 respectively. The construction licence for the Darlington Used Fuel Dry Storage Facility (DUFDSF) was obtained in August 2004. Safety assessment of the used fuel for dry storage is required to support each request for regulatory approval to construct and operate a dry storage facility. The objective of the safety assessment is to assess the used fuel performance under normal operation and postulated credible accident scenarios. A reference used fuel bundle is defined based on the operating history and data on fuel discharged from the reactors of the specific nuclear generating station. The characteristics of the reference used fuel bundle are used to calculate the nuclide inventory, source term and decay heat used for the assessment. When assessing malfunctions and accidents, postulated external and internal events are considered. Consideration is also given to the design basis accidents of the specific nuclear generating station that could affect the used fuel under dry storage. For those events deemed credible (i.e. probability > 10 -7 ), a bounding fuel failure consequence is predicted. Given the chemical characteristics of the radionuclides in used fuel, the design of the CANDU fuel and the conditions inside the DSC, in the event that a used fuel bundle should become damaged during used fuel dry storage operations, the only significant radionuclides species that are volatile are krypton-85 and tritium. Release of these radionuclides is considered in calculating public and worker doses. (author)

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

    NARCIS (Netherlands)

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

    2012-01-01

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

  18. Nirex Safety Assessment Research Programme bibliography, 1987

    International Nuclear Information System (INIS)

    Cooper, M.J.; Hodgkinson, D.P.

    1987-06-01

    This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW). All work referred to has been funded, or partly funded, by UK Nirex Limited, previously known as the Nuclear Industry Radioactive Waste Executive (NIREX). The bibliography has been divided into two sections, a List of Publications in roughly chronological order and an Author Index. The topics involved include near-field and far-field studies. The near-field includes the waste, its immobilising medium, its container, the engineered structure in which the containers are emplaced, and the immediately adjacent geological formation disturbed by the construction of the repository. The far-field is the undisturbed geological formation between the near-field and the biosphere. (author)

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    2011-01-01

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

  2. Safety Assessment in Installation of Precast Concrete

    Directory of Open Access Journals (Sweden)

    Yashrri S.N.

    2014-03-01

    Full Text Available This study was carried out to identify the safety aspects and the level of safety during the installation process in construction sites. A questionnaire survey and interviews were done to provide data on safety requirements in precast concrete construction. All of the interviews and the research questionnaire survey were conducted among contractors that are registered as class 1 to class 7 with the Construction Industry Development Board (CIDB and class A to class G with Pusat Khidmat Kontraktor (PKK in Penang. Returned questionnaires were analysed with the use of simple percentages and the Likert Scale analysis method to identify safety aspects of precast construction. The results indicate that the safety aspect implemented by companies involved in the precast construction process is at a good level in the safety aspect during bracing, propping, welding and grouting processes and at a very good level of safety in general aspects and safety aspects during lifting processes.

  3. ASSESSMENT OF FOOD SAFETY PRACTICES AMONG CASSAVA ...

    African Journals Online (AJOL)

    Philips Olusola

    public policies on food safety management. ... in the contemporary world and a significant factor in the reduction of productivity. There is some .... adduced to the views of the majority who considered food safety practices a big ..... Contingency.

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

    International Nuclear Information System (INIS)

    Armstrong, D.L.

    1994-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, D.L.

    1994-08-01

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

  6. Assessment of fragment projection hazard: probability distributions for the initial direction of fragments.

    Science.gov (United States)

    Tugnoli, Alessandro; Gubinelli, Gianfilippo; Landucci, Gabriele; Cozzani, Valerio

    2014-08-30

    The evaluation of the initial direction and velocity of the fragments generated in the fragmentation of a vessel due to internal pressure is an important information in the assessment of damage caused by fragments, in particular within the quantitative risk assessment (QRA) of chemical and process plants. In the present study an approach is proposed to the identification and validation of probability density functions (pdfs) for the initial direction of the fragments. A detailed review of a large number of past accidents provided the background information for the validation procedure. A specific method was developed for the validation of the proposed pdfs. Validated pdfs were obtained for both the vertical and horizontal angles of projection and for the initial velocity of the fragments. Copyright © 2014 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  8. Nirex safety assessment research programme bibliography, 1991

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1991-09-01

    This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research Programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW) and associated radiological assessments. The bibliography has been divided into two sections, a list of Publications in roughly chronological order and an Author Index. The topics involved include near-field and far-field studies. The near-field includes the waste package, filling or sealing materials, and those parts of the host medium whose characteristics have been or could be altered by the repository or its content. The far-field is the rock formation outside the repository, including the surrounding strata, at a distance from the waste disposal site such that, for modelling purposes, the site may be considered as a single entity, and the effects of individual waste packages are indistinguishable in the effects of the whole. The far-field includes also the biosphere, into which radionuclides from the waste could conceivably migrate in the future. (author)

  9. The Nirex safety assessment research programme

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1988-07-01

    This report describes progress on the Nirex Safety Assessment Research Programme in 1987/88. The programme is concerned with research into the disposal of low-level waste (LLW) and intermediate-level waste (ILW) into underground repositories. At the beginning of 1987/88 a range of techniques for measuring and modelling far-field phenomena were being applied to near-surface disposal of low-level waste in clay. However, during the year the far-field studies were redirected to consider generic geological materials of interest for deep disposal of low and intermediate-level waste, which is now the preferred option in the UK. A substantial part of the programme is concerned with the effectiveness of near-field barriers to water-borne leakage of radionuclides from cementitious repositories. Considerable progress has been made in quantifying this and laying the foundations for robust and reliable radiological assessments to be made with appropriate models. New projects have also been initiated to study the evolution and migration of gases from an underground repository and to consider the contribution of the biosphere to the retardation of radionuclides. (author)

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

    Science.gov (United States)

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

    2013-12-01

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

  11. Geosphere process report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    Skagius, Kristina

    2010-11-01

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

  12. Geosphere process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-11-15

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

  13. Safety assessment of radioactive wastes storage 'Mironova Gora'

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. Probabilistic safety assessment as a standpoint for decision making

    International Nuclear Information System (INIS)

    Cepin, M.

    2001-01-01

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

  16. Probabilistic safety assessment in nuclear power plant management

    International Nuclear Information System (INIS)

    Holloway, N.J.

    1989-06-01

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

  17. [Patient safety in antibiotics administration: Risk assessment].

    Science.gov (United States)

    Maqueda Palau, M; Pérez Juan, E

    To determine the level of risk in the preparation and administration of antibiotics frequently used in the Intensive Care Unit using a risk matrix. A study was conducted using situation analysis and literature review of databases, protocols and good practice guidelines on intravenous therapy, drugs, and their administration routes. The most used antibiotics in the ICU registered in the ENVIN-HELICS program from 1 April to 30 June 2015 were selected. In this period, 257 patients received antimicrobial treatment and 26 antibiotics were evaluated. Variables studied: A risk assessment of each antibiotic using the scale Risk Assessment Tool, of the National Patient Safety Agency, as well as pH, osmolarity, type of catheter recommended for administration, and compatibility and incompatibility with other antibiotics studied. Almost two-thirds (65.3%) of antibiotics had more than 3 risk factors (represented by a yellow stripe), with the remaining 34.7% of antibiotics having between 0 and 2 risk factors (represented by a green stripe). There were no antibiotics with 6 or more risk factors (represented by a red stripe). Most drugs needed reconstitution, additional dilution, and the use of part of the vial to administer the prescribed dose. More than half of the antibiotics studied had a moderate risk level; thus measures should be adopted in order to reduce it. The risk matrix is a useful tool for the assessment and detection of weaknesses associated with the preparation and administration of intravenous antibiotics. Copyright © 2016 SECA. Publicado por Elsevier España, S.L.U. All rights reserved.

  18. Safety assessment of the BELENE NPP

    International Nuclear Information System (INIS)

    Bruna, G.B.

    2010-01-01

    The most important outcome of the RISKAUDIT’s review of the BELENE’s ISAR was that the design of the plant is in general conformity with the applied requirements, the Bulgarian regulation and the international practice.No substantial design flaws have been identified and no major issues are found as missing. However, the ISAR should be significantly improved before issuing the construction permit. RISKAUDIT asked for elaboration of a revised version of the ISAR, in which all the recommendations, the requests for information as well the questions should be comprehensively addressed, and explicitly referenced to facilitate their assessment. The satisfactory accomplishment of this duty has been considered. The considered as a as prerequisite for the issuance of the construction permit, 2.The REV 2 of the ISAR have been issued at the beginning of this year, BNRA and RISKAUDIT signed on a new collaboration under the above-mentioned Contract, which has the following main objectives:a. Assessing the implementation of the above-mentioned recommendations, b.Verifying and integrating the answers provided to the questions,c.Performing additional in-depth analysis aimed at completing and finalizing the previous investigations as well as addressing some complementary relevant safety issues.3.The inception meeting of this Contract is planed June 17 –18 in Sofia. Details on the planning have to be defined, but, tentatively, the assessment activity should be completed by the end of current year and the accompanying in-depth analysis at the latest at the beginnings of 2011. 4.4.The issuance of those Reports is an important technical support., The support to to the process for granting the construction permit

  19. Qualitative uncertainty analysis in probabilistic safety assessment context

    International Nuclear Information System (INIS)

    Apostol, M.; Constantin, M; Turcu, I.

    2007-01-01

    In Probabilistic Safety Assessment (PSA) context, an uncertainty analysis is performed either to estimate the uncertainty in the final results (the risk to public health and safety) or to estimate the uncertainty in some intermediate quantities (the core damage frequency, the radionuclide release frequency or fatality frequency). The identification and evaluation of uncertainty are important tasks because they afford credit to the results and help in the decision-making process. Uncertainty analysis can be performed qualitatively or quantitatively. This paper performs a preliminary qualitative uncertainty analysis, by identification of major uncertainty in PSA level 1- level 2 interface and in the other two major procedural steps of a level 2 PSA i.e. the analysis of accident progression and of the containment and analysis of source term for severe accidents. One should mention that a level 2 PSA for a Nuclear Power Plant (NPP) involves the evaluation and quantification of the mechanisms, amount and probabilities of subsequent radioactive material releases from the containment. According to NUREG 1150, an important task in source term analysis is fission products transport analysis. The uncertainties related to the isotopes distribution in CANDU NPP primary circuit and isotopes' masses transferred in the containment, using SOPHAEROS module from ASTEC computer code will be also presented. (authors)

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

    International Nuclear Information System (INIS)

    Metcalf, P.

    2003-01-01

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

  1. Using Prediction Markets to Generate Probability Density Functions for Climate Change Risk Assessment

    Science.gov (United States)

    Boslough, M.

    2011-12-01

    Climate-related uncertainty is traditionally presented as an error bar, but it is becoming increasingly common to express it in terms of a probability density function (PDF). PDFs are a necessary component of probabilistic risk assessments, for which simple "best estimate" values are insufficient. Many groups have generated PDFs for climate sensitivity using a variety of methods. These PDFs are broadly consistent, but vary significantly in their details. One axiom of the verification and validation community is, "codes don't make predictions, people make predictions." This is a statement of the fact that subject domain experts generate results using assumptions within a range of epistemic uncertainty and interpret them according to their expert opinion. Different experts with different methods will arrive at different PDFs. For effective decision support, a single consensus PDF would be useful. We suggest that market methods can be used to aggregate an ensemble of opinions into a single distribution that expresses the consensus. Prediction markets have been shown to be highly successful at forecasting the outcome of events ranging from elections to box office returns. In prediction markets, traders can take a position on whether some future event will or will not occur. These positions are expressed as contracts that are traded in a double-action market that aggregates price, which can be interpreted as a consensus probability that the event will take place. Since climate sensitivity cannot directly be measured, it cannot be predicted. However, the changes in global mean surface temperature are a direct consequence of climate sensitivity, changes in forcing, and internal variability. Viable prediction markets require an undisputed event outcome on a specific date. Climate-related markets exist on Intrade.com, an online trading exchange. One such contract is titled "Global Temperature Anomaly for Dec 2011 to be greater than 0.65 Degrees C." Settlement is based

  2. Safety Assessment of Polyether Lanolins as Used in Cosmetics.

    Science.gov (United States)

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

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

  3. Risk Assessment Using the Three Dimensions of Probability (Likelihood), Severity, and Level of Control

    Science.gov (United States)

    Watson, Clifford C.

    2011-01-01

    Traditional hazard analysis techniques utilize a two-dimensional representation of the results determined by relative likelihood and severity of the residual risk. These matrices present a quick-look at the Likelihood (Y-axis) and Severity (X-axis) of the probable outcome of a hazardous event. A three-dimensional method, described herein, utilizes the traditional X and Y axes, while adding a new, third dimension, shown as the Z-axis, and referred to as the Level of Control. The elements of the Z-axis are modifications of the Hazard Elimination and Control steps (also known as the Hazard Reduction Precedence Sequence). These steps are: 1. Eliminate risk through design. 2. Substitute less risky materials for more hazardous materials. 3. Install safety devices. 4. Install caution and warning devices. 5. Develop administrative controls (to include special procedures and training.) 6. Provide protective clothing and equipment. When added to the two-dimensional models, the level of control adds a visual representation of the risk associated with the hazardous condition, creating a tall-pole for the least-well-controlled failure while establishing the relative likelihood and severity of all causes and effects for an identified hazard. Computer modeling of the analytical results, using spreadsheets and three-dimensional charting gives a visual confirmation of the relationship between causes and their controls.

  4. 76 FR 74723 - New Car Assessment Program (NCAP); Safety Labeling

    Science.gov (United States)

    2011-12-01

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

  5. Promoting and assessment of safety culture within regulatory body

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Science.gov (United States)

    2013-03-08

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

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

    International Nuclear Information System (INIS)

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. Assessing progress in the development of safety culture

    International Nuclear Information System (INIS)

    Rotaru, Ioan; Ghita, Sorin

    1999-01-01

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

  11. Development of Safety Culture Assessment Strategy for Korean NPP

    International Nuclear Information System (INIS)

    Park, Jung Hwan; Kim, Jong Hyun

    2014-01-01

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

  12. Development of Safety Culture Assessment Strategy for Korean NPP

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  13. The role of probabilistic safety assessment in the design

    International Nuclear Information System (INIS)

    Green, A.; Ingham, E.L.

    1989-01-01

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

  14. Plant assessment system and safety culture

    International Nuclear Information System (INIS)

    Chun, Chuyoung

    1996-01-01

    The government, upon these events, keenly felt the necessity for developing the safety culture which was already forwarded in nuclear industries and started taking actions to propagate it to all parts of society. The government established a social safety director position under the Prime Minister's jurisdiction and also established a Safety Culture Promotion Headquarters in which 7 ministries and other organizations, such as Korea Economic Council, Federation of Korea Trade Union and Women's Federation Council were participating. In accordance with the government's strong will to enhance the safety consciousness of people, safety campaigns are being developed voluntarily in the private sector. The formation of non-governmental organizations, such as People's Central Council of Safety Culture Promotion, shows a good example of such movement

  15. Failure rate data for fusion safety and risk assessment

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1993-01-01

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

  16. Role of physician perception of patient smile on pretest probability assessment for acute pulmonary embolism.

    Science.gov (United States)

    Kline, Jeffrey A; Neumann, Dawn; Hall, Cassandra L; Capito, Jacob

    2017-02-01

    Many clinicians use a global visual interpretation of patient appearance to decide if a patient looks sick or not. For patients with suspected acute pulmonary embolism (PE), we tested the relationship between visual appearance of a happy patient facial affect and probability of PE+ on CT pulmonary angiography (CTPA). Eligible patients were selected by usual care to undergo CTPA, the criterion standard for PE+ or PE-. Prior to CTPA result, trained study personnel obtained physician pretest probability using the gestalt method (visual analogue scale, 0%-100%), the Wells score (0-12) and physicians' impression of whether the patient smiled during the initial examination (smile+). Patients' faces were also video recorded and analysed with an automated neural network-based algorithm (Noldus FaceReader) for happy affect. Of the 208 patients enrolled, 27 were PE+ and smile+ was more frequent in patients with PE+ than PE-, a finding confirmed by the Noldus. The diagnostic sensitivity and specificity of smile was low, and physicians overestimated presence of an alternative diagnosis more likely to PE with smile+ than smile- patients in patients with true PE. As a result, the area under the receiver operating characteristic curve (AUROC) was lower for the Wells score in smile+ patients. However, the physicians' mean gestalt estimate of PE did not differ with smile status, nor did smile status affect the AUROC for gestalt. In patients with suspected PE, physician recollection of patients' smile+ was more common in PE+ patients, and was associated with a less accurate Wells score, primarily because physicians overestimated probability of alternative diagnosis. However, the overall diagnostic accuracy of physicians' gestalt did not differ with perceived smile status. These data suggest that the patients' smile had less effect on the numeric gestalt pretest probability assessment than on the binary decision about an alternative diagnosis. Published by the BMJ Publishing Group

  17. ITER plasma safety interface models and assessments

    International Nuclear Information System (INIS)

    Uckan, N.A.; Bartels, H-W.; Honda, T.; Amano, T.; Boucher, D.; Post, D.; Wesley, J.

    1996-01-01

    Physics models and requirements to be used as a basis for safety analysis studies are developed and physics results motivated by safety considerations are presented for the ITER design. Physics specifications are provided for enveloping plasma dynamic events for Category I (operational event), Category II (likely event), and Category III (unlikely event). A safety analysis code SAFALY has been developed to investigate plasma anomaly events. The plasma response to ex-vessel component failure and machine response to plasma transients are considered

  18. Assessing verticalization effects on urban safety perception

    OpenAIRE

    Lourenço, Ricardo Barros

    2017-01-01

    We describe an experiment with the modeling of urban verticalization effects on perceived safety scores as obtained with computer vision on Google Streetview data for New York City. Preliminary results suggests that for smaller buildings (between one and seven floors), perceived safety increases with building height, but that for high-rise buildings, perceived safety decreases with increased height. We also determined that while height contributing for this relation, other zonal aspects also ...

  19. Risk measures in living probabilistic safety assessment

    International Nuclear Information System (INIS)

    Holmberg, J.; Niemelae, I.

    1993-05-01

    The main objectives of the study are: to define risk measures and suggested uses of them in various living PSA applications for the operational safety management and to describe specific model features required for living PSA applications. The report is based on three case studies performed within the Nordic research project Safety Evaluation by Use of Living PSA and Safety Indicators. (48 refs., 11 figs., 17 tabs.)

  20. Systematic assessment of laser safety in otolaryngology

    Science.gov (United States)

    Oswal, V. H.

    2001-01-01

    Risk management of lasers can be broadly define das a process of identification of the risk, assessment of the risk and steps taken to avert the risk. The risk management may be divided into: Risk inherent to the technology and risk in clinical use. Within the National Health Service in the UK, a useful document, which provides hospital laser users with advice on safety, is the 'Guidance on the Safe Use of Lasers in Medical and Dental Practice' issued by the Medical Devices Agency for the Department of Health in the UK. It recommends the appointment of a Laser Protection Adviser (LPA) who is knowledgeable in the evaluation of laser hazards. One of the duties LPA is to ensure that Local Rules are drawn up for each specific application of a laser. A Laser Protection Supervisor (LPS) should also be appointed with responsibility to ensure that the Local Rules are observed. It is a sensible precaution that laser users should be those approved by the Laser Protection Supervisor in consultation with the Laser Protection Advisor. All laser users should sign a statement that they have read and understood the Local Rules.

  1. Reliability assessment for thickness measurements of pipe wall using probability of detection

    International Nuclear Information System (INIS)

    Nakamoto, Hiroyuki; Kojima, Fumio; Kato, Sho

    2013-01-01

    This paper proposes a reliability assessment method for thickness measurements of pipe wall using probability of detection (POD). Thicknesses of pipes are measured by qualified inspectors with ultrasonic thickness gauges. The inspection results are affected by human factors of the inspectors and include some errors, because the inspectors have different experiences and frequency of inspections. In order to ensure reliability for inspection results, first, POD evaluates experimental results of pipe-wall thickness inspection. We verify that the results have differences depending on inspectors including qualified inspectors. Second, two human factors that affect POD are indicated. Finally, it is confirmed that POD can identify the human factors and ensure reliability for pipe-wall thickness inspections. (author)

  2. The safety assessment of OPR-1000 nuclear power plant for station blackout accident applying the combined deterministic and probabilistic procedure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu, E-mail: littlewing@kins.re.kr [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2014-08-15

    Highlights: • The combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. • The safety assessment of OPR-1000 nuclear power plant for SBO accident is performed by applying the CDPP. • By estimating the offsite power restoration time appropriately, the SBO risk is reevaluated. • It is concluded that the CDPP is applicable to safety assessment of BDBAs without significant erosion of the safety margin. - Abstract: Station blackout (SBO) is a typical beyond design basis accident (BDBA) and significant contributor to overall plant risk. The risk analysis of SBO could be important basis of rulemaking, accident mitigation strategy, etc. Recently, studies on the integrated approach of deterministic and probabilistic method for nuclear safety in nuclear power plants have been done, and among them, the combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident was performed by applying the CDPP. It was confirmed that the SBO risk should be reevaluated by eliminating excessive conservatism in existing probabilistic safety assessment to meet the targeted core damage frequency and conditional core damage probability. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it is concluded that the CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

  3. Track 2 sites: Guidance for assessing low probability hazard sites at the INEL. Revision 6

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-01

    This document presents guidance for assessment of Track 2 low probability hazard sites (LPHS) at the Idaho National Engineering Laboratory (INEL). The Track 2 classification was developed specifically for the INEL to streamline the implementation of Comprehensive Environmental Response, Compensation, and Liability Act. Track 2 LPHSs are described as sites where insufficient data are available to make a decision concerning the risk level or to select or design a remedy. As such, these types of sites are not described in the National Contingency Plan or existing regulatory guidance. The goal of the Track 2 process is to evaluate LPHSs using existing qualitative and quantitative data to minimize the collection of new environmental data. To this end, this document presents a structured format consisting of a series of questions and tables. A qualitative risk assessment is used. The process is iterative, and addresses an LPHS from multiple perspectives (i.e., historical, empirical, process) in an effort to generate a reproducible and defensible method. This rigorous approach follows the data quality objective process and establishes a well organized, logical approach to consolidate and assess existing data, and set decision criteria. If necessary, the process allows for the design of a sampling and analysis strategy to obtain new environmental data of appropriate quality to support decisions for each LPHS. Finally, the guidance expedites consensus between regulatory parties by emphasizing a team approach to Track 2 investigations.

  4. Track 2 sites: Guidance for assessing low probability hazard sites at the INEL

    International Nuclear Information System (INIS)

    1994-01-01

    This document presents guidance for assessment of Track 2 low probability hazard sites (LPHS) at the Idaho National Engineering Laboratory (INEL). The Track 2 classification was developed specifically for the INEL to streamline the implementation of Comprehensive Environmental Response, Compensation, and Liability Act. Track 2 LPHSs are described as sites where insufficient data are available to make a decision concerning the risk level or to select or design a remedy. As such, these types of sites are not described in the National Contingency Plan or existing regulatory guidance. The goal of the Track 2 process is to evaluate LPHSs using existing qualitative and quantitative data to minimize the collection of new environmental data. To this end, this document presents a structured format consisting of a series of questions and tables. A qualitative risk assessment is used. The process is iterative, and addresses an LPHS from multiple perspectives (i.e., historical, empirical, process) in an effort to generate a reproducible and defensible method. This rigorous approach follows the data quality objective process and establishes a well organized, logical approach to consolidate and assess existing data, and set decision criteria. If necessary, the process allows for the design of a sampling and analysis strategy to obtain new environmental data of appropriate quality to support decisions for each LPHS. Finally, the guidance expedites consensus between regulatory parties by emphasizing a team approach to Track 2 investigations

  5. Reliability Assessment of Wind Farm Electrical System Based on a Probability Transfer Technique

    Directory of Open Access Journals (Sweden)

    Hejun Yang

    2018-03-01

    Full Text Available The electrical system of a wind farm has a significant influence on the wind farm reliability and electrical energy yield. The disconnect switch installed in an electrical system cannot only improve the operating flexibility, but also enhance the reliability for a wind farm. Therefore, this paper develops a probabilistic transfer technique for integrating the electrical topology structure, the isolation operation of disconnect switch, and stochastic failure of electrical equipment into the reliability assessment of wind farm electrical system. Firstly, as the traditional two-state reliability model of electrical equipment cannot consider the isolation operation, so the paper develops a three-state reliability model to replace the two-state model for incorporating the isolation operation. In addition, a proportion apportion technique is presented to evaluate the state probability. Secondly, this paper develops a probabilistic transfer technique based on the thoughts that through transfer the unreliability of electrical system to the energy transmission interruption of wind turbine generators (WTGs. Finally, some novel indices for describing the reliability of wind farm electrical system are designed, and the variance coefficient of the designed indices is used as a convergence criterion to determine the termination of the assessment process. The proposed technique is applied to the reliability assessment of a wind farm with the different topologies. The simulation results show that the proposed techniques are effective in practical applications.

  6. Tree Simulation Techniques for Integrated Safety Assessment

    International Nuclear Information System (INIS)

    Melendez Asensio, E.; Izquierdo Rocha, J.M.; Sanchez Perez, M.; Hortal Reymundo, J.; Perez Mulas, A.

    1999-01-01

    techniques are: (a) An unifying theory that should (i) establish the relationship among different approaches and, in particular, be able to demonstrate the standard safety assessment approach as a particular case, (ii) identify implicit assumptions in present practice and (iii) establish a sound scientific reference for an ideal treatment in order to judge the relative importance of implicit and explicit assumptions. In addition, the theoretical developments help to identify the type of applications where the new developments will be a necessary requirement. (b) The capability for simulation of trees. By this we mean the techniques required to be able to efficiently simulate all branches. Historically algorithms able to do this were already implemented in earlier pioneering work for discrete number of branches while stochastic branching requires Montecarlo techniques. (c) The capability to incorporate new types of branching, particularly operator actions. This paper shortly reviews these aspects and justifies in that frame our particular development, denoted here as Integrated Safety Assessment methodology. In this method, the dynamics of the event is followed by transient simulation in tree form, building a Setpoint or Deterministic Dynamic Event Tree (DDET). When a setpoint that should trigger the actuation of a protection is crossed, the tree is opened in branches corresponding to different functioning states of the protection device and each branch followed by the engineering simulator. One of these states is the nominal state, which, in the PSAs, is Associated to the success criterion of the system

  7. The Safety Assessment of Long term Interim Storage at Sellafield

    International Nuclear Information System (INIS)

    Buchan, Andrew B.

    2014-01-01

    The Sellafield site comprises a wide range of nuclear facilities, including operating facilities associated with the Magnox reprocessing programme, the Thermal Oxide Reprocessing Plant (THORP) and a range of waste treatment plants. The operational life of some of the site facilities currently extends to 2120, requiring the retrieval, treatment, consolidation and safe extended storage of a variety of radioactive materials. Sellafield has utilised its existing safety assessment processes to inform and prioritise studies into beyond design basis events and resilience evaluation required following Fukushima by UK regulators and industry bodies such as WANO. There are significant differences between NPPs, for which the ENSREG 'stress tests' were originally intended, and the Sellafield site which is instead centered around two reprocessing facilities (Magnox and THORP), with a supporting infrastructure of waste processing and storage facilities, coupled with a legacy of high hazard older facilities. In the former case the consequences of a catastrophic failure are promptly realised, leading to significant problems such as fuel failure / meltdown in AGRs and LWRs respectively in the event of complete loss of cooling. At Sellafield, the processes are carried out at comparatively low temperatures and pressures with relatively low rates of change following any loss of cooling. Instead the consequences of catastrophic failure at Sellafield are more directly related to the very large inventories of radioactive materials, including high level liquid wastes and unprocessed fuels, present in specific plants and to the condition of the ageing assets, holding legacy wastes. The paper will consider how a number of analysis techniques have been used to establish a safety case for fault and accident conditions, i.e. design basis analysis (DBA), probabilistic safety analysis (PSA) and severe accident analysis (SAA) DBA is focused on the key safety measures for those initiating faults

  8. Safety culture' is integrating 'human' into risk assessment

    International Nuclear Information System (INIS)

    Sugimoto, Taiji

    2014-01-01

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

  9. Key issues on safety design basis selection and safety assessment

    International Nuclear Information System (INIS)

    An, S.; Togo, Y.

    1976-01-01

    In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

  10. Safety assessment and detection methods of genetically modified organisms.

    Science.gov (United States)

    Xu, Rong; Zheng, Zhe; Jiao, Guanglian

    2014-01-01

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

  11. Online probabilistic operational safety assessment of multi-mode engineering systems using Bayesian methods

    International Nuclear Information System (INIS)

    Lin, Yufei; Chen, Maoyin; Zhou, Donghua

    2013-01-01

    In the past decades, engineering systems become more and more complex, and generally work at different operational modes. Since incipient fault can lead to dangerous accidents, it is crucial to develop strategies for online operational safety assessment. However, the existing online assessment methods for multi-mode engineering systems commonly assume that samples are independent, which do not hold for practical cases. This paper proposes a probabilistic framework of online operational safety assessment of multi-mode engineering systems with sample dependency. To begin with, a Gaussian mixture model (GMM) is used to characterize multiple operating modes. Then, based on the definition of safety index (SI), the SI for one single mode is calculated. At last, the Bayesian method is presented to calculate the posterior probabilities belonging to each operating mode with sample dependency. The proposed assessment strategy is applied in two examples: one is the aircraft gas turbine, another is an industrial dryer. Both examples illustrate the efficiency of the proposed method

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

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

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

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

    NARCIS (Netherlands)

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

    2016-01-01

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

  14. Safety assessment of primary system components at the USNRC

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-12-31

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

  15. Safety assessment requirements for onsite transfers of radioactive material

    International Nuclear Information System (INIS)

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

    1992-05-01

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

  16. In prospect: role of safety assessment and risk regulation

    International Nuclear Information System (INIS)

    Novegno, A.; Askulaj, Eh.

    1987-01-01

    Problems of accident prevention in industry and power engineering are considered for the sake of environment and human health protection. Investigations into comparison of power system risks are conducted; based on the data obtained a possibility to control the risk has appeared. The IAEA provides an active assistance in realization of a program of coordinated investigations on the risk assessment using the cost-benefit method. For each NPP investigation into all types of its effect on the environment (risk for personnel and population under normal radioactivity releases and in case of accidents), is conducted. Two approaches to calculating the impacts of accidents at NPPs-'determination' one, based on the designed accident and safety probability evaluation exist. Regional approach appears to be the best one when solving the problems of risk control. Attention is paid to a joint project of the IAEA-UNO and WHO related to risk assessment and control for human health and environment protection at power and other complex commercial systems

  17. Patient safety culture assessment in oman.

    Science.gov (United States)

    Al-Mandhari, Ahmed; Al-Zakwani, Ibrahim; Al-Kindi, Moosa; Tawilah, Jihane; Dorvlo, Atsu S S; Al-Adawi, Samir

    2014-07-01

    To illustrate the patient safety culture in Oman as gleaned via 12 indices of patient safety culture derived from the Hospital Survey on Patient Safety Culture (HSPSC) and to compare the average positive response rates in patient safety culture between Oman and the USA, Taiwan, and Lebanon. This was a cross-sectional research study employed to gauge the performance of HSPSC safety indices among health workers representing five secondary and tertiary care hospitals in the northern region of Oman. The participants (n=398) represented different professional designations of hospital staff. Analyses were performed using univariate statistics. The overall average positive response rate for the 12 patient safety culture dimensions of the HSPSC survey in Oman was 58%. The indices from HSPSC that were endorsed the highest included 'organizational learning and continuous improvement' while conversely, 'non-punitive response to errors' was ranked the least. There were no significant differences in average positive response rates between Oman and the United States (58% vs. 61%; p=0.666), Taiwan (58% vs. 64%; p=0.386), and Lebanon (58% vs. 61%; p=0.666). This study provides the first empirical study on patient safety culture in Oman which is similar to those rates reported elsewhere. It highlights the specific strengths and weaknesses which may stem from the specific milieu prevailing in Oman.

  18. Patient Safety Culture Assessment in Oman

    Science.gov (United States)

    Al-Mandhari, Ahmed; Al-Zakwani, Ibrahim; Al-Kindi, Moosa; Tawilah, Jihane; Dorvlo, Atsu S.S.; Al-Adawi, Samir

    2014-01-01

    Objective To illustrate the patient safety culture in Oman as gleaned via 12 indices of patient safety culture derived from the Hospital Survey on Patient Safety Culture (HSPSC) and to compare the average positive response rates in patient safety culture between Oman and the USA, Taiwan, and Lebanon. Methods This was a cross-sectional research study employed to gauge the performance of HSPSC safety indices among health workers representing five secondary and tertiary care hospitals in the northern region of Oman. The participants (n=398) represented different professional designations of hospital staff. Analyses were performed using univariate statistics. Results The overall average positive response rate for the 12 patient safety culture dimensions of the HSPSC survey in Oman was 58%. The indices from HSPSC that were endorsed the highest included ‘organizational learning and continuous improvement’ while conversely, ‘non-punitive response to errors’ was ranked the least. There were no significant differences in average positive response rates between Oman and the United States (58% vs. 61%; p=0.666), Taiwan (58% vs. 64%; p=0.386), and Lebanon (58% vs. 61%; p=0.666). Conclusion This study provides the first empirical study on patient safety culture in Oman which is similar to those rates reported elsewhere. It highlights the specific strengths and weaknesses which may stem from the specific milieu prevailing in Oman. PMID:25170407

  19. 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

  20. Total Probability of Collision as a Metric for Finite Conjunction Assessment and Collision Risk Management

    Science.gov (United States)

    Frigm, Ryan C.; Hejduk, Matthew D.; Johnson, Lauren C.; Plakalovic, Dragan

    2015-01-01

    On-orbit collision risk is becoming an increasing mission risk to all operational satellites in Earth orbit. Managing this risk can be disruptive to mission and operations, present challenges for decision-makers, and is time-consuming for all parties involved. With the planned capability improvements to detecting and tracking smaller orbital debris and capacity improvements to routinely predict on-orbit conjunctions, this mission risk will continue to grow in terms of likelihood and effort. It is very real possibility that the future space environment will not allow collision risk management and mission operations to be conducted in the same manner as it is today. This paper presents the concept of a finite conjunction assessment-one where each discrete conjunction is not treated separately but, rather, as a continuous event that must be managed concurrently. The paper also introduces the Total Probability of Collision as an analogous metric for finite conjunction assessment operations and provides several options for its usage in a Concept of Operations.

  1. Physical method to assess a probable maximum precipitation, using CRCM datas

    International Nuclear Information System (INIS)

    Beauchamp, J.

    2009-01-01

    'Full text:' For Nordic hydropower facilities, spillways are designed with a peak discharge based on extreme conditions. This peak discharge is generally derived using the concept of a probable maximum flood (PMF), which results from the combined effect of abundant downpours (probable maximum precipitation - PMP) and rapid snowmelt. On a gauged basin, the weather data record allows for the computation of the PMF. However, uncertainty in the future climate raises questions as to the accuracy of current PMP estimates for existing and future hydropower facilities. This project looks at the potential use of the Canadian Regional Climate Model (CRCM) data to compute the PMF in ungauged basins and to assess potential changes to the PMF in a changing climate. Several steps will be needed to accomplish this task. This paper presents the first step that aims at applying/adapting to CRCM data the in situ moisture maximization technique developed by the World Meteorological Organization, in order to compute the PMP at the watershed scale. The CRCM provides output data on a 45km grid at a six hour time step. All of the needed atmospheric data is available at sixteen different pressure levels. The methodology consists in first identifying extreme precipitation events under current climate conditions. Then, a maximum persisting twelve hours dew point is determined at each grid point and pressure level for the storm duration. Afterwards, the maximization ratio is approximated by merging the effective temperature with dew point and relative humidity values. The variables and maximization ratio are four-dimensional (x, y, z, t) values. Consequently, two different approaches are explored: a partial ratio at each step and a global ratio for the storm duration. For every identified extreme precipitation event, a maximized hyetograph is computed from the application of this ratio, either partial or global, on CRCM precipitation rates. Ultimately, the PMP is the depth of the

  2. Physical method to assess a probable maximum precipitation, using CRCM datas

    Energy Technology Data Exchange (ETDEWEB)

    Beauchamp, J. [Univ. de Quebec, Ecole de technologie superior, Quebec (Canada)

    2009-07-01

    'Full text:' For Nordic hydropower facilities, spillways are designed with a peak discharge based on extreme conditions. This peak discharge is generally derived using the concept of a probable maximum flood (PMF), which results from the combined effect of abundant downpours (probable maximum precipitation - PMP) and rapid snowmelt. On a gauged basin, the weather data record allows for the computation of the PMF. However, uncertainty in the future climate raises questions as to the accuracy of current PMP estimates for existing and future hydropower facilities. This project looks at the potential use of the Canadian Regional Climate Model (CRCM) data to compute the PMF in ungauged basins and to assess potential changes to the PMF in a changing climate. Several steps will be needed to accomplish this task. This paper presents the first step that aims at applying/adapting to CRCM data the in situ moisture maximization technique developed by the World Meteorological Organization, in order to compute the PMP at the watershed scale. The CRCM provides output data on a 45km grid at a six hour time step. All of the needed atmospheric data is available at sixteen different pressure levels. The methodology consists in first identifying extreme precipitation events under current climate conditions. Then, a maximum persisting twelve hours dew point is determined at each grid point and pressure level for the storm duration. Afterwards, the maximization ratio is approximated by merging the effective temperature with dew point and relative humidity values. The variables and maximization ratio are four-dimensional (x, y, z, t) values. Consequently, two different approaches are explored: a partial ratio at each step and a global ratio for the storm duration. For every identified extreme precipitation event, a maximized hyetograph is computed from the application of this ratio, either partial or global, on CRCM precipitation rates. Ultimately, the PMP is the depth of the

  3. A Computer Program for Assessing Nuclear Safety Culture Impact

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  4. 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.

  5. Reliability considerations of NDT by probability of detection (POD). Determination using ultrasound phased array. Results from a project in frame of the German nuclear safety research program

    International Nuclear Information System (INIS)

    Kurz, Jochen H.; Dugan, Sandra; Juengert, Anne

    2013-01-01

    Reliable assessment procedures are an important aspect of maintenance concepts. Non-destructive testing (NDT) methods are an essential part of a variety of maintenance plans. Fracture mechanical assessments require knowledge of flaw dimensions, loads and material parameters. NDT methods are able to acquire information on all of these areas. However, it has to be considered that the level of detail information depends on the case investigated and therefore on the applicable methods. Reliability aspects of NDT methods are of importance if quantitative information is required. Different design concepts e.g. the damage tolerance approach in aerospace already include reliability criteria of NDT methods applied in maintenance plans. NDT is also an essential part during construction and maintenance of nuclear power plants. In Germany, type and extent of inspection are specified in Safety Standards of the Nuclear Safety Standards Commission (KTA). Only certified inspections are allowed in the nuclear industry. The qualification of NDT is carried out in form of performance demonstrations of the inspection teams and the equipment, witnessed by an authorized inspector. The results of these tests are mainly statements regarding the detection capabilities of certain artificial flaws. In other countries, e.g. the U.S., additional blind tests on test blocks with hidden and unknown flaws may be required, in which a certain percentage of these flaws has to be detected. The knowledge of the probability of detection (POD) curves of specific flaws in specific testing conditions is often not present. This paper shows the results of a research project designed for POD determination of ultrasound phased array inspections of real and artificial cracks. The continuative objective of this project was to generate quantitative POD results. The distribution of the crack sizes of the specimens and the inspection planning is discussed, and results of the ultrasound inspections are presented. In

  6. A generalized framework for assessment of safety margins in nuclear power plants

    International Nuclear Information System (INIS)

    Gavrilas, M.; Youngblood, B.; Prelewicz, D.; Meyer, Jim

    2004-01-01

    The protection of public health and safety, and the environment from inadvertent releases of radioactive materials from nuclear power plants relies on the implementation of the defense-in-depth strategy. The term defense-in-depth evolved historically, and thus its application has not always been uniform. The use of the term in the context of the U.S. Nuclear Regulatory Commission (NRC) safety philosophy entails the reliance of a nuclear facility on successive compensatory measures in preventing accidents or mitigating damage caused by malfunctions, accidents, or naturally occurring events. The introduction of probabilistic risk analyses with NUREG-74/014 and subsequent evolution in risk assessment techniques, are leading to the implementation of risk informed regulation to ensure the safety of the public and the environment. Risk informed regulation minimizes the likelihood of overlooking potentially significant accident sequences while limiting unnecessary burdens imposed on licensees. The proposed framework merges fundamental elements of safety regulation: defense-in depth, safety margins and probabilistic risk. It formalizes the relationship between probabilistic risk assessment (PRA) methods and data, and deterministic analyses in a manner consistent with NRC's defense-in-depth philosophy. Succinctly put, the likelihood and consequences of accident scenarios are considered simultaneously and quantified by a plant safety metric. The integration of these fundamental elements into a practically applicable safety framework is consistent with the NRC policy statement on use of probabilistic risk assessment methods and the November 2002 Regulatory Guide on risk informed decisions on plant-specific changes to the licensing basis. Safety information resulting from the application of the framework supersedes traditional safety figures of merit. Safety quantifiers, referred herein as safety indices, expand on the qualifier outcomes that currently accompany fault tree

  7. Pedagogical Utilization and Assessment of the Statistic Online Computational Resource in Introductory Probability and Statistics Courses.

    Science.gov (United States)

    Dinov, Ivo D; Sanchez, Juana; Christou, Nicolas

    2008-01-01

    Technology-based instruction represents a new recent pedagogical paradigm that is rooted in the realization that new generations are much more comfortable with, and excited about, new technologies. The rapid technological advancement over the past decade has fueled an enormous demand for the integration of modern networking, informational and computational tools with classical pedagogical instruments. Consequently, teaching with technology typically involves utilizing a variety of IT and multimedia resources for online learning, course management, electronic course materials, and novel tools of communication, engagement, experimental, critical thinking and assessment.The NSF-funded Statistics Online Computational Resource (SOCR) provides a number of interactive tools for enhancing instruction in various undergraduate and graduate courses in probability and statistics. These resources include online instructional materials, statistical calculators, interactive graphical user interfaces, computational and simulation applets, tools for data analysis and visualization. The tools provided as part of SOCR include conceptual simulations and statistical computing interfaces, which are designed to bridge between the introductory and the more advanced computational and applied probability and statistics courses. In this manuscript, we describe our designs for utilizing SOCR technology in instruction in a recent study. In addition, present the results of the effectiveness of using SOCR tools at two different course intensity levels on three outcome measures: exam scores, student satisfaction and choice of technology to complete assignments. Learning styles assessment was completed at baseline. We have used three very different designs for three different undergraduate classes. Each course included a treatment group, using the SOCR resources, and a control group, using classical instruction techniques. Our findings include marginal effects of the SOCR treatment per individual

  8. Method for assessing the probability of accumulated doses from an intermittent source using the convolution technique

    International Nuclear Information System (INIS)

    Coleman, J.H.

    1980-10-01

    A technique is discussed for computing the probability distribution of the accumulated dose received by an arbitrary receptor resulting from several single releases from an intermittent source. The probability density of the accumulated dose is the convolution of the probability densities of doses from the intermittent releases. Emissions are not assumed to be constant over the brief release period. The fast fourier transform is used in the calculation of the convolution

  9. Preliminary investigation on reliability assessment of passive safety system

    International Nuclear Information System (INIS)

    Huang Changfan; Kuang Bo

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Alujevic, Luka

    2014-01-01

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

  11. Suggestions on the Development of Safety Culture Assessment Method

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  12. Improvement of safety approach for accident during operation of LILW disposal facility: Application for operational safety assessment of the near-surface LILW disposal facility in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Joo; Kim, Min Seong; Park, Jin Beak [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2017-06-15

    To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classifcation logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper.

  13. Improvement of safety approach for accident during operation of LILW disposal facility: Application for operational safety assessment of the near-surface LILW disposal facility in Korea

    International Nuclear Information System (INIS)

    Kim, Hyun Joo; Kim, Min Seong; Park, Jin Beak

    2017-01-01

    To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classifcation logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper

  14. The role of natural analogues in safety assessment and acceptability

    International Nuclear Information System (INIS)

    Papp, Toenis

    1987-01-01

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

  15. Safety assessment document for the Dynamic Test Complex B854

    International Nuclear Information System (INIS)

    Odell, B.N.; Pfeifer, H.E.

    1981-01-01

    A safety assessment was performed to determine if potential accidents at the 854 Complex at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The credible accidents that might have an effect on these facilities or have off-site consequences were considered. These were earthquake, extreme wind (including missiles), lightning, flood, criticality, high explosive (HE) detonation that disperses uranium and beryllium, spontaneous oxidation of plutonium, explosions due to finely divided particles, and a fire. Seismic and extreme wind (including missiles) analyses indicate that the buildings are basically sound. The lightning protection system is in the process of being upgraded to meet AMCR 385-100. These buildings are located high above the dry creek bed so that a flood is improbable. The probability of high explosive detonation involving plutonium is very remote since the radioactive materials are encased and plutonium and HE are not permitted concurrently in the same area at Site 300. (The exception to this policy is that explosive actuating devices are sometimes located in assemblies containing fissile materials. However, an accidental actuation will not affect the safe containment of the plutonium within the assembly.) There is a remote possibility of an HE explosion involving uranium and beryllium since these are permitted in the same area.The possibility of a criticality accident is very remote since the fissile materials are doubly encased in stout metal containers. All operations involving these materials are independently reviewed and inspected by the Criticality Safety Office. It was determined that a fire was unlikely due to the low fire loading and the absence of ignition sources. It was also determined that the consequences of any accidents were reduced by the remote location of these facilities, their design, and by administrative controls

  16. Ultraviolet safety assessments of insect light traps

    OpenAIRE

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

    2016-01-01

    ABSTRACT Near-ultraviolet (UV-A: 315?400?nm), ?black-light,? electric lamps were invented in 1935 and ultraviolet insect light traps (ILTs) were introduced for use in agriculture around that time. Today ILTs are used indoors in several industries and in food-service as well as in outdoor settings. With recent interest in photobiological lamp safety, safety standards are being developed to test for potentially hazardous ultraviolet emissions. A variety of UV ?Black-light? ILTs were measured at...

  17. Self-assessment of operational safety for nuclear power plants

    International Nuclear Information System (INIS)

    1999-12-01

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

  18. Safety assessment as basis for the decision making process

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Syarip; Hauptmanns, U.

    2000-01-01

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

  20. Application of a few orthogonal polynomials to the assessment of the fracture failure probability of a spherical tank

    International Nuclear Information System (INIS)

    Cao Tianjie; Zhou Zegong

    1993-01-01

    This paper presents some methods to assess the fracture failure probability of a spherical tank. These methods convert the assessment of the fracture failure probability into the calculation of the moment of cracks and a one-dimensional integral. In the paper, we first derive series' formulae to calculation the moments of cracks on the occasion of the crack fatigue growth and the moments of crack opening displacements according to JWES-2805 code. We then use the first n moments of crack opening displacements and a few orthogonal polynomials to compose the probability density function of the crack opening displacement. Lastly, the fracture failure probability is obtained according to the interference theory. An example proves that these methods are simpler, quicker, and more accurate. At the same time, these methods avoid the disadvantage of Edgeworth's series method. (author)

  1. Nirex safety assessment research programme: 1987/88

    International Nuclear Information System (INIS)

    George, D.; Hodgkinson, D.P.

    1987-01-01

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

  2. Epigenetics and cancer: implications for drug discovery and safety assessment

    International Nuclear Information System (INIS)

    Moggs, Jonathan G.; Goodman, Jay I.; Trosko, James E.; Roberts, Ruth A.

    2004-01-01

    It is necessary to determine whether chemicals or drugs have the potential to pose a threat to human health. Research conducted over the last two decades has led to the paradigm that chemicals can cause cancer either by damaging DNA or by altering cellular growth, probably via receptor-mediated changes in gene expression. However, recent evidence suggests that gene expression can be altered markedly via several diverse epigenetic mechanisms that can lead to permanent or reversible changes in cellular behavior. Key molecular events underlying these mechanisms include the alteration of DNA methylation and chromatin, and changes in the function of cell surface molecules. Thus, for example, DNA methyltransferase enzymes together with chromatin-associated proteins such as histone modifying enzymes and remodelling factors can modify the genetic code and contribute to the establishment and maintenance of altered epigenetic states. This is relevant to many types of toxicity including but not limited to cancer. In this paper, we describe the potential for interplay between genetic alteration and epigenetic changes in cell growth regulation and discuss the implications for drug discovery and safety assessment

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  4. Procedures for self-assessment of operational safety

    International Nuclear Information System (INIS)

    1997-08-01

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

  5. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  6. Visualization of Safety Assessment Result Using GIS in SITES

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

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

    1985-07-01

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

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

    International Nuclear Information System (INIS)

    Ebashi, Takeshi; Kaku, Kenichi; Ishiguro, Katsuhiko

    2012-01-01

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

  9. Seismic and tsunami safety margin assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

  10. Safety assessment in primary Mycobacterium tuberculosis smear ...

    African Journals Online (AJOL)

    Introduction Tuberculosis (TB) is caused by Mycobacterium tuberculosis and is transmitted mainly through aerosolization of infected sputum which puts laboratory workers at risk in spite of the laboratory workersf risk of infection being at 3 to 9 times higher than the general public. Laboratory safety should therefore be ...

  11. Seismic and tsunami safety margin assessment

    International Nuclear Information System (INIS)

    2013-01-01

    Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

  12. Application of probability distributions for quantifying uncertainty in radionuclide source terms for Seabrook risk assessment

    International Nuclear Information System (INIS)

    Walker, D.H.; Savin, N.L.

    1985-01-01

    The calculational models developed for the Reactor Safety Study (RSS) have traditionally been used to generate 'point estimate values' for radionuclide release to the environment for nuclear power plant risk assessments. The point estimate values so calculated are acknowledged by most knowledgeable individuals to be conservatively high. Further, recent evaluations of the overall uncertainties in the various components that make up risk estimates for nuclear electric generating stations show that one of the large uncertainties is associated with the magnitude of the radionuclide release to the environment. In the approach developed for the RSS, values for fission product release from the fuel are derived from data obtained from small experiments. A reappraisal of the RSS release fractions was published in 1981 in NUREG-0772. Estimates of fractional releases from fuel are similar to those of the RSS. In the RSS approach, depletion during transport from the core (where the fission products are released) to the containment is assumed to be zero for calculation purposes. In the containment, the CORRAL code is applied to calculate radioactivity depletion by containment processes and to calculate the quantity and timing of release to the environment

  13. Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

    International Nuclear Information System (INIS)

    Frid, W.; Sandervaag, O.

    1997-01-01

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

  14. Complementary assessment of the safety of French nuclear power plants

    International Nuclear Information System (INIS)

    Camarcat, N.; Pouget-Abadie, X.

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  18. Assessing propensity to learn from safety-related events

    NARCIS (Netherlands)

    Drupsteen, L.; Wybo, J.L.

    2015-01-01

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

  19. 76 FR 45453 - New Car Assessment Program (NCAP); Safety Labeling

    Science.gov (United States)

    2011-07-29

    ... Testing F. Consumer Survey and Label Research G. Other Issues 1. Legend for Star Ratings 2. Overall... consumers to consider and assess the relative safety of vehicles before deciding which new vehicle they want..., and prominent fashion, and that the safety rating area of the Monroney label meet minimum size...

  20. Assessing the safety of the road network : a simple method.

    NARCIS (Netherlands)

    Janssen, S.T.M.C.

    1993-01-01

    Research by the SWOV Institute for Road Safety Research in The Netherlands assessing the safety of the Dutch road network is presented. In The Netherlands urban and rural roads and motorways are used roughly equally, however accident rates vary greatly. These are discussed in this paper. Tables show

  1. Incidents in nuclear research reactor examined by deterministic probability and probabilistic safety analysis; Incidentes em reatores nucleares de pesquisa examinados por analise de probabilidade deterministica e analise probabilistica de seguranca

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Valdir Maciel

    2010-07-01

    This study aims to evaluate the potential risks submitted by the incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency, IAEA, were used, the Incident Report System for Research Reactor and Research Reactor Data Base. For this type of assessment was used the Probabilistic Safety Analysis (PSA), within a confidence level of 90% and the Deterministic Probability Analysis (DPA). To obtain the results of calculations of probabilities for PSA, were used the theory and equations in the paper IAEA TECDOC - 636. The development of the calculations of probabilities for PSA was used the program Scilab version 5.1.1, free access, executable on Windows and Linux platforms. A specific program to get the results of probability was developed within the main program Scilab 5.1.1., for two distributions Fischer and Chi-square, both with the confidence level of 90%. Using the Sordi equations and Origin 6.0 program, were obtained the maximum admissible doses related to satisfy the risk limits established by the International Commission on Radiological Protection, ICRP, and were also obtained these maximum doses graphically (figure 1) resulting from the calculations of probabilities x maximum admissible doses. It was found that the reliability of the results of probability is related to the operational experience (reactor x year and fractions) and that the larger it is, greater the confidence in the outcome. Finally, a suggested list of future work to complement this paper was gathered. (author)

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

    NARCIS (Netherlands)

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

    1995-01-01

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

  3. Development of Safety Assessment Information System (SAIS)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Byung Shik; Lee, Kyung Jin; Lee, Byung Chul [FNC Tech. Co. Ltd. SNU, Seoul (Korea, Republic of); Song, Tae Young; Lee, Chang Ho [KHNP, Daejeon (Korea, Republic of)

    2007-10-15

    Many reports and documents about nuclear power plant safety analysis like a Periodic Safe Review (PSR), Periodic Safety Analysis (PSA) and Severe Accident Management Guideline (SAMG) come to be drawn up from KHNP. Since these are not arranged easy to look up, the systematic arrangement of data was necessary. The solution against hereupon is to store database, and it was developed with the name, SAIS, by FNC Tech. Co. together with NETEC KHNP. In this web program it is easy to manage (registration, search and statistics) data. And the authorized user can approach this system. This was developed, and was verified under the development environment of; - Web Server : Apache 2.2.5 - Program Language : PHP 5.2 - DBMS : Oracle 10g.

  4. Development of Safety Assessment Information System (SAIS)

    International Nuclear Information System (INIS)

    Park, Byung Shik; Lee, Kyung Jin; Lee, Byung Chul; Song, Tae Young; Lee, Chang Ho

    2007-01-01

    Many reports and documents about nuclear power plant safety analysis like a Periodic Safe Review (PSR), Periodic Safety Analysis (PSA) and Severe Accident Management Guideline (SAMG) come to be drawn up from KHNP. Since these are not arranged easy to look up, the systematic arrangement of data was necessary. The solution against hereupon is to store database, and it was developed with the name, SAIS, by FNC Tech. Co. together with NETEC KHNP. In this web program it is easy to manage (registration, search and statistics) data. And the authorized user can approach this system. This was developed, and was verified under the development environment of; - Web Server : Apache 2.2.5 - Program Language : PHP 5.2 - DBMS : Oracle 10g

  5. Safety Assessment of Radioactive waste Repositories

    International Nuclear Information System (INIS)

    1991-01-01

    It is planned to dispose of high-level radioactive wastes in deep geological formations. To access the long-term safety of radioactive waste disposal systems, mathematical models are used to describe groundwater flow, chemistry and potential radionuclide migration through these formations. Establishing the validity of such models is important in order to obtain the necessary confidence in the safety of the disposal method. The papers in these proceedings of the GEOVAL'90 Symposium describe the current state of knowledge on the validation of geosphere flow and transport models. This symposium, divided into five sessions, contains 65 technical papers: session 1 - Necessity of validation. Session 2 - Progress in validation of flow and transport models in orystalline rock, unsaturated media, salt media or clay. Session 3 - Progress in validation of geochemical models. Session 4 - Progress in validation of coupled thermo-hydro-mechanical effects. Session 5 - Validation strategy

  6. Safety assessment of the liquid-fed ceramic melter process

    International Nuclear Information System (INIS)

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

    1980-08-01

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

  7. 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

  8. Ensuring the quality of occupational safety risk assessment.

    Science.gov (United States)

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

    2013-03-01

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

  9. Data used for safety assessment of reprocessing facilities

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Suzuki, Atsuyuki; Kanagawa, Akira

    1990-08-01

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

  10. Re-assessment of safety and ageing

    International Nuclear Information System (INIS)

    Boudot, R.; Dallery, D.; Delmas, J.; Balley, J.; Acalet, R.; Bignan, G.; Mergui, C.; Mazoyer, R.; Maegey, M.; Laigneau, P.

    2007-01-01

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

  11. Modern Aspects of Safety Assessment of Foodstuff

    Directory of Open Access Journals (Sweden)

    Tetiana Chorna

    2018-04-01

    Full Text Available Food safety is one of the decisive components of the economic security of each state and is determined by the ability of the country to control effectively the production and import of safe and high-quality food on the generally accepted principles of the world. This sphere of activity in human society has extremely important humanitarian, social, economic and political aspects. The food raw materials and food products quality and safety control is currently the most relevant analytical task. It is more important than environmental pollution, according to some data, more than 70 % of harmful pollutants in the human body gets through food, 20% of water and 10 % of the air. Technogenic pollution of the environment through soil, water and air gets directly into the food. However, food products are contaminated with natural harmful substances that appear in improper storage, in violation of technologies, food processing and processing charts. The article is devoted to the main factors analysis influencing the safety of food products and the improvement of instrumental methods for the study of quality aromatic products (for example, coffee.

  12. Assessment of public safety around EDF facilities

    Energy Technology Data Exchange (ETDEWEB)

    Poupart, M. [Electricite de France, Paris (France)

    2004-09-01

    Electricite de France (EDF) recognizes that a dam's structural resistance and its capacity to withstand heavy flooding are 2 of the most significant safety aspects for hydroelectric power stations. However, in addition to dam failure, there are safety risks for the public who frequent the rivers up and down stream from the dam, as well as on property and the environment. A fatal accident which occurred in 1995 down river from EDF's Monteynard hydroelectric facility on the Drac River prompted the utility to take measures to improve control over this type of hazard. Collaboration with public authorities led to an action plan to educate the public about possible danger areas and to improve methods of reducing risks. Regulations regarding access to these areas were also studied along with ways of informing and warning the public. All the stretches of river directly above and below the power stations and dams were listed systematically and a qualitative analysis was carried out of any possible dangers. This led to changes in operating rules, technical instructions and power plant operating regulations. Certain areas are designated as prohibited areas, such as places subject to hazards caused by violent and unexpected water discharges. This paper outlined the Hydraulic Safety Significant Event (HSSE) classification which relates to an operational event related to water that is liable to affect human beings, the environment, water level or flow rate. 9 figs.

  13. Risk assessment of safety violations for coal mines

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-15

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

  14. Use of agent based simulation for traffic safety assessment

    CSIR Research Space (South Africa)

    Conradie, Dirk CU

    2008-07-01

    Full Text Available This paper describes the development of an agent based Computational Building Simulation (CBS) tool, termed KRONOS that is being used to work on advanced research questions such as traffic safety assessment and user behaviour in buildings...

  15. NANoREG framework for the safety assessment of nanomaterials

    OpenAIRE

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

    2017-01-01

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

  16. Selected component failure rate values from fusion safety assessment tasks

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1998-01-01

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

  17. Selected component failure rate values from fusion safety assessment tasks

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, L.C.

    1998-09-01

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

  18. Selected Component Failure Rate Values from Fusion Safety Assessment Tasks

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles

    1998-09-01

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

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

    International Nuclear Information System (INIS)

    1989-10-01

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

  20. Assessing progress in the development of safety culture

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    Science.gov (United States)

    Ancel, Ersin; Shih, Ann T.

    2014-01-01

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

  2. Savannah River Site K-Reactor Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  3. AP1000 Containment Design and Safety Assessment

    International Nuclear Information System (INIS)

    Wright, Richard F.; Ofstun, Richard P.; Bachere, Sebastien

    2002-01-01

    The AP1000 is an up-rated version of the AP600 passive plant design that recently received final design certification from the US NRC. Like AP600, the AP1000 is a two-loop, pressurized water reactor featuring passive core cooling and passive containment safety systems. One key safety feature of the AP1000 is the passive containment cooling system which maintains containment integrity in the event of a design basis accident. This system utilizes a high strength, steel containment vessel inside a concrete shield building. In the event of a pipe break inside containment, a high pressure signal actuates valves which allow water to drain from a storage tank atop the shield building. Water is applied to the top of the containment shell, and evaporates, thereby removing heat. An air flow path is formed between the shield building and the containment to aid in the evaporation and is exhausted through a chimney at the top of the shield building. Extensive testing and analysis of this system was performed as part of the AP600 design certification process. The AP1000 containment has been designed to provide increased safety margin despite the increased reactor power. The containment volume was increased to accommodate the larger steam generators, and to provide increased margin for containment pressure response to design basis events. The containment design pressure was increased from AP600 by increasing the shell thickness and by utilizing high strength steel. The passive containment cooling system water capacity has been increased and the water application rate has been scaled to the higher decay heat level. The net result is higher margins to the containment design pressure limit than were calculated for AP600 for all design basis events. (authors)

  4. Assessment of clinical utility of 18F-FDG PET in patients with head and neck cancer: a probability analysis

    International Nuclear Information System (INIS)

    Goerres, Gerhard W.; Mosna-Firlejczyk, Katarzyna; Schulthess, Gustav K. von; Steurer, Johann; Bachmann, Lucas M.

    2003-01-01

    The purpose of this study was to calculate disease probabilities based on data of patients with head and neck cancer in the register of our institution and to perform a systematic review of the available data on the accuracy of PET in the primary assessment and follow-up of patients with head and neck cancer. The pre-test probability of head and neck cancer among patients in our institutional data registry was assessed. Then the published literature was selected and appraised according to a standard protocol of systematic reviews. Two reviewers independently selected and extracted data on study characteristics, quality and accuracy. Accuracy data were used to form 2 x 2 contingency tables and were pooled to produce summary receiver operating characteristic (ROC) curves and summary likelihood ratios for positive and negative testing. Finally post-test probabilities were calculated on the basis of the pre-test probabilities of this patient group. All patients had cytologically or histologically proven cancer. The prevalence of additional lymph node metastases on PET in staging examinations was 19.6% (11/56), and that of locoregional recurrence on restaging PET was 28.6% (12/42). In the primary assessment of patients, PET had positive and negative likelihood ratios of 3.9 (2.56-5.93) and 0.24 (0.14-0.41), respectively. Disease probabilities were therefore 49.4% for a positive test result and 5.7% for a negative test result. In the assessment of recurrence these values were 3.96 (2.8-5.6) and 0.16 (0.1-0.25), resulting in probabilities of 49.7% and 3.8%. PET evaluation for involvement of lymph nodes had positive and negative likelihood ratios of 17.26 (10.9-27.3) and 0.19 (0.13-0.27) for primary assessment and 11.0 (2.93-41.24) and 0.14 (0.01-1.88) for detection of recurrence. The probabilities were 81.2% and 4.5% for primary assessment and 73.3% and 3.4% for assessment of recurrence. It is concluded that in this clinical setting the main advantage of PET is the

  5. Guidelines for pressure vessel safety assessment

    Science.gov (United States)

    Yukawa, S.

    1990-04-01

    A technical overview and information on metallic pressure containment vessels and tanks is given. The intent is to provide Occupational Safety and Health Administration (OSHA) personnel and other persons with information to assist in the evaluation of the safety of operating pressure vessels and low pressure storage tanks. The scope is limited to general industrial application vessels and tanks constructed of carbon or low alloy steels and used at temperatures between -75 and 315 C (-100 and 600 F). Information on design codes, materials, fabrication processes, inspection and testing applicable to the vessels and tanks are presented. The majority of the vessels and tanks are made to the rules and requirements of ASME Code Section VIII or API Standard 620. The causes of deterioration and damage in operation are described and methods and capabilities of detecting serious damage and cracking are discussed. Guidelines and recommendations formulated by various groups to inspect for the damages being found and to mitigate the causes and effects of the problems are presented.

  6. LANL Safety Conscious Work Environment (SCWE) Self-Assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-29

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

  7. Research on advanced system safety assessment procedures (4)

    International Nuclear Information System (INIS)

    Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-03-01

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

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

    International Nuclear Information System (INIS)

    Xiang Yuanyuan; Chen Xukun; Xu Rongbin

    2012-01-01

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

  9. Development of risk assessment simulation tool for optimal control of a low probability-high consequence disaster

    International Nuclear Information System (INIS)

    Yotsumoto, Hiroki; Yoshida, Kikuo; Genchi, Hiroshi

    2011-01-01

    In order to control low probability-high consequence disaster which causes huge social and economic damage, it is necessary to develop simultaneous risk assessment simulation tool based on the scheme of disaster risk including diverse effects of primary disaster and secondary damages. We propose the scheme of this risk simulation tool. (author)

  10. Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S; Streit, R D; Chou, C K

    1980-01-01

    This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10{sup -12}). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

  11. Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

    International Nuclear Information System (INIS)

    Lu, S.; Streit, R.D.; Chou, C.K.

    1980-01-01

    This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10 -12 ). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

  12. Uncertainty and sensitivity analysis on probabilistic safety assessment of an experimental facility

    International Nuclear Information System (INIS)

    Burgazzi, L.

    2000-01-01

    The aim of this work is to perform an uncertainty and sensitivity analysis on the probabilistic safety assessment of the International Fusion Materials Irradiation Facility (IFMIF), in order to assess the effect on the final risk values of the uncertainties associated with the generic data used for the initiating events and component reliability and to identify the key quantities contributing to this uncertainty. The analysis is conducted on the expected frequency calculated for the accident sequences, defined through the event tree (ET) modeling. This is in order to increment credit to the ET model quantification, to calculate frequency distributions for the occurrence of events and, consequently, to assess if sequences have been correctly selected on the probability standpoint and finally to verify the fulfillment of the safety conditions. Uncertainty and sensitivity analysis are performed using respectively Monte Carlo sampling and an importance parameter technique. (author)

  13. Survey of probabilistic methods in safety and risk assessment for nuclear power plant licensing

    International Nuclear Information System (INIS)

    1984-04-01

    After an overview about the goals and general methods of probabilistic approaches in nuclear safety the main features of probabilistic safety or risk assessment (PRA) methods are discussed. Mostly in practical applications not a full-fledged PRA is applied but rather various levels of analysis leading from unavailability assessment of systems over the more complex analysis of the probable core damage stages up to the assessment of the overall health effects on the total population from a certain practice. The various types of application are discussed in relation to their limitation and benefits for different stages of design or operation of nuclear power plants. This gives guidance for licensing staff to judge the usefulness of the various methods for their licensing decisions. Examples of the application of probabilistic methods in several countries are given. Two appendices on reliability analysis and on containment and consequence analysis provide some more details on these subjects. (author)

  14. Training courses on integrated safety assessment modelling for waste repositories

    International Nuclear Information System (INIS)

    Mallants, D.

    2007-01-01

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

  15. Neuropsychological assessment of driving safety risk in older adults with and without neurologic disease.

    Science.gov (United States)

    Anderson, Steven W; Aksan, Nazan; Dawson, Jeffrey D; Uc, Ergun Y; Johnson, Amy M; Rizzo, Matthew

    2012-01-01

    Decline in cognitive abilities can be an important contributor to the driving problems encountered by older adults, and neuropsychological assessment may provide a practical approach to evaluating this aspect of driving safety risk. The purpose of the present study was to evaluate several commonly used neuropsychological tests in the assessment of driving safety risk in older adults with and without neurological disease. A further goal of this study was to identify brief combinations of neuropsychological tests that sample performances in key functional domains and thus could be used to efficiently assess driving safety risk. A total of 345 legally licensed and active drivers over the age of 50, with no neurologic disease (N = 185), probable Alzheimer's disease (N = 40), Parkinson's disease (N = 91), or stroke (N = 29), completed vision testing, a battery of 10 neuropsychological tests, and an 18-mile drive on urban and rural roads in an instrumented vehicle. Performances on all neuropsychological tests were significantly correlated with driving safety errors. Confirmatory factor analysis was used to identify 3 key cognitive domains assessed by the tests (speed of processing, visuospatial abilities, and memory), and several brief batteries consisting of one test from each domain showed moderate corrected correlations with driving performance. These findings are consistent with the notion that driving places demands on multiple cognitive abilities that can be affected by aging and age-related neurological disease, and that neuropsychological assessment may provide a practical off-road window into the functional status of these cognitive systems.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

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

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

    International Nuclear Information System (INIS)

    1990-04-01

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

  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. Probabilistic assessment of NPP safety under aircraft impact

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  2. Assessment of the factors with significant influence on safety culture

    International Nuclear Information System (INIS)

    Farcasiu, M.; Nitoi, M.

    2013-01-01

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

  3. Safety assessment of existing highway bridges and viaducts

    NARCIS (Netherlands)

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

    2012-01-01

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

  4. Risk assessment and safety regulations in offshore oil and gas ...

    African Journals Online (AJOL)

    Risk management of which risk assessment is part, and safety regulations are common in the offshore oil and gas industry management system. The process of conducting risk assessment is mostly a challenge for operational personnel assigned to perform this function. The most significant problem is the decision to use ...

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

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1992-01-01

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

  6. Assessment of safety regulation using an artificial society

    International Nuclear Information System (INIS)

    Furuta, Kazuo; Nagase, Masaya

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. Probability of Detection Study to Assess the Performance of Nondestructive Inspection Methods for Wind Turbine Blades.

    Energy Technology Data Exchange (ETDEWEB)

    Roach, Dennis P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rice, Thomas M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Paquette, Joshua [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-07-01

    Wind turbine blades pose a unique set of inspection challenges that span from very thick and attentive spar cap structures to porous bond lines, varying core material and a multitude of manufacturing defects of interest. The need for viable, accurate nondestructive inspection (NDI) technology becomes more important as the cost per blade, and lost revenue from downtime, grows. NDI methods must not only be able to contend with the challenges associated with inspecting extremely thick composite laminates and subsurface bond lines, but must also address new inspection requirements stemming from the growing understanding of blade structural aging phenomena. Under its Blade Reliability Collaborative program, Sandia Labs quantitatively assessed the performance of a wide range of NDI methods that are candidates for wind blade inspections. Custom wind turbine blade test specimens, containing engineered defects, were used to determine critical aspects of NDI performance including sensitivity, accuracy, repeatability, speed of inspection coverage, and ease of equipment deployment. The detection of fabrication defects helps enhance plant reliability and increase blade life while improved inspection of operating blades can result in efficient blade maintenance, facilitate repairs before critical damage levels are reached and minimize turbine downtime. The Sandia Wind Blade Flaw Detection Experiment was completed to evaluate different NDI methods that have demonstrated promise for interrogating wind blades for manufacturing flaws or in-service damage. These tests provided the Probability of Detection information needed to generate industry-wide performance curves that quantify: 1) how well current inspection techniques are able to reliably find flaws in wind turbine blades (industry baseline) and 2) the degree of improvements possible through integrating more advanced NDI techniques and procedures. _____________ S a n d i a N a t i o n a l L a b o r a t o r i e s i s a m u l t i

  9. Criticality safety evaluations - a open-quotes stalking horseclose quotes for integrated safety assessment

    International Nuclear Information System (INIS)

    Williams, R.A.

    1995-01-01

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility's criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE

  10. Criticality safety evaluations - a {open_quotes}stalking horse{close_quotes} for integrated safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R.A. [Westinghouse Electric Corp., Columbia, SC (United States)

    1995-12-31

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility`s criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE.

  11. Preclosure radiological safety assessment for the ground support system in the exploratory studies facility

    International Nuclear Information System (INIS)

    Smith, A.J.; Tsai, F.C.

    1995-01-01

    An initial probabilistic safety assessment was performed for the exploratory studies facility underground opening to determine whether the ground support system should be classified as an item important to safety. The initiating event was taken to be a rock fall in an operational facility impacting a loaded waste transporter. Rock fall probability rates were estimated from data reported by commercial mining operations. This information was retrieved from the data base compiled by the Mining Safety and Health Administration from the mandatory reporting of incidents. The statistical distribution of the rock fall magnitude was estimated from the horizontal and vertical spacing fractures measured at the Yucca Mountain repository horizon. Simple models were developed to estimate container deformation and radionuclide releases arising from the projected distribution of impacts. Accepted techniques were used to calculate atmospheric dispersion and obtain the committed dose to individuals

  12. Incorporation of formal safety assessment and Bayesian network in navigational risk estimation of the Yangtze River

    International Nuclear Information System (INIS)

    Zhang, D.; Yan, X.P.; Yang, Z.L.; Wall, A.; Wang, J.

    2013-01-01

    Formal safety assessment (FSA), as a structured and systematic risk evaluation methodology, has been increasingly and broadly used in the shipping industry around the world. Concerns have been raised as to navigational safety of the Yangtze River, China's largest and the world's busiest inland waterway. Over the last few decades, the throughput of ships in the Yangtze River has increased rapidly due to the national development of the Middle and Western parts of China. Accidents such as collisions, groundings, contacts, oil-spills and fires occur repeatedly, often causing serious consequences. In order to improve the navigational safety in the Yangtze River, this paper estimates the navigational risk of the Yangtze River using the FSA concept and a Bayesian network (BN) technique. The navigational risk model is established by considering both probability and consequences of accidents with respect to a risk matrix method, followed by a scenario analysis to demonstrate the application of the proposed model

  13. Analysis of truncation limit in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Cepin, Marko

    2005-01-01

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

  14. Assessment of different models for computing the probability of a clear line of sight

    Science.gov (United States)

    Bojin, Sorin; Paulescu, Marius; Badescu, Viorel

    2017-12-01

    This paper is focused on modeling the morphological properties of the cloud fields in terms of the probability of a clear line of sight (PCLOS). PCLOS is defined as the probability that a line of sight between observer and a given point of the celestial vault goes freely without intersecting a cloud. A variety of PCLOS models assuming the cloud shape hemisphere, semi-ellipsoid and ellipsoid are tested. The effective parameters (cloud aspect ratio and absolute cloud fraction) are extracted from high-resolution series of sunshine number measurements. The performance of the PCLOS models is evaluated from the perspective of their ability in retrieving the point cloudiness. The advantages and disadvantages of the tested models are discussed, aiming to a simplified parameterization of PCLOS models.

  15. Results of the CANDU 3 probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    1995-01-01

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10 -6 /year. 4 refs., 1 fig

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

    International Nuclear Information System (INIS)

    EVANS, C.B.

    2004-01-01

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

  17. 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.

  18. 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

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

    Science.gov (United States)

    Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

    2012-07-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

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

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

    Science.gov (United States)

    Haslberger, Alexander G

    2006-05-03

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

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

    International Nuclear Information System (INIS)

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    2009-01-01

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

  6. 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

  7. Use of the t-distribution to construct seismic hazard curves for seismic probabilistic safety assessments

    Energy Technology Data Exchange (ETDEWEB)

    Yee, Eric [KEPCO International Nuclear Graduate School, Dept. of Nuclear Power Plant Engineering, Ulsan (Korea, Republic of)

    2017-03-15

    Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered.

  8. Use of the t-distribution to construct seismic hazard curves for seismic probabilistic safety assessments

    International Nuclear Information System (INIS)

    Yee, Eric

    2017-01-01

    Seismic probabilistic safety assessments are used to help understand the impact potential seismic events can have on the operation of a nuclear power plant. An important component to seismic probabilistic safety assessment is the seismic hazard curve which shows the frequency of seismic events. However, these hazard curves are estimated assuming a normal distribution of the seismic events. This may not be a strong assumption given the number of recorded events at each source-to-site distance. The use of a normal distribution makes the calculations significantly easier but may underestimate or overestimate the more rare events, which is of concern to nuclear power plants. This paper shows a preliminary exploration into the effect of using a distribution that perhaps more represents the distribution of events, such as the t-distribution to describe data. The integration of a probability distribution with potentially larger tails basically pushes the hazard curves outward, suggesting a different range of frequencies for use in seismic probabilistic safety assessments. Therefore the use of a more realistic distribution results in an increase in the frequency calculations suggesting rare events are less rare than thought in terms of seismic probabilistic safety assessment. However, the opposite was observed with the ground motion prediction equation considered

  9. Safety assessment of discharge chute isolation barrier preparation and installation

    International Nuclear Information System (INIS)

    Meichle, R.H.

    1994-01-01

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

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

    Science.gov (United States)

    Pöting, A; Schauzu, M

    2010-06-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  13. A Guide to the Application of Probability Risk Assessment Methodology and Hazard Risk Frequency Criteria as a Hazard Control for the Use of the Mobile Servicing System on the International Space Station

    Science.gov (United States)

    D'silva, Oneil; Kerrison, Roger

    2013-09-01

    A key feature for the increased utilization of space robotics is to automate Extra-Vehicular manned space activities and thus significantly reduce the potential for catastrophic hazards while simultaneously minimizing the overall costs associated with manned space. The principal scope of the paper is to evaluate the use of industry standard accepted Probability risk/safety assessment (PRA/PSA) methodologies and Hazard Risk frequency Criteria as a hazard control. This paper illustrates the applicability of combining the selected Probability risk assessment methodology and hazard risk frequency criteria, in order to apply the necessary safety controls that allow for the increased use of the Mobile Servicing system (MSS) robotic system on the International Space Station. This document will consider factors such as component failure rate reliability, software reliability, and periods of operation and dormancy, fault tree analyses and their effects on the probability risk assessments. The paper concludes with suggestions for the incorporation of existing industry Risk/Safety plans to create an applicable safety process for future activities/programs

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  16. Interim process report for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Sellin, Patrick

    2004-08-01

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

  17. Interim process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Sellin, Patrick (ed.)

    2004-08-01

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

  18. Use of standard reliability levels in design and safety assessment of in-pile loops

    International Nuclear Information System (INIS)

    Bogani, G.; Verre, A.; Balestreri, S.; Colombo, A.G.; Luisi, T.

    1975-01-01

    This paper describes a logic-probabilistic analysis technique for a critical design review and safety assessment of in-pile loops. The examples in this paper refer to the analysis performed for the experimental loops already constructed or under construction in the ESSOR reactor of the Joint Research Centre of Ispra, as irradiation facilities for fuel element research and development tests. The proposed technique is based on the classification into categories of components and protective device malfunctions. Such subdivision into categories was agreed upon by the Italian Safety Authority and Euratom JRC, and adopted for the safety assessment of the ESSOR reactor in-pile loops. For each category, the method makes a link with a corresponding malfunction probability range (probability level). This probability level is defined taking into account design, construction, inspection and maintenance criteria as well as periodic controls; therefore the quality level and consequently the reliability level are thus also defined. The analysis is developed in the following stages: (1) definition of the analysis object (top event) and drawing of the relative fault-tree; (2) loop design analysis and preliminary optimization based on logic criteria; (3) classification into categories of the fault-tree primary events; (4) final loop design analysis and optimization based on defined component quality requirements. Stages 2 and 4 are quite different since stage 2 mainly consists of a redundance optimization, while stage 4 acts on the component quality level in such a way that each minimum cut-set leading to the top has an acceptable probability level. During analysis development, use is made of computer codes which, among other things enable the verification of fault-tree logic makeup, the listing of the minimum cut-sets with and without event categorization, and the evaluation of each cut-set order. (author)

  19. Seismic performance assessment of base-isolated safety-related nuclear structures

    Science.gov (United States)

    Huang, Y.-N.; Whittaker, A.S.; Luco, N.

    2010-01-01

    Seismic or base isolation is a proven technology for reducing the effects of earthquake shaking on buildings, bridges and infrastructure. The benefit of base isolation has been presented in terms of reduced accelerations and drifts on superstructure components but never quantified in terms of either a percentage reduction in seismic loss (or percentage increase in safety) or the probability of an unacceptable performance. Herein, we quantify the benefits of base isolation in terms of increased safety (or smaller loss) by comparing the safety of a sample conventional and base-isolated nuclear power plant (NPP) located in the Eastern U.S. Scenario- and time-based assessments are performed using a new methodology. Three base isolation systems are considered, namely, (1) Friction Pendulum??? bearings, (2) lead-rubber bearings and (3) low-damping rubber bearings together with linear viscous dampers. Unacceptable performance is defined by the failure of key secondary systems because these systems represent much of the investment in a new build power plant and ensure the safe operation of the plant. For the scenario-based assessments, the probability of unacceptable performance is computed for an earthquake with a magnitude of 5.3 at a distance 7.5 km from the plant. For the time-based assessments, the annual frequency of unacceptable performance is computed considering all potential earthquakes that may occur. For both assessments, the implementation of base isolation reduces the probability of unacceptable performance by approximately four orders of magnitude for the same NPP superstructure and secondary systems. The increase in NPP construction cost associated with the installation of seismic isolators can be offset by substantially reducing the required seismic strength of secondary components and systems and potentially eliminating the need to seismically qualify many secondary components and systems. ?? 2010 John Wiley & Sons, Ltd.

  20. A proposal of safety indicators aggregation to assess the safety management effectiveness of nuclear power plants

    International Nuclear Information System (INIS)

    Carvalho, Jose Antonio B.; Saldanha, Pedro L.C.; Melo, Paulo F.F. Frutuoso e

    2009-01-01

    Safety management has changed with the evolution of management methods, named Quality Systems, moving from Quality Control, where the focus was the product, passing through Quality Assurance, which takes care of the whole manufacturing process and reaching the Total Quality Management, where policies and goals are established. Nowadays, there is a trend towards Management Systems, which integrate all different aspects related to the management of an organization (safety, environment, security, quality, costs and, etc), but it is necessary to have features to establish and assure that safety overrides the remaining aspects. The most usual way to reach this goal is to establish a policy where safety is a priority, but its implementation and the assessment of its effectiveness are no so simple. Nuclear power plants usually have over a hundred safety indicators in many processes dedicated to prevent and detect problems, although a lot of them do not evaluate these indicators in an integrated manner or point out degradation trends of organizational aspects, which can affect the plant safety. This work develops an aggregation of proactive and reactive safety indicators in order to evaluate the effectiveness of nuclear power plant safety management and to detect, at early stages, signs of process degradation or activities used to establish, maintain and assure safety conditions. The aggregation integrates indicators of the usual processes and is based on the manner the management activities have been developed in the last decades, that is: Planning, Doing, Checking and Acting - known as PDCA cycle - plus a fifth element related to the capability of those who perform safety activities. The proposed aggregation is in accordance to Brazilian standards and international recommendations and constitutes a friendly link between the top management level and the daily aspects of the organization. (author)

  1. A proposal of safety indicators aggregation to assess the safety management effectiveness of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Jose Antonio B.; Saldanha, Pedro L.C. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Coordenacao-Geral de Reatores e Ciclo Combustivel], e-mail: jantonio@cnen.gov.br, e-mail: saldanha@cnen.gov.br; Melo, Paulo F.F. Frutuoso e [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: frutuoso@con.ufrj.br

    2009-07-01

    Safety management has changed with the evolution of management methods, named Quality Systems, moving from Quality Control, where the focus was the product, passing through Quality Assurance, which takes care of the whole manufacturing process and reaching the Total Quality Management, where policies and goals are established. Nowadays, there is a trend towards Management Systems, which integrate all different aspects related to the management of an organization (safety, environment, security, quality, costs and, etc), but it is necessary to have features to establish and assure that safety overrides the remaining aspects. The most usual way to reach this goal is to establish a policy where safety is a priority, but its implementation and the assessment of its effectiveness are no so simple. Nuclear power plants usually have over a hundred safety indicators in many processes dedicated to prevent and detect problems, although a lot of them do not evaluate these indicators in an integrated manner or point out degradation trends of organizational aspects, which can affect the plant safety. This work develops an aggregation of proactive and reactive safety indicators in order to evaluate the effectiveness of nuclear power plant safety management and to detect, at early stages, signs of process degradation or activities used to establish, maintain and assure safety conditions. The aggregation integrates indicators of the usual processes and is based on the manner the management activities have been developed in the last decades, that is: Planning, Doing, Checking and Acting - known as PDCA cycle - plus a fifth element related to the capability of those who perform safety activities. The proposed aggregation is in accordance to Brazilian standards and international recommendations and constitutes a friendly link between the top management level and the daily aspects of the organization. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. Data report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    2010-12-01

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

  4. Building quality into performance and safety assessment software

    International Nuclear Information System (INIS)

    Wojciechowski, L.C.

    2011-01-01

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

  5. Safety and security risk assessments--now demystified!

    Science.gov (United States)

    White, Donald E

    2011-01-01

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

  6. Data report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

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

  7. Assessment of patient safety culture in clinical laboratories in the Spanish National Health System.

    Science.gov (United States)

    Giménez-Marín, Angeles; Rivas-Ruiz, Francisco; García-Raja, Ana M; Venta-Obaya, Rafael; Fusté-Ventosa, Margarita; Caballé-Martín, Inmaculada; Benítez-Estevez, Alfonso; Quinteiro-García, Ana I; Bedini, José Luis; León-Justel, Antonio; Torra-Puig, Montserrat

    2015-01-01

    There is increasing awareness of the importance of transforming organisational culture in order to raise safety standards. This paper describes the results obtained from an evaluation of patient safety culture in a sample of clinical laboratories in public hospitals in the Spanish National Health System. A descriptive cross-sectional study was conducted among health workers employed in the clinical laboratories of 27 public hospitals in 2012. The participants were recruited by the heads of service at each of the participating centers. Stratified analyses were performed to assess the mean score, standardized to a base of 100, of the six survey factors, together with the overall patient safety score. 740 completed questionnaires were received (88% of the 840 issued). The highest standardized scores were obtained in Area 1 (individual, social and cultural) with a mean value of 77 (95%CI: 76-78), and the lowest ones, in Area 3 (equipment and resources), with a mean value of 58 (95%CI: 57-59). In all areas, a greater perception of patient safety was reported by the heads of service than by other staff. We present the first multicentre study to evaluate the culture of clinical safety in public hospital laboratories in Spain. The results obtained evidence a culture in which high regard is paid to safety, probably due to the pattern of continuous quality improvement. Nevertheless, much remains to be done, as reflected by the weaknesses detected, which identify areas and strategies for improvement.

  8. 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.

  9. Guidelines for Self-assessment of Research Reactor Safety

    International Nuclear Information System (INIS)

    2018-01-01

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

  10. Assessment of private security guards by Suicide Probability Scale and Brief Symptom Inventory.

    Science.gov (United States)

    Dogan, Bulent; Canturk, Gurol; Canturk, Nergis; Guney, Sevgi; Özcan, Ebru

    2016-01-01

    The aim of the present study was to investigate the influence of suicide probability and relevant sociodemographic features and to provide information for preventing suicide in private security guards working under the stressful conditions and continuous exposure to the negative and traumatic life events. 200 private security guards and 200 personnels of Ankara University participated in the study. A sociodemographic information questionnaire, the Suicide Probability Scale (SPS) and the Brief Symptom Inventory (BSI) were used to collect the data. Gender, marital status, income, religious beliefs, experiencing a life-threatening situation, history of a suicide attempt, smoking and not having a chronic disease caused statistically significant differences in the scores for SPS between the private security guards group and the controls. Moreover there was a statistically significant positive correlation between the total scores of the subscales of SPS and the total scores of BSI. Like police officers and gendarmes, private security guards are at high risk of committing and attempting suicide because of being at stressful work settings and also suffering from secondary trauma. It is required that they should be aware of their tendency to commit suicide and have regular psychiatric screenings.

  11. Assessing changes in failure probability of dams in a changing climate

    Science.gov (United States)

    Mallakpour, I.; AghaKouchak, A.; Moftakhari, H.; Ragno, E.

    2017-12-01

    Dams are crucial infrastructures and provide resilience against hydrometeorological extremes (e.g., droughts and floods). In 2017, California experienced series of flooding events terminating a 5-year drought, and leading to incidents such as structural failure of Oroville Dam's spillway. Because of large socioeconomic repercussions of such incidents, it is of paramount importance to evaluate dam failure risks associated with projected shifts in the streamflow regime. This becomes even more important as the current procedures for design of hydraulic structures (e.g., dams, bridges, spillways) are based on the so-called stationary assumption. Yet, changes in climate are anticipated to result in changes in statistics of river flow (e.g., more extreme floods) and possibly increasing the failure probability of already aging dams. Here, we examine changes in discharge under two representative concentration pathways (RCPs): RCP4.5 and RCP8.5. In this study, we used routed daily streamflow data from ten global climate models (GCMs) in order to investigate possible climate-induced changes in streamflow in northern California. Our results show that while the average flow does not show a significant change, extreme floods are projected to increase in the future. Using the extreme value theory, we estimate changes in the return periods of 50-year and 100-year floods in the current and future climates. Finally, we use the historical and future return periods to quantify changes in failure probability of dams in a warming climate.

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

    Directory of Open Access Journals (Sweden)

    S. Boubaker

    2017-12-01

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

  13. National and international standards and recommendations on fire protection and fire safety assessment

    International Nuclear Information System (INIS)

    Berg, H.P.

    2007-01-01

    Experience feedback from events in nuclear facilities worldwide has shown that fire can represent a safety significant hazard. Thus, the primary objectives of fire protection programmes are to minimize both the probability of occurrence and the consequences of a fire. The regulator body expects that the licensees justify their arrangements for identifying how fires can occur and spread, assess the vulnerability of plant equipment and structures, determine how the safe operation of a plant is affected, and introduce measures to prevent a fire hazard from developing and propagating as well as to mitigate its effects in case the fire cannot be prevented. For that purpose usually a comprehensive regulatory framework for fire protection has been elaborated, based on national industrial regulations, nuclear specific regulations as well as international recommendations or requirements. Examples of such national and international standards and recommendations on fire protection and fire safety assessment as well as ongoing activities in this field are described. (orig.)

  14. Probabilistic safety assessment (Cernavoda). Experience and strategies

    International Nuclear Information System (INIS)

    Mircea, Mariana

    2000-01-01

    An IAEA project named 'Support for PSA related activities for Cernavoda NPP' was agreed at the beginning of 2000. The objectives were: upgrading of capability and framework to perform deterministic analyses as support for PSA (accident analyses and severe accident analyses); upgrading of capability and framework to extend the scope of PSA model for Cernavoda NPP to include internal and external hazards (internal fire, internal flooding, earthquake); upgrading of capability and framework to perform the Level 2 PSA for Cernavoda NPP. valuation was done for the status of the development of the seismic PSA, fire PSA and flooding PSA. For seismic PSA it was concluded by IAEA experts that this work needs adequate human and financial resources. Decision was taken to coordinate this project from Cernavoda but using specialists from external institutions. A Fire Hazard Assessment-FHA is in progress for Unit 1. First stage, regarding the methodology, was reviewed by IAEA experts in November 1999. In present, work is done for Reactor and Service Buildings. Work on flooding PSA was not started yet. To extend the PSA scope: Capability will be extended to develop the seismic PSA, fire PSA, flooding PSA (procurement of supplementary computer codes and specialist training); the extension of PSA scope to include internal and external hazards will continue after the completion of deterministic studies and is expected that the effective inclusion in the PSA model will start at the end of 2002

  15. Probabilistic safety assessment in the chemical and nuclear industries

    CERN Document Server

    Fullwood, Ralph R

    2000-01-01

    Probabilistic Safety Analysis (PSA) determines the probability and consequences of accidents, hence, the risk. This subject concerns policy makers, regulators, designers, educators and engineers working to achieve maximum safety with operational efficiency. Risk is analyzed using methods for achieving reliability in the space program. The first major application was to the nuclear power industry, followed by applications to the chemical industry. It has also been applied to space, aviation, defense, ground, and water transportation. This book is unique in its treatment of chemical and nuclear risk. Problems are included at the end of many chapters, and answers are in the back of the book. Computer files are provided (via the internet), containing reliability data, a calculator that determines failure rate and uncertainty based on field experience, pipe break calculator, event tree calculator, FTAP and associated programs for fault tree analysis, and a units conversion code. It contains 540 references and many...

  16. Earthquake induced liquefaction hazard, probability and risk assessment in the city of Kolkata, India: its historical perspective and deterministic scenario

    Science.gov (United States)

    Nath, Sankar Kumar; Srivastava, Nishtha; Ghatak, Chitralekha; Adhikari, Manik Das; Ghosh, Ambarish; Sinha Ray, S. P.

    2018-01-01

    Liquefaction-induced ground failure is one amongst the leading causes of infrastructure damage due to the impact of large earthquakes in unconsolidated, non-cohesive, water saturated alluvial terrains. The city of Kolkata is located on the potentially liquefiable alluvial fan deposits of Ganga-Bramhaputra-Meghna Delta system with subsurface litho-stratigraphic sequence comprising of varying percentages of clay, cohesionless silt, sand, and gravel interbedded with decomposed wood and peat. Additionally, the region has moderately shallow groundwater condition especially in the post-monsoon seasons. In view of burgeoning population, there had been unplanned expansion of settlements in the hazardous geological, geomorphological, and hydrological conditions exposing the city to severe liquefaction hazard. The 1897 Shillong and 1934 Bihar-Nepal earthquakes both of M w 8.1 reportedly induced Modified Mercalli Intensity of IV-V and VI-VII respectively in the city reportedly triggering widespread to sporadic liquefaction condition with surface manifestation of sand boils, lateral spreading, ground subsidence, etc., thus posing a strong case for liquefaction potential analysis in the terrain. With the motivation of assessing seismic hazard, vulnerability, and risk of the city of Kolkata through a consorted federal funding stipulated for all the metros and upstart urban centers in India located in BIS seismic zones III, IV, and V with population more than one million, an attempt has been made here to understand the liquefaction susceptibility condition of Kolkata under the impact of earthquake loading employing modern multivariate techniques and also to predict deterministic liquefaction scenario of the city in the event of a probabilistic seismic hazard condition with 10% probability of exceedance in 50 years and a return period of 475 years. We conducted in-depth geophysical and geotechnical investigations in the city encompassing 435 km2 area. The stochastically

  17. A Framework for Assessment of Aviation Safety Technology Portfolios

    Science.gov (United States)

    Jones, Sharon M.; Reveley, Mary S.

    2014-01-01

    The programs within NASA's Aeronautics Research Mission Directorate (ARMD) conduct research and development to improve the national air transportation system so that Americans can travel as safely as possible. NASA aviation safety systems analysis personnel support various levels of ARMD management in their fulfillment of system analysis and technology prioritization as defined in the agency's program and project requirements. This paper provides a framework for the assessment of aviation safety research and technology portfolios that includes metrics such as projected impact on current and future safety, technical development risk and implementation risk. The paper also contains methods for presenting portfolio analysis and aviation safety Bayesian Belief Network (BBN) output results to management using bubble charts and quantitative decision analysis techniques.

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

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

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

  19. Real-time safety risk assessment based on a real-time location system for hydropower construction sites.

    Science.gov (United States)

    Jiang, Hanchen; Lin, Peng; Fan, Qixiang; Qiang, Maoshan

    2014-01-01

    The concern for workers' safety in construction industry is reflected in many studies focusing on static safety risk identification and assessment. However, studies on real-time safety risk assessment aimed at reducing uncertainty and supporting quick response are rare. A method for real-time safety risk assessment (RTSRA) to implement a dynamic evaluation of worker safety states on construction site has been proposed in this paper. The method provides construction managers who are in charge of safety with more abundant information to reduce the uncertainty of the site. A quantitative calculation formula, integrating the influence of static and dynamic hazards and that of safety supervisors, is established to link the safety risk of workers with the locations of on-site assets. By employing the hidden Markov model (HMM), the RTSRA provides a mechanism for processing location data provided by the real-time location system (RTLS) and analyzing the probability distributions of different states in terms of false positives and negatives. Simulation analysis demonstrated the logic of the proposed method and how it works. Application case shows that the proposed RTSRA is both feasible and effective in managing construction project safety concerns.

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

    Directory of Open Access Journals (Sweden)

    Baranowski Tom

    2010-05-01

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

  1. Mathematical Safety Assessment Approaches for Thermal Power Plants

    Directory of Open Access Journals (Sweden)

    Zong-Xiao Yang

    2014-01-01

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

  2. Safety assessment of geologic repositories for nuclear waste

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. Safety Assessment of Talc as Used in Cosmetics.

    Science.gov (United States)

    Fiume, Monice M; Boyer, Ivan; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan

    2015-01-01

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

  4. Safety assessment for the underground disposal of radioactive wastes

    International Nuclear Information System (INIS)

    1981-01-01

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

  5. Cyber Security Risk Assessment for the KNICS Safety Systems

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  6. Safety assessment for the underground disposal of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

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

  7. Hazard Identification and Risk Assessment in Water Treatment Plant considering Environmental Health and Safety Practice

    Science.gov (United States)

    Falakh, Fajrul; Setiani, Onny

    2018-02-01

    Water Treatment Plant (WTP) is an important infrastructure to ensure human health and the environment. In its development, aspects of environmental safety and health are of concern. This paper case study was conducted at the Water Treatment Plant Company in Semarang, Central Java, Indonesia. Hazard identification and risk assessment is one part of the occupational safety and health program at the risk management stage. The purpose of this study was to identify potential hazards using hazard identification methods and risk assessment methods. Risk assessment is done using criteria of severity and probability of accident. The results obtained from this risk assessment are 22 potential hazards present in the water purification process. Extreme categories that exist in the risk assessment are leakage of chlorine and industrial fires. Chlorine and fire leakage gets the highest value because its impact threatens many things, such as industrial disasters that could endanger human life and the environment. Control measures undertaken to avoid potential hazards are to apply the use of personal protective equipment, but management will also be better managed in accordance with hazard control hazards, occupational safety and health programs such as issuing work permits, emergency response training is required, Very useful in overcoming potential hazards that have been determined.

  8. Hazard Identification and Risk Assessment in Water Treatment Plant considering Environmental Health and Safety Practice

    Directory of Open Access Journals (Sweden)

    Falakh Fajrul

    2018-01-01

    Full Text Available Water Treatment Plant (WTP is an important infrastructure to ensure human health and the environment. In its development, aspects of environmental safety and health are of concern. This paper case study was conducted at the Water Treatment Plant Company in Semarang, Central Java, Indonesia. Hazard identification and risk assessment is one part of the occupational safety and health program at the risk management stage. The purpose of this study was to identify potential hazards using hazard identification methods and risk assessment methods. Risk assessment is done using criteria of severity and probability of accident. The results obtained from this risk assessment are 22 potential hazards present in the water purification process. Extreme categories that exist in the risk assessment are leakage of chlorine and industrial fires. Chlorine and fire leakage gets the highest value because its impact threatens many things, such as industrial disasters that could endanger human life and the environment. Control measures undertaken to avoid potential hazards are to apply the use of personal protective equipment, but management will also be better managed in accordance with hazard control hazards, occupational safety and health programs such as issuing work permits, emergency response training is required, Very useful in overcoming potential hazards that have been determined.

  9. Value-impact assessment of safety-related modifications

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  10. Current issues and perspectives in food safety and risk assessment.

    Science.gov (United States)

    Eisenbrand, G

    2015-12-01

    In this review, current issues and opportunities in food safety assessment are discussed. Food safety is considered an essential element inherent in global food security. Hazard characterization is pivotal within the continuum of risk assessment, but it may be conceived only within a very limited frame as a true alternative to risk assessment. Elucidation of the mode of action underlying a given hazard is vital to create a plausible basis for human toxicology evaluation. Risk assessment, to convey meaningful risk communication, must be based on appropriate and reliable consideration of both exposure and mode of action. New perspectives, provided by monitoring human exogenous and endogenous exposure biomarkers, are considered of great promise to support classical risk extrapolation from animal toxicology. © The Author(s) 2015.

  11. 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)

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

    International Nuclear Information System (INIS)

    Edwards, Gareth

    2005-01-01

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

  13. Microbial safety assessment of recreation water at Lake Nabugabo ...

    African Journals Online (AJOL)

    EJIRO

    Key words: Lake Nabugabo, microbial safety assessment, recreation water, water quality. ... the environment is favourable for growth (Jaiani et al., ... Swimming and bathing in inland waters are recognized .... in India. This can be attributed to variation in number of recreational users and the frequency of use of the various.

  14. Nuclear data for radiation damage assessment and related safety aspects

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1989-12-01

    The IAEA Advisory Group Meeting on Nuclear Data for Radiation Damage Assessment and Related Safety Aspects was held at the IAEA Headquarters in Vienna, 19-22 September 1989. This report contains the conclusions and recommendations of this meeting. The papers which the participants prepared for and presented at the meeting will be published as an IAEA Technical Document. (author)

  15. Safety assessment standards for modern plants in the UK

    International Nuclear Information System (INIS)

    Harbison, S.A.; Hannaford, J.

    1993-01-01

    The NII has revised its safety assessment principles (SAPs). This paper discusses the revised SAPs and their links with international standards. It considers the licensing of foreign designs of plant - a matter under active consideration in the UK -and discusses how the SAPs and the licensing process cater for that possibility. (author)

  16. Assessing the Food Safety Knowledge of University of Maine Students

    Science.gov (United States)

    Ferk, Chelsea C.; Calder, Beth L.; Camire, Mary Ellen

    2016-01-01

    Foodborne illness is a global public health issue. Young adults may work in foodservice while they are university students, and their habits may later shape the practices and well-being of their children. The objective of this study was to establish baseline data and assess the food safety knowledge of 18- to 26-year-old Univ. of Maine students.…

  17. Waste isolation safety assessment program. Summary of FY-77 progress

    International Nuclear Information System (INIS)

    Burkholder, H.C.; Greenborg, J.; Stottlemyre, J.A.; Bradley, D.J.; Raymond, J.R.; Serne, R.J.

    1977-11-01

    Objective is to provide long-term safety information for the National Waste Terminal Storage Program. Work in FY 77 supported the development of the generic assessment method (release scenario analysis, release consequence analysis) and of the generic data base (waste form release rate data, radionuclide geochemical interaction data)

  18. A review of international sources for road safety measures assessment.

    NARCIS (Netherlands)

    Yannis, G. Weijermars, W. & Kauppila, J.

    2012-01-01

    The efficiency assessment of road safety measures is considered to be an extremely useful tool in decision making; in particular, cost-benefit and cost-effectiveness analyses are carried out in several countries, in a more or less systematic way. The objective of this paper is to present findings

  19. Safety assessment of indoor live fire range, May 1989

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-05-01

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

  20. Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

    International Nuclear Information System (INIS)

    Ergun, E; Kadiroglu, O.S.

    1999-01-01

    The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

  1. Nuclear Criticality Safety Assessment for Tank 38H Salt Dissolution

    International Nuclear Information System (INIS)

    Davis, P.L.

    1996-01-01

    This assessment report of sample results of the accumulating insoluble solids from Tank 38H demonstrates that an inherent subcritical condition for nuclear criticality safety exists during saltcake dissolution. This report also defines criteria for future sampling of Tank 38H for continued verification of the inherent subcritical condition as saltcake dissolution proceeds

  2. Comparative safety assessment of plant-derived foods

    NARCIS (Netherlands)

    Kok, E.J.; Keijer, J.; Kleter, G.A.; Kuiper, H.A.

    2008-01-01

    The second generation of genetically modified (GM) plants that are moving towards the market are characterized by modifications that may be more complex and traits that more often are to the benefit of the consumer. These developments will have implications for the safety assessment of the resulting

  3. Method for quantitative assessment of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Dearien, J.A.; Davis, C.B.; Matthews, L.J.

    1979-01-01

    A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

  4. Safety assessment of the Area 6 Decontamination Pad and Laundry

    International Nuclear Information System (INIS)

    Chilton, M.W.; Orcutt, J.A.

    1984-10-01

    The Safety Assessment of the Area 6 Decontamination Pad and Laundry, prepared in accordance with DOE Order 5481.1A, identifies and evaluates potential radiation and chemical hazards to personnel, and impacts on the environment. Site and facility characteristics, as well as routine and nonroutine operations are discussed. Hypothetical incidents and accidents are described and evaluated. 3 figures, 1 table

  5. Learning Safety Assessment from Accidents in a University Environment

    OpenAIRE

    Jensen, Niels; Jørgensen, Sten Bay

    2013-01-01

    This contribution describes how a chemical engineering department started learning from accidents during experimental work and ended up implementing an industrially inspired system for risk assessment of new and existing experimental setups as well as a system for assessing potential risk from the chemicals used in the experimental work. These experiences have led to recent developments which focus increasingly on the a theoretical basis for modeling and reasoning on safety as well as operati...

  6. Assessing the Effectiveness of the Cone of Probability as a Visual Means of Communicating Scientific Forecasts

    Science.gov (United States)

    Orlove, B. S.; Broad, K.; Meyer, R.

    2010-12-01

    We review the evolution, communication, and differing interpretations of the National Hurricane Center (NHC)'s "cone of uncertainty" hurricane forecast graphic, drawing on several related disciplines—cognitive psychology, visual anthropology, and risk communication theory. We examine the 2004 hurricane season, two specific hurricanes (Katrina 2005 and Ike 2008) and the 2010 hurricane season, still in progress. During the 2004 hurricane season, five named storms struck Florida. Our analysis of that season draws upon interviews with key government officials and media figures, archival research of Florida newspapers, analysis of public comments on the NHC cone of uncertainty graphic and a multiagency study of 2004 hurricane behavior. At that time, the hurricane forecast graphic was subject to misinterpretation by many members of the public. We identify several characteristics of this graphic that contributed to public misinterpretation. Residents overemphasized the specific track of the eye, failed to grasp the width of hurricanes, and generally did not recognize the timing of the passage of the hurricane. Little training was provided to emergency response managers in the interpretation of forecasts. In the following year, Katrina became a national scandal, further demonstrating the limitations of the cone as a means of leading to appropriate responses to forecasts. In the second half of the first decade of the 21st century, three major changes occurred in hurricane forecast communication: the forecasts themselves improved in terms of accuracy and lead time, the NHC made minor changes in the graphics and expanded the explanatory material that accompanies the graphics, and some efforts were made to reach out to emergency response planners and municipal officials to enhance their understanding of the forecasts and graphics. There were some improvements in the responses to Ike, though a number of deaths were due to inadequate evacuations, and property damage probably

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

    International Nuclear Information System (INIS)

    Andersson, Johan

    2012-01-01

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

  8. Upgrading the safety assessment of exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

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

  9. Assessment of safety culture in the Iranian nuclear installations

    International Nuclear Information System (INIS)

    Farahani, H.F.; Davilu, H.; Sepanloo, K.

    2005-01-01

    The deficient safety culture (S.C) is the center of safety issues of nuclear industry. To benefit from the advantages of nuclear technology and considering the fact of potential hazards of accidents in nuclear installations it is essential to view safety as the highest priority. S.C is an amalgamation of values, standards, morals and norms of acceptable behavior. Organizations having effective S.C show constant commitment to safety as a top level priority. Furthermore, the personnel of a nuclear facility shall recognize the safety significance of their tasks. Many people even those who work in the field of safety do not have a correct understanding of what S.C looks like in practical sense. In this study, by conducting a survey according to IAEA-TECDOC-1329 in some nuclear facilities, the S.C within the Iranian nuclear facilities is assessed. The human and organizational factors in Tehran Research Reactor are evaluated using a questionnaire method with active participation of the reactor operators. The results sho w that the operators are pretty aware of the subject. Also it has been identified some areas of improvement. (authors)

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

    International Nuclear Information System (INIS)

    2012-01-01

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

  11. Concrete structures. Contribution to the safety assessment of existing structures

    Directory of Open Access Journals (Sweden)

    D. COUTO

    Full Text Available The safety evaluation of an existing concrete structure differs from the design of new structures. The partial safety factors for actions and resistances adopted in the design phase consider uncertainties and inaccuracies related to the building processes of structures, variability of materials strength and numerical approximations of the calculation and design processes. However, when analyzing a finished structure, a large number of unknown factors during the design stage are already defined and can be measured, which justifies a change in the increasing factors of the actions or reduction factors of resistances. Therefore, it is understood that safety assessment in existing structures is more complex than introducing security when designing a new structure, because it requires inspection, testing, analysis and careful diagnose. Strong knowledge and security concepts in structural engineering are needed, as well as knowledge about the materials of construction employed, in order to identify, control and properly consider the variability of actions and resistances in the structure. With the intention of discussing this topic considered complex and diffuse, this paper presents an introduction to the safety of concrete structures, a synthesis of the recommended procedures by Brazilian standards and another codes, associated with the topic, as well a realistic example of the safety assessment of an existing structure.

  12. 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

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

    Science.gov (United States)

    Działak, Paulina; Karliński, Jacek; Rusiński, Eugeniusz

    2018-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Działak Paulina

    2018-01-01

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

  15. Assessing nuclear power plant safety and recovery from earthquakes using a system-of-systems approach

    International Nuclear Information System (INIS)

    Ferrario, E.; Zio, E.

    2014-01-01

    We adopt a ‘system-of-systems’ framework of analysis, previously presented by the authors, to include the interdependent infrastructures which support a critical plant in the study of its safety with respect to the occurrence of an earthquake. We extend the framework to consider the recovery of the system of systems in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant (the critical plant) embedded in the connected power and water distribution, and transportation networks which support its operation. The Seismic Probabilistic Risk Assessment of such system of systems is carried out by Hierarchical modeling and Monte Carlo simulation. First, we perform a top-down analysis through a hierarchical model to identify the elements that at each level have most influence in restoring safety, adopting the criticality importance measure as a quantitative indicator. Then, we evaluate by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe state and the time needed to recover its safety. The results obtained allow the identification of those elements most critical for the safety and recovery of the nuclear power plant; this is relevant for determining improvements of their structural/functional responses and supporting the decision-making process on safety critical-issues. On the test system considered, under the given assumptions, the components of the external and internal water systems (i.e., pumps and pool) turn out to be the most critical for the safety and recovery of the plant. - Highlights: • We adopt a system-of-system framework to analyze the safety of a critical plant exposed to risk from external events, considering also the interdependent infrastructures that support the plant. • We develop a hierarchical modeling framework to represent the system of systems, accounting also for its recovery. • Monte Carlo simulation is used for the quantitative evaluation of the

  16. Environment, Safety and Health Progress Assessment of the Hanford Site

    International Nuclear Information System (INIS)

    1992-05-01

    This report documents the result of the US Department of Energy (DOE) Environment, Safety and Health (ES ampersand H) Progress Assessment of the Hanford Site, in Richland, Washington. The assessment, which was conducted from May 11 through May 22, 1992, included a selective-review of the ES ampersand H management systems and programs of the responsible DOE Headquarters Program Offices the DOE Richland Field Office, and the site contractors. The ES ampersand H Progress Assessments are part of the Secretary of Energy's continuing effort to institutionalize line management accountability and the self-assessment process throughout DOE and its contractor organizations. The purpose of the Hanford Site ES ampersand H Progress Assessment is to provide the Secretary with an independent assessment of the adequacy and effectiveness of the DOE and contractor management structures, resources, and systems to address ES ampersand H problems and requirements. They are not intended to be comprehensive compliance assessments of ES ampersand H activities. The point of reference for assessing programs at the Hanford Site was, for the most part, the Tiger Team Assessment of the Hanford Site, which was conducted from May 21 through July 18, 1990. A summary of issues and progress in the areas of environment, safety and health, and management is included

  17. The use of safety indicators in the assessment of radioactive waste disposal

    International Nuclear Information System (INIS)

    Wingefors, S.; Westerlind, M.; Gera, F.

    1999-01-01

    The most widely used criteria for disposal are limits or constraints on individual dose or risk, and these have been introduced in most national legal frameworks. There is general agreement that future generations have the right to the same level of protection as the current generation. Even if quantitative criteria corresponding to the required level of protection can be (and have been) defined, it is a great challenge to demonstrate compliance with these criteria. The difficulties are to large extent due to the long time-scales needed to be considered in radioactive waste disposal. The future cannot be predicted in detail but instead different scenarios, with different probabilities of occurrence, must be assessed. Some parts of a disposal system can be predicted or analysed with high confidence for very long periods of time, e.g. geological formations, while for example the evolution of the biosphere, and in particular the society, become quite uncertain within less than one thousand years. Thus, there may be considerable uncertainty in doses (or risks) derived from the safety assessment of a repository. Due to these unavoidable uncertainties it is believed advantageous to use multiple approaches in the safety assessment and to identify different indicators for the repository safety ('multiple-lines-of-reasoning'). The most fundamental safety indicators are dose/risk but complementary indicators have been suggested, in particular flux and environmental concentration of radionuclides. This presentation is focussed on fluxes and concentrations as complementary safety indicators. Other safety indicators, e.g. transfer times, are mentioned only briefly

  18. Flamanville 3 EPR, Safety Assessment and On-site Inspections

    International Nuclear Information System (INIS)

    Piedagnel, Corinne; Tarallo, Francois; Monnot, Bernard

    2011-01-01

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

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

    Science.gov (United States)

    Mendoza, Jason A; Watson, Kathy; Baranowski, Tom; Nicklas, Theresa A; Uscanga, Doris K; Hanfling, Marcus J

    2010-05-18

    Safe Routes to School (SRTS) programs are designed to make walking and bicycling to school safe and accessible for children. Despite their growing popularity, few validated measures exist for assessing important outcomes such as type of student transport or pedestrian safety behaviors. This research validated the SRTS school travel survey and a pedestrian safety behavior checklist. Fourth grade students completed a brief written survey on how they got to school that day with set responses. Test-retest reliability was obtained 3-4 hours apart. Convergent validity of the SRTS travel survey was assessed by comparison to parents' report. For the measure of pedestrian safety behavior, 10 research assistants observed 29 students at a school intersection for completion of 8 selected pedestrian safety behaviors. Reliability was determined in two ways: correlations between the research assistants' ratings to that of the Principal Investigator (PI) and intraclass correlations (ICC) across research assistant ratings. The SRTS travel survey had high test-retest reliability (kappa = 0.97, n = 96, p < 0.001) and convergent validity (kappa = 0.87, n = 81, p < 0.001). The pedestrian safety behavior checklist had moderate reliability across research assistants' ratings (ICC = 0.48) and moderate correlation with the PI (r = 0.55, p = < 0.01). When two raters simultaneously used the instrument, the ICC increased to 0.65. Overall percent agreement (91%), sensitivity (85%) and specificity (83%) were acceptable. These validated instruments can be used to assess SRTS programs. The pedestrian safety behavior checklist may benefit from further formative work.

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

    Science.gov (United States)

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

    2008-12-15

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

  1. 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)

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

    Science.gov (United States)

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

    2014-02-01

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

  3. Assessment of the probability of introducing Mycobacterium tuberculosis into Danish cattle herds

    DEFF Research Database (Denmark)

    Foddai, Alessandro; Nielsen, Liza Rosenbaum; Krogh, Kaspar

    2015-01-01

    working in Danish cattle herds was obtained through a questionnaire sent to Danish cattle farmers. The gained inputs were fed into three stochastic scenario trees to assess the PIntro for the current and alternative test-and-manage strategies, such as testing of imported animals and/or testing immigrant...

  4. Statistical analysis applied to safety culture self-assessment

    International Nuclear Information System (INIS)

    Macedo Soares, P.P.

    2002-01-01

    Interviews and opinion surveys are instruments used to assess the safety culture in an organization as part of the Safety Culture Enhancement Programme. Specific statistical tools are used to analyse the survey results. This paper presents an example of an opinion survey with the corresponding application of the statistical analysis and the conclusions obtained. Survey validation, Frequency statistics, Kolmogorov-Smirnov non-parametric test, Student (T-test) and ANOVA means comparison tests and LSD post-hoc multiple comparison test, are discussed. (author)

  5. Results of the CANDU 3 probabilistic safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaitly, R K [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10{sup -6}/year. 4 refs., 1 fig.

  6. Safety assessment of a robotic system handling nuclear material

    International Nuclear Information System (INIS)

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

    1996-01-01

    This paper outlines the use of a Failure Modes and Effects Analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, The Weigh and Leak Check System, is to replace a manual process at the Department of Energy facility at Pantex by which nuclear material is inspected for weight and leakage. Failure Modes and Effects Analyses were completed for the robotics process to ensure that safety goals for the system had been meet. These analyses showed that the risks to people and the internal and external environment were acceptable

  7. ProZES. A tool for assessment of assigned share of radiation in probability of cancer development (Pt. 2)

    Energy Technology Data Exchange (ETDEWEB)

    Ulanowski, Alexander; Eidemueller, Markus; Guethlin, Denise; Kaiser, Jan Christian; Shemiakina, Elena; Jacob, Peter [Helmholtz Zentrum Muenchen - Deutsches Forschungszentrum fuer Gesundheit und Umwelt, Muenchen (Germany). Inst fuer Strahlenschutz

    2016-11-15

    Methodology and a corresponding computer program ProZES were developed to estimate the probability that a previous radiation exposure for a specific person and a given exposure situation has resulted in cancer (probability of causation or relationship between the exposure and the disease, Z). ProZES can provide the scientific basis to support making decisions on compensation claims due to cancer following occupational exposure to radiation. Starting from the results achieved in the first version of ProZES, when the general methodology and risk models for colon, stomach, lung, and female breast were implemented, the second stage of the ProZES development was focused on the development of risk models for all other cancer locations, including leukaemias and lymphomas as well as risk models for lung cancer after exposure to radon. The models for estimating the cancer risks and the associated probability Z are mostly based on the observed cancer incidence in the cohort of the atomic bomb survivors in Hiroshima and Nagasaki. Most of the models are newly developed for the project. For the frequent types of cancer, specific models of radiation risk have been developed, while for the less common diseases the risk models were developed for the groups of functionally similar diseases. Since various models built upon the basis of the same data can result in different predictions for ''dose-effect'' relationships, so the method of ''multi-model inference'' is used for some types of cancer to derive risk factors, which are less dependent on individual models and take model uncertainties into account. Risk estimates for the Japanese population must be transferred to the German population. An essential element is the estimation of the uncertainty of the associated probability. ProZES was developed as a user-friendly stand-alone program, which can assess and present the individualised estimate of probability of relationship between radiation

  8. Assessment of climate change using methods of mathematic statistics and theory of probability

    International Nuclear Information System (INIS)

    Trajanoska, Lidija; Kaevski, Ivancho

    2004-01-01

    In simple terms: 'Climate' is the average of 'weather'. The Earth's weather system is a complex machine composed of coupled sub-systems (ocean, air, land, ice and the biosphere) between which energy are exchanged. The understanding and study of climate change does not only rely on the understanding of the physics of climate change but is linked to the following question: 'How we can detect change in a system that is changing all the time under its own volition'? What is even the meaning of 'change' in such a situation? The concept of 'change' we should transform into the concept of 'significant and long-term' then this re-phrasing allows for a definition in mathematical terms. Significant change in a system becomes a measure of how large an observed change is in terms of the variability one would see under 'normal' conditions. Example could be the analyses of the yearly temperature of the air and precipitations, like in this paper. A large amount of data are selected as representing the 'before' case (change) and another set of data are selected as being the 'after' case and then the average in these two cases are compared. These comparisons are in the form of 'hypothesis tests' in which one tests whether the hypothesis that there has Open no change can be rejected. Both parameter and nonparametric statistic methods are used in the theory of mathematic statistic. The most indicative changeable which show global change is an average, standard deviation and probability function distribution on examined time series. Examined meteorological series are taken like haphazard process so we can mathematic statistic applied.(Author)

  9. Aspects of safety assessments for package with additional equipment components

    International Nuclear Information System (INIS)

    Reiche, I.; Boerst, F.M.; Krietsch, T.

    2004-01-01

    Many paragraphs in TS-R-1 contain the terms ''package'' or ''packaging''. These terms are defined in TS-R-1 paras 230 and 231 and explained in TS-G-1.1 paras 230.1 - 230.6. The importance of a consistent understanding of these definitions has been shown by recent discussions during the assessment of applications for package design approval. There was disagreement, if equipment components attached to the container body during transport, e.g. a transport frame, should be considered part of the package and taken into account in the safety assessment for the package. Discussions were also caused by the way inner design components are treated in the safety assessment of the package. This paper summarises the regulatory requirements to such additional equipment components and presents the way of their inclusion into the package design approval process in Germany

  10. Containment response analysis for the PSA (Probabilistic Safety Assessment) of the CAREM-25 nuclear power plant

    International Nuclear Information System (INIS)

    Baron, J.H.

    1997-01-01

    This work is part of the probabilistic safety assessment actually under development for the CAREM-25 nuclear power station, and departs from the accident sequences already obtained and quantified by the Event Trees/Fault Trees techniques. At first, the potential containment failure modes for nuclear stations are listed, based on the experience. Then, the CAREM-25 design peculiarities are analyzed, on their possible influence on the containment behavior during severe accidents. Then plan damage states are defined. Furthermore, containment damage states are also defined, and containment event trees are built for each plant damage state. Those sequences considered representative from the annual probability (those which exceed or probability of IE-09 per year, are used to quantify the combinations of plant damage states/containment damage states, based on the estimation of a vulnerability matrix. (author) [es

  11. Human reliability analysis for probabilistic safety assessments - review of methods and issues

    International Nuclear Information System (INIS)

    Srinivas, G.; Guptan, Rajee; Malhotra, P.K.; Ghadge, S.G.; Chandra, Umesh

    2011-01-01

    It is well known that the two major events in World Nuclear Power Plant Operating history, namely the Three Mile Island and Chernobyl, were Human failure events. Subsequent to these two events, several significant changes have been incorporated in Plant Design, Control Room Design and Operator Training to reduce the possibility of Human errors during plant transients. Still, human error contribution to Risk in Nuclear Power Plant operations has been a topic of continued attention for research, development and analysis. Probabilistic Safety Assessments attempt to capture all potential human errors with a scientifically computed failure probability, through Human Reliability Analysis. Several methods are followed by different countries to quantify the Human error probability. This paper reviews the various popular methods being followed, critically examines them with reference to their criticisms and brings out issues for future research. (author)

  12. Results of the probabilistic safety assessment to the cobalt-therapy process

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Ferro, R.; Lozano, B.; De la Fuente Puch, Andres; Dumenigo Gonzalez, Cruz; Troncoso, M.; Perez, Y.; Alemany, J.; Leon, L.; Amador, R.; Lazo, R.; Labrador, F.; Blanco, A.; Betancourt, L.; Crespo, D.; Silvestre, I.

    2004-01-01

    This paper presents the results of the Probabilistic Safety Assessment (PSA) to the Cobalt Therapy Treatment Process in the Oncological Unit of Pinar del Rio city to evaluate occupational, public and medical exposures during cobalt therapy treatment. Equipment's Failures Modes and Human Error were evaluated for each system and treatment stage aimed at obtaining an exhaustive list of the deviations with a reasonable probability to occur and may produce significant adverse outcomes. The lowest exposures probabilities correspond to the public exposures during the treatment process; around 10-10 per year, being the workers exposures around 10-4 per year. Regarding the patient, exposures frequencies vary in dependence of the extent to which the error affect individual treatment, individual patients, or all the patients treated on a specific unit

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

    International Nuclear Information System (INIS)

    2011-12-01

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

  14. Concept of risk: risk assessment and nuclear safety

    International Nuclear Information System (INIS)

    Thompson, P.B.

    1980-01-01

    The dissertation is a critical examination of risk assessment and its role in public policy. Nuclear power safety safety issues are selected as the primary source of illustrations and examples. The dissertation examines how risk assessment studies develop a concept of risk which becomes decisive for policy choices. Risk-assessment techniques are interpreted as instruments which secure an evaluation of risk which, in turn, figures prominently in technical reports on nuclear power. The philosophical critique is mounted on two levels. First, an epistemological critique surveys distinctions between the technical