WorldWideScience

Sample records for reliable core analysis

  1. Reliability and accuracy of a video analysis protocol to assess core ability.

    Science.gov (United States)

    McDonald, Dawn A; Delgadillo, James Q; Fredericson, Michael; McConnell, Jennifer; Hodgins, Melissa; Besier, Thor F

    2011-03-01

    To develop and test a method to measure core ability in healthy athletes with 2-dimensional video analysis software (SiliconCOACH). Specific objectives were to: (1) develop a standardized exercise battery with progressions of increasing difficulty to evaluate areas of core ability in elite athletes; (2) develop an objective and quantitative grading rubric with the use of video analysis software; (3) assess the test-retest reliability of the exercise battery; (4) assess the interrater and intrarater reliability of the video analysis system; and (5) assess the accuracy of the assessment. Test-retest repeatability and accuracy. Testing was conducted in the Stanford Human Performance Laboratory, Stanford University, Stanford, CA. Nine female gymnasts currently training with the Stanford Varsity Women's Gymnastics Team participated in testing. Participants completed a test battery composed of planks, side planks, and leg bridges of increasing difficulty. Subjects completed two 20-minute testing sessions within a 4- to 10-day period. Two-dimensional sagittal-plane video was captured simultaneously with 3-dimensional motion capture. The main outcome measures were pelvic displacement and time that elapsed until failure occurred, as measured with SiliconCOACH video analysis software. Test-retest and interrater and intrarater reliability of the video analysis measures was assessed. Accuracy as compared with 3-dimensional motion capture also was assessed. Levels reached during the side planks and leg bridges had an excellent test-retest correlation (r(2) = 0.84, r(2) = 0.95). Pelvis displacements measured by examiner 1 and examiner 2 had an excellent correlation (r(2) = 0.86, intraclass correlation coefficient = 0.92). Pelvis displacements measured by examiner 1 during independent grading sessions had an excellent correlation (r(2) = 0.92). Pelvis displacements from the plank and from a set of combined plank and side plank exercises both had an excellent correlation with 3

  2. Reliability Analysis of Core Protection Calculator System by Combining Petri Net and Fault Tree

    International Nuclear Information System (INIS)

    Kim, Hyejin; Kim, Jonghyun

    2013-01-01

    This paper proposes an approach to analyzing the reliability of digital systems by combining Petri net (PN) and Fault tree. The Petri net allows modeling event dependencies and interaction, to represent the time sequence, and to model assumptions for dynamic events. The Petri net model can be straightforwardly transformed to fault tree using the gate. Then, the FT can be integrated into the existing PSA. This paper applies the approach to the reliability analysis of Core Protection Calculator System (CPCS). Digital technology is replacing the analog instrumentation and control (I and C) systems in both new and upgraded nuclear power plants. As digital systems are introduced to nuclear power plants, issues related with reliability analyses of these digital systems are being raised. One of these issues is that static fault tree (FT) and event tree (ET) approach cannot properly account for dynamic interactions in the digital systems, such as multiple top events, logic loops and time delay. Many methods have been proposed to solve the problems, but there is no single method that is universally accepted for the application to the current generation probabilistic safety analysis (PSA)

  3. Reliability Analysis of Core Protection Calculator System by Combining Petri Net and Fault Tree

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyejin; Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    This paper proposes an approach to analyzing the reliability of digital systems by combining Petri net (PN) and Fault tree. The Petri net allows modeling event dependencies and interaction, to represent the time sequence, and to model assumptions for dynamic events. The Petri net model can be straightforwardly transformed to fault tree using the gate. Then, the FT can be integrated into the existing PSA. This paper applies the approach to the reliability analysis of Core Protection Calculator System (CPCS). Digital technology is replacing the analog instrumentation and control (I and C) systems in both new and upgraded nuclear power plants. As digital systems are introduced to nuclear power plants, issues related with reliability analyses of these digital systems are being raised. One of these issues is that static fault tree (FT) and event tree (ET) approach cannot properly account for dynamic interactions in the digital systems, such as multiple top events, logic loops and time delay. Many methods have been proposed to solve the problems, but there is no single method that is universally accepted for the application to the current generation probabilistic safety analysis (PSA)

  4. Reliable core thermal design

    International Nuclear Information System (INIS)

    Amendola, A.

    1974-01-01

    The hot spot analysis is no longer limited to the calculation of a simple safety factor against overtemperature, but is now integrated in the overall design philosophy. This paper describes the development of a probabilistic method of analysis and compares it with other advanced calculation methods. Feedbacks from the analysis act: - on the nominal temperature distribution in order to satisfy the maximum temperature limit and in the same time to optimize the coolant temperature for maximum plant efficiency, and - on the specifications of manufacturing tolerances and experimental investigations in order to identify and to reduce the most important design uncertainties. Moreover the computer codes SHOSPA and THEDRA are briefly discussed. Both codes deliver the zero hot spot probability as a function of the geometrical size assumed for a ''spot''. THEDRA delivers also the expected hot spot distribution. By means of THEDRA it is possible to evaluate the pins failure expectation if the distribution of pin failures versus operating temperature is known. (author)

  5. Unavailability Analysis of the Reactor Core Protection System using Reliability Block Diagram

    International Nuclear Information System (INIS)

    Shin, Hyun Kook; Kim, Sung Ho; Choi, Woong Suk; Kim, Jae Hack

    2006-01-01

    The reactor core of nuclear power plants needs to be monitored for the early detection of core abnormal conditions to protect plants from a severe accident. The core protection calculator system (CPCS) has been provided to calculate the departure from nucleate boiling ratio (DNBR) and the local power density (LPD) based on measured parameters of reactor and coolant system. The original CPCS for OPR 1000 has been designed and implemented based on the concurrent 3205 computer system whose components are obsolete. The CPCS based on Westinghouse Common-Q system has recently been implemented for the Shin-Kori Nuclear Power Plant, Units 1 and 2(SKN 1 and 2). An R and D project has been launched to develop new core protection system called as RCOPS (Reactor Core Protection System) with the partnership of KOPEC and Doosan Heavy Industries and Construction Co. RCOPS is implemented on the HFC-6000 safety class programmable logic controller (PLC). In this paper, the reliability of RCOPS is analyzed using the reliability block diagram (RBD) method. The calculated results are compared with that of the CPCS for SKN 1 and 2

  6. Citation analysis did not provide a reliable assessment of core outcome set uptake.

    Science.gov (United States)

    Barnes, Karen L; Kirkham, Jamie J; Clarke, Mike; Williamson, Paula R

    2017-06-01

    The aim of the study was to evaluate citation analysis as an approach to measuring core outcome set (COS) uptake, by assessing whether the number of citations for a COS report could be used as a surrogate measure of uptake of the COS by clinical trialists. Citation data were obtained for COS reports published before 2010 in five disease areas (systemic sclerosis, rheumatoid arthritis, eczema, sepsis and critical care, and female sexual dysfunction). Those publications identified as a report of a clinical trial were examined to identify whether or not all outcomes in the COS were measured in the trial. Clinical trials measuring the relevant COS made up a small proportion of the total number of citations for COS reports. Not all trials citing a COS report measured all the recommended outcomes. Some trials cited the COS reports for other reasons, including the definition of a condition or other trial design issues addressed by the COS report. Although citation data can be readily accessed, it should not be assumed that the citing of a COS report indicates that a trial has measured the recommended COS. Alternative methods for assessing COS uptake are needed. Copyright © 2017 The Authors. Published by Elsevier Inc. All rights reserved.

  7. The use of reliability analysis techniques applied to nuclear power station emergency core cooling systems

    International Nuclear Information System (INIS)

    Danielsen, A.; Snaith, E.R.

    1975-01-01

    A reliability investigation carried out by the Safety and Reliability Services of the UKAEA, and the SSEB, of the essential system/reactor coolant system for a large nuclear power station is described. In AGR type reactors, after all reactor shutdown conditions, it is necessary to restore forced gas circulation and sufficient boiler feed to maintain the heat removal capacity of the boilers. The coolant requirements are provided by several independent mechanical systems of primary coolant fans, feedwater pumps, and valves integrated with electrical power sources, switchgear, and automatic control equipment. Reliability is treated as one aspect of system performance and quantified in terms of failure to meet a specific objective. Based on the reliability performance of the constituent components the optimum system configuration is determined together with the preferred plant operating procedures and maintenance requirements. (author)

  8. Challenges Regarding IP Core Functional Reliability

    Science.gov (United States)

    Berg, Melanie D.; LaBel, Kenneth A.

    2017-01-01

    For many years, intellectual property (IP) cores have been incorporated into field programmable gate array (FPGA) and application specific integrated circuit (ASIC) design flows. However, the usage of large complex IP cores were limited within products that required a high level of reliability. This is no longer the case. IP core insertion has become mainstream including their use in highly reliable products. Due to limited visibility and control, challenges exist when using IP cores and subsequently compromise product reliability. We discuss challenges and suggest potential solutions to critical application IP insertion.

  9. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    International Nuclear Information System (INIS)

    Choi, Y.A.; Feltus, M.A.

    1995-01-01

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates

  10. Power electronics reliability analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Mark A.; Atcitty, Stanley

    2009-12-01

    This report provides the DOE and industry with a general process for analyzing power electronics reliability. The analysis can help with understanding the main causes of failures, downtime, and cost and how to reduce them. One approach is to collect field maintenance data and use it directly to calculate reliability metrics related to each cause. Another approach is to model the functional structure of the equipment using a fault tree to derive system reliability from component reliability. Analysis of a fictitious device demonstrates the latter process. Optimization can use the resulting baseline model to decide how to improve reliability and/or lower costs. It is recommended that both electric utilities and equipment manufacturers make provisions to collect and share data in order to lay the groundwork for improving reliability into the future. Reliability analysis helps guide reliability improvements in hardware and software technology including condition monitoring and prognostics and health management.

  11. Human reliability analysis

    International Nuclear Information System (INIS)

    Dougherty, E.M.; Fragola, J.R.

    1988-01-01

    The authors present a treatment of human reliability analysis incorporating an introduction to probabilistic risk assessment for nuclear power generating stations. They treat the subject according to the framework established for general systems theory. Draws upon reliability analysis, psychology, human factors engineering, and statistics, integrating elements of these fields within a systems framework. Provides a history of human reliability analysis, and includes examples of the application of the systems approach

  12. Multidisciplinary System Reliability Analysis

    Science.gov (United States)

    Mahadevan, Sankaran; Han, Song; Chamis, Christos C. (Technical Monitor)

    2001-01-01

    The objective of this study is to develop a new methodology for estimating the reliability of engineering systems that encompass multiple disciplines. The methodology is formulated in the context of the NESSUS probabilistic structural analysis code, developed under the leadership of NASA Glenn Research Center. The NESSUS code has been successfully applied to the reliability estimation of a variety of structural engineering systems. This study examines whether the features of NESSUS could be used to investigate the reliability of systems in other disciplines such as heat transfer, fluid mechanics, electrical circuits etc., without considerable programming effort specific to each discipline. In this study, the mechanical equivalence between system behavior models in different disciplines are investigated to achieve this objective. A new methodology is presented for the analysis of heat transfer, fluid flow, and electrical circuit problems using the structural analysis routines within NESSUS, by utilizing the equivalence between the computational quantities in different disciplines. This technique is integrated with the fast probability integration and system reliability techniques within the NESSUS code, to successfully compute the system reliability of multidisciplinary systems. Traditional as well as progressive failure analysis methods for system reliability estimation are demonstrated, through a numerical example of a heat exchanger system involving failure modes in structural, heat transfer and fluid flow disciplines.

  13. Analysis and Application of Reliability

    International Nuclear Information System (INIS)

    Jeong, Hae Seong; Park, Dong Ho; Kim, Jae Ju

    1999-05-01

    This book tells of analysis and application of reliability, which includes definition, importance and historical background of reliability, function of reliability and failure rate, life distribution and assumption of reliability, reliability of unrepaired system, reliability of repairable system, sampling test of reliability, failure analysis like failure analysis by FEMA and FTA, and cases, accelerated life testing such as basic conception, acceleration and acceleration factor, and analysis of accelerated life testing data, maintenance policy about alternation and inspection.

  14. Method of core thermodynamic reliability determination in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ackermann, G.; Horche, W. (Ingenieurhochschule Zittau (German Democratic Republic). Sektion Kraftwerksanlagenbau und Energieumwandlung)

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied.

  15. Method of core thermodynamic reliability determination in pressurized water reactors

    International Nuclear Information System (INIS)

    Ackermann, G.; Horche, W.

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied. (author)

  16. Waste package reliability analysis

    International Nuclear Information System (INIS)

    Pescatore, C.; Sastre, C.

    1983-01-01

    Proof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt. 8 references, 1 table

  17. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  18. Further HTGR core support structure reliability studies. Interim report No. 1

    International Nuclear Information System (INIS)

    Platus, D.L.

    1976-01-01

    Results of a continuing effort to investigate high temperature gas cooled reactor (HTGR) core support structure reliability are described. Graphite material and core support structure component physical, mechanical and strength properties required for the reliability analysis are identified. Also described are experimental and associated analytical techniques for determining the required properties, a procedure for determining number of tests required, properties that might be monitored by special surveillance of the core support structure to improve reliability predictions, and recommendations for further studies. Emphasis in the study is directed towards developing a basic understanding of graphite failure and strength degradation mechanisms; and validating analytical methods for predicting strength and strength degradation from basic material properties

  19. LMFBR core design analysis

    International Nuclear Information System (INIS)

    Cho, M.; Yang, J.C.; Yoh, K.C.; Suk, S.D.; Soh, D.S.; Kim, Y.M.

    1980-01-01

    The design parameters of a commercial-scale fast breeder reactor which is currently under construction by regeneration of these data is preliminary analyzed. The analysis of nuclear and thermal characteristics as well as safety features of this reactor is emphasized. And the evaluation of the initial core mentioned in the system description is carried out in the areas of its kinetics and control system, and, at the same time, the flow distribution of sodium and temperature distribution of the initial FBR core system are calculated. (KAERI INIS Section)

  20. Integrated system reliability analysis

    DEFF Research Database (Denmark)

    Gintautas, Tomas; Sørensen, John Dalsgaard

    Specific targets: 1) The report shall describe the state of the art of reliability and risk-based assessment of wind turbine components. 2) Development of methodology for reliability and risk-based assessment of the wind turbine at system level. 3) Describe quantitative and qualitative measures...

  1. Structural systems reliability analysis

    International Nuclear Information System (INIS)

    Frangopol, D.

    1975-01-01

    For an exact evaluation of the reliability of a structure it appears necessary to determine the distribution densities of the loads and resistances and to calculate the correlation coefficients between loads and between resistances. These statistical characteristics can be obtained only on the basis of a long activity period. In case that such studies are missing the statistical properties formulated here give upper and lower bounds of the reliability. (orig./HP) [de

  2. Swimming pool reactor reliability and safety analysis

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1997-01-01

    A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

  3. A reliability simulation language for reliability analysis

    International Nuclear Information System (INIS)

    Deans, N.D.; Miller, A.J.; Mann, D.P.

    1986-01-01

    The results of work being undertaken to develop a Reliability Description Language (RDL) which will enable reliability analysts to describe complex reliability problems in a simple, clear and unambiguous way are described. Component and system features can be stated in a formal manner and subsequently used, along with control statements to form a structured program. The program can be compiled and executed on a general-purpose computer or special-purpose simulator. (DG)

  4. Scyllac equipment reliability analysis

    International Nuclear Information System (INIS)

    Gutscher, W.D.; Johnson, K.J.

    1975-01-01

    Most of the failures in Scyllac can be related to crowbar trigger cable faults. A new cable has been designed, procured, and is currently undergoing evaluation. When the new cable has been proven, it will be worked into the system as quickly as possible without causing too much additional down time. The cable-tip problem may not be easy or even desirable to solve. A tightly fastened permanent connection that maximizes contact area would be more reliable than the plug-in type of connection in use now, but it would make system changes and repairs much more difficult. The balance of the failures have such a low occurrence rate that they do not cause much down time and no major effort is underway to eliminate them. Even though Scyllac was built as an experimental system and has many thousands of components, its reliability is very good. Because of this the experiment has been able to progress at a reasonable pace

  5. Integrating reliability analysis and design

    International Nuclear Information System (INIS)

    Rasmuson, D.M.

    1980-10-01

    This report describes the Interactive Reliability Analysis Project and demonstrates the advantages of using computer-aided design systems (CADS) in reliability analysis. Common cause failure problems require presentations of systems, analysis of fault trees, and evaluation of solutions to these. Results have to be communicated between the reliability analyst and the system designer. Using a computer-aided design system saves time and money in the analysis of design. Computer-aided design systems lend themselves to cable routing, valve and switch lists, pipe routing, and other component studies. At EG and G Idaho, Inc., the Applicon CADS is being applied to the study of water reactor safety systems

  6. Reliability analysis of shutdown system

    International Nuclear Information System (INIS)

    Kumar, C. Senthil; John Arul, A.; Pal Singh, Om; Suryaprakasa Rao, K.

    2005-01-01

    This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 x 10 -8 /de for failure of shutdown function in case of global faults and 4.4 x 10 -8 /de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 x 10 -6 /ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is -3 /ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability

  7. Risk analysis and reliability

    International Nuclear Information System (INIS)

    Uppuluri, V.R.R.

    1979-01-01

    Mathematical foundations of risk analysis are addressed. The importance of having the same probability space in order to compare different experiments is pointed out. Then the following topics are discussed: consequences as random variables with infinite expectations; the phenomenon of rare events; series-parallel systems and different kinds of randomness that could be imposed on such systems; and the problem of consensus of estimates of expert opinion

  8. TMI-2 core debris analysis

    International Nuclear Information System (INIS)

    Cook, B.A.; Carlson, E.R.

    1985-01-01

    One of the ongoing examination tasks for the damaged TMI-2 reactor is analysis of samples of debris obtained from the debris bed presently at the top of the core. This paper summarizes the results reported in the TMI-2 Core Debris Grab Sample Examination and Analysis Report, which will be available early in 1986. The sampling and analysis procedures are presented, and information is provided on the key results as they relate to the present core condition, peak temperatures during the transient, temperature history, chemical interactions, and core relocation. The results are then summarized

  9. Reliability analysis and operator modelling

    International Nuclear Information System (INIS)

    Hollnagel, Erik

    1996-01-01

    The paper considers the state of operator modelling in reliability analysis. Operator models are needed in reliability analysis because operators are needed in process control systems. HRA methods must therefore be able to account both for human performance variability and for the dynamics of the interaction. A selected set of first generation HRA approaches is briefly described in terms of the operator model they use, their classification principle, and the actual method they propose. In addition, two examples of second generation methods are also considered. It is concluded that first generation HRA methods generally have very simplistic operator models, either referring to the time-reliability relationship or to elementary information processing concepts. It is argued that second generation HRA methods must recognise that cognition is embedded in a context, and be able to account for that in the way human reliability is analysed and assessed

  10. Reliability Analysis of Wind Turbines

    DEFF Research Database (Denmark)

    Toft, Henrik Stensgaard; Sørensen, John Dalsgaard

    2008-01-01

    In order to minimise the total expected life-cycle costs of a wind turbine it is important to estimate the reliability level for all components in the wind turbine. This paper deals with reliability analysis for the tower and blades of onshore wind turbines placed in a wind farm. The limit states...... consideres are in the ultimate limit state (ULS) extreme conditions in the standstill position and extreme conditions during operating. For wind turbines, where the magnitude of the loads is influenced by the control system, the ultimate limit state can occur in both cases. In the fatigue limit state (FLS......) the reliability level for a wind turbine placed in a wind farm is considered, and wake effects from neighbouring wind turbines is taken into account. An illustrative example with calculation of the reliability for mudline bending of the tower is considered. In the example the design is determined according...

  11. Reliability analysis in intelligent machines

    Science.gov (United States)

    Mcinroy, John E.; Saridis, George N.

    1990-01-01

    Given an explicit task to be executed, an intelligent machine must be able to find the probability of success, or reliability, of alternative control and sensing strategies. By using concepts for information theory and reliability theory, new techniques for finding the reliability corresponding to alternative subsets of control and sensing strategies are proposed such that a desired set of specifications can be satisfied. The analysis is straightforward, provided that a set of Gaussian random state variables is available. An example problem illustrates the technique, and general reliability results are presented for visual servoing with a computed torque-control algorithm. Moreover, the example illustrates the principle of increasing precision with decreasing intelligence at the execution level of an intelligent machine.

  12. Under sodium reliability tests on core components and in-core instrumentation

    International Nuclear Information System (INIS)

    Ruppert, E.; Stehle, H.; Vinzens, K.

    1977-01-01

    A sodium test facility for fast breeder core components (AKB), built by INTERATOM at Bensberg, has been operating since 1971 to test fuel dummies and blanket elements as well as absorber elements under simulated normal and extreme reactor conditions. Individual full-scale fuel or blanket elements and arrays of seven elements, modelling a section of the SNR-300 reactor core, have been tested under a wide range of sodium mass flow and isothermal test conditions up to 925K as well as under cyclic changed temperature transients. Besides endurance testing of the core components a special sodium and high-temperature instrumentation is provided to investigate thermohydraulic and vibrational behaviour of the test objects. During all test periods the main subassembly characteristics could be reproduced and the reliability of the instrumentation could be proven. (orig.) [de

  13. RELIABILITY ANALYSIS OF BENDING ELIABILITY ANALYSIS OF ...

    African Journals Online (AJOL)

    eobe

    Reliability analysis of the safety levels of the criteria slabs, have been .... was also noted [2] that if the risk level or β < 3.1), the ... reliability analysis. A study [6] has shown that all geometric variables, ..... Germany, 1988. 12. Hasofer, A. M and ...

  14. Reliability analysis under epistemic uncertainty

    International Nuclear Information System (INIS)

    Nannapaneni, Saideep; Mahadevan, Sankaran

    2016-01-01

    This paper proposes a probabilistic framework to include both aleatory and epistemic uncertainty within model-based reliability estimation of engineering systems for individual limit states. Epistemic uncertainty is considered due to both data and model sources. Sparse point and/or interval data regarding the input random variables leads to uncertainty regarding their distribution types, distribution parameters, and correlations; this statistical uncertainty is included in the reliability analysis through a combination of likelihood-based representation, Bayesian hypothesis testing, and Bayesian model averaging techniques. Model errors, which include numerical solution errors and model form errors, are quantified through Gaussian process models and included in the reliability analysis. The probability integral transform is used to develop an auxiliary variable approach that facilitates a single-level representation of both aleatory and epistemic uncertainty. This strategy results in an efficient single-loop implementation of Monte Carlo simulation (MCS) and FORM/SORM techniques for reliability estimation under both aleatory and epistemic uncertainty. Two engineering examples are used to demonstrate the proposed methodology. - Highlights: • Epistemic uncertainty due to data and model included in reliability analysis. • A novel FORM-based approach proposed to include aleatory and epistemic uncertainty. • A single-loop Monte Carlo approach proposed to include both types of uncertainties. • Two engineering examples used for illustration.

  15. Reliability analysis techniques in power plant design

    International Nuclear Information System (INIS)

    Chang, N.E.

    1981-01-01

    An overview of reliability analysis techniques is presented as applied to power plant design. The key terms, power plant performance, reliability, availability and maintainability are defined. Reliability modeling, methods of analysis and component reliability data are briefly reviewed. Application of reliability analysis techniques from a design engineering approach to improving power plant productivity is discussed. (author)

  16. Update Knowledge Base for Long-term Core Cooling Reliability

    International Nuclear Information System (INIS)

    Agrell, Maria; Sandervag, Oddbjoern; Amri, Abdallah; ); Bang, Young S.; Blomart, Philippe; Broecker, Annette; Pointner, Winfried; Ganzmann, Ingo; Lenogue, Bruno; Guzonas, David; Herer, Christophe; Mattei, Jean-Marie; Tricottet, Matthieu; Masaoka, Hideaki; Soltesz, Vojtech; Tarkiainen, Seppo; Ui, Atsushi; Villalba, Cristina; Zigler, Gilbert

    2013-11-01

    This revision of the Knowledge Base for Emergency Core Cooling System Recirculation Reliability (NEA/CSNI/R (95)11) describes the current status (late 2012) of the knowledge base on emergency core cooling system (ECCS) and containment spray system (CSS) suction strainer performance and long-term cooling in operating power reactors. New reactors, such as the AP1000, EPR and APR1400 that are under construction in some Organization for Economic Co-operation and Development (OECD) member countries, are not addressed in detail in this revision. The containment sump (also known as the emergency or recirculation sump in pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs) or the suppression pools or wet wells in boiling water reactors (BWRs)) and associated ECCS strainers are parts of the ECCS in both reactor types. All nuclear power plants (NPPs) are required to have an ECCS that is capable of mitigating a design basis accident (DBA). The containment sump collects reactor coolant, ECCS injection water, and containment spray solutions, if applicable, after a loss-of-coolant accident (LOCA). The sump serves as the water source to support long-term recirculation for residual heat removal, emergency core cooling, and containment atmosphere clean-up. This water source, the related pump suction inlets, and the piping between the source and inlets are important safety-related components. In addition, if fibrous material is deposited at the fuel element spacers, core cooling can be endangered. The performance of ECCS/CSS strainers was recognized many years ago as an important regulatory and safety issue. One of the primary concerns is the potential for debris generated by a jet of high-pressure coolant during a LOCA to clog the strainer and obstruct core cooling. The issue was considered resolved for all reactor types in the mid-1990s and the OECD/NEA/CSNI published report NEA/CSNI/R(95)11 in 1996 to document the state of knowledge of ECCS performance

  17. Human Reliability Analysis: session summary

    International Nuclear Information System (INIS)

    Hall, R.E.

    1985-01-01

    The use of Human Reliability Analysis (HRA) to identify and resolve human factors issues has significantly increased over the past two years. Today, utilities, research institutions, consulting firms, and the regulatory agency have found a common application of HRA tools and Probabilistic Risk Assessment (PRA). The ''1985 IEEE Third Conference on Human Factors and Power Plants'' devoted three sessions to the discussion of these applications and a review of the insights so gained. This paper summarizes the three sessions and presents those common conclusions that were discussed during the meeting. The paper concludes that session participants supported the use of an adequately documented ''living PRA'' to address human factors issues in design and procedural changes, regulatory compliance, and training and that the techniques can produce cost effective qualitative results that are complementary to more classical human factors methods

  18. Contributions to the determination of the thermal core reliability of pressurized water reactors

    International Nuclear Information System (INIS)

    Ackermann, G.; Horche, W.; Melchior, H.; Prasser, H.M.

    1982-09-01

    The investigations in the field of thermohydraulics of PW reactors are aimed at a possible increase of economy and reliability of WWER-type-reactors. In detail the flow distribution at the core entrance, the modification of the power distribution as a result of an irregular temperature distribution at the core entrance, and based on the theory of hot spots the thermic core reliability are studied. In this connection qualitatively new methods are applied characterized by low expenditure. (author)

  19. Fundamentals and applications of systems reliability analysis

    International Nuclear Information System (INIS)

    Boesebeck, K.; Heuser, F.W.; Kotthoff, K.

    1976-01-01

    The lecture gives a survey on the application of methods of reliability analysis to assess the safety of nuclear power plants. Possible statements of reliability analysis in connection with specifications of the atomic licensing procedure are especially dealt with. Existing specifications of safety criteria are additionally discussed with the help of reliability analysis by the example of the reliability analysis of a reactor protection system. Beyond the limited application to single safety systems, the significance of reliability analysis for a closed risk concept is explained in the last part of the lecture. (orig./LH) [de

  20. On Bayesian System Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Soerensen Ringi, M

    1995-05-01

    The view taken in this thesis is that reliability, the probability that a system will perform a required function for a stated period of time, depends on a person`s state of knowledge. Reliability changes as this state of knowledge changes, i.e. when new relevant information becomes available. Most existing models for system reliability prediction are developed in a classical framework of probability theory and they overlook some information that is always present. Probability is just an analytical tool to handle uncertainty, based on judgement and subjective opinions. It is argued that the Bayesian approach gives a much more comprehensive understanding of the foundations of probability than the so called frequentistic school. A new model for system reliability prediction is given in two papers. The model encloses the fact that component failures are dependent because of a shared operational environment. The suggested model also naturally permits learning from failure data of similar components in non identical environments. 85 refs.

  1. On Bayesian System Reliability Analysis

    International Nuclear Information System (INIS)

    Soerensen Ringi, M.

    1995-01-01

    The view taken in this thesis is that reliability, the probability that a system will perform a required function for a stated period of time, depends on a person's state of knowledge. Reliability changes as this state of knowledge changes, i.e. when new relevant information becomes available. Most existing models for system reliability prediction are developed in a classical framework of probability theory and they overlook some information that is always present. Probability is just an analytical tool to handle uncertainty, based on judgement and subjective opinions. It is argued that the Bayesian approach gives a much more comprehensive understanding of the foundations of probability than the so called frequentistic school. A new model for system reliability prediction is given in two papers. The model encloses the fact that component failures are dependent because of a shared operational environment. The suggested model also naturally permits learning from failure data of similar components in non identical environments. 85 refs

  2. Cost analysis of reliability investigations

    International Nuclear Information System (INIS)

    Schmidt, F.

    1981-01-01

    Taking Epsteins testing theory as a basis, premisses are formulated for the selection of cost-optimized reliability inspection plans. Using an example, the expected testing costs and inspection time periods of various inspection plan types, standardized on the basis of the exponential distribution, are compared. It can be shown that sequential reliability tests usually involve lower costs than failure or time-fixed tests. The most 'costly' test is to be expected with the inspection plan type NOt. (orig.) [de

  3. The validity and reliability of a dynamic neuromuscular stabilization-heel sliding test for core stability.

    Science.gov (United States)

    Cha, Young Joo; Lee, Jae Jin; Kim, Do Hyun; You, Joshua Sung H

    2017-10-23

    Core stabilization plays an important role in the regulation of postural stability. To overcome shortcomings associated with pain and severe core instability during conventional core stabilization tests, we recently developed the dynamic neuromuscular stabilization-based heel sliding (DNS-HS) test. The purpose of this study was to establish the criterion validity and test-retest reliability of the novel DNS-HS test. Twenty young adults with core instability completed both the bilateral straight leg lowering test (BSLLT) and DNS-HS test for the criterion validity study and repeated the DNS-HS test for the test-retest reliability study. Criterion validity was determined by comparing hip joint angle data that were obtained from BSLLT and DNS-HS measures. The test-retest reliability was determined by comparing hip joint angle data. Criterion validity was (ICC2,3) = 0.700 (preliability was (ICC3,3) = 0.953 (pvalidity data demonstrated a good relationship between the gold standard BSLLT and DNS-HS core stability measures. Test-retest reliability data suggests that DNS-HS core stability was a reliable test for core stability. Clinically, the DNS-HS test is useful to objectively quantify core instability and allow early detection and evaluation.

  4. Diakoptical reliability analysis of transistorized systems

    International Nuclear Information System (INIS)

    Kontoleon, J.M.; Lynn, J.W.; Green, A.E.

    1975-01-01

    Limitations both on high-speed core availability and computation time required for assessing the reliability of large-sized and complex electronic systems, such as used for the protection of nuclear reactors, are very serious restrictions which continuously confront the reliability analyst. Diakoptic methods simplify the solution of the electrical-network problem by subdividing a given network into a number of independent subnetworks and then interconnecting the solutions of these smaller parts by a systematic process involving transformations based on connection-matrix elements associated with the interconnecting links. However, the interconnection process is very complicated and it may be used only if the original system has been cut in such a manner that a relation can be established between the constraints appearing at both sides of the cut. Also, in dealing with transistorized systems, one of the difficulties encountered is that of modelling adequately their performance under various operating conditions, since their parameters are strongly affected by the imposed voltage and current levels. In this paper a new interconnection approach is presented which may be of use in the reliability analysis of large-sized transistorized systems. This is based on the partial optimization of the subdivisions of the torn network as well as on the optimization of the torn paths. The solution of the subdivisions is based on the principles of algebraic topology, with an algebraic structure relating the physical variables in a topological structure which defines the interconnection of the discrete elements. Transistors, and other nonlinear devices, are modelled using their actual characteristics, under normal and abnormal operating conditions. Use of so-called k factors is made to facilitate accounting for use of electrical stresses. The approach is demonstrated by way of an example. (author)

  5. Reliability Analysis of Money Habitudes

    Science.gov (United States)

    Delgadillo, Lucy M.; Bushman, Brittani S.

    2015-01-01

    Use of the Money Habitudes exercise has gained popularity among various financial professionals. This article reports on the reliability of this resource. A survey administered to young adults at a western state university was conducted, and each Habitude or "domain" was analyzed using Cronbach's alpha procedures. Results showed all six…

  6. Power system reliability analysis using fault trees

    International Nuclear Information System (INIS)

    Volkanovski, A.; Cepin, M.; Mavko, B.

    2006-01-01

    The power system reliability analysis method is developed from the aspect of reliable delivery of electrical energy to customers. The method is developed based on the fault tree analysis, which is widely applied in the Probabilistic Safety Assessment (PSA). The method is adapted for the power system reliability analysis. The method is developed in a way that only the basic reliability parameters of the analysed power system are necessary as an input for the calculation of reliability indices of the system. The modeling and analysis was performed on an example power system consisting of eight substations. The results include the level of reliability of current power system configuration, the combinations of component failures resulting in a failed power delivery to loads, and the importance factors for components and subsystems. (author)

  7. Reliability analysis of software based safety functions

    International Nuclear Information System (INIS)

    Pulkkinen, U.

    1993-05-01

    The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

  8. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  9. Reliability Analysis of Adhesive Bonded Scarf Joints

    DEFF Research Database (Denmark)

    Kimiaeifar, Amin; Toft, Henrik Stensgaard; Lund, Erik

    2012-01-01

    element analysis (FEA). For the reliability analysis a design equation is considered which is related to a deterministic code-based design equation where reliability is secured by partial safety factors together with characteristic values for the material properties and loads. The failure criteria......A probabilistic model for the reliability analysis of adhesive bonded scarfed lap joints subjected to static loading is developed. It is representative for the main laminate in a wind turbine blade subjected to flapwise bending. The structural analysis is based on a three dimensional (3D) finite...... are formulated using a von Mises, a modified von Mises and a maximum stress failure criterion. The reliability level is estimated for the scarfed lap joint and this is compared with the target reliability level implicitly used in the wind turbine standard IEC 61400-1. A convergence study is performed to validate...

  10. Core analysis: new features and applications

    International Nuclear Information System (INIS)

    Edenius, M.; Kurcyusz, E.; Molina, D.; Wiksell, G.

    1995-01-01

    Today, core analysis may be performed with sophisticated software capable of both steady state and transient analysis using a common methodology for BWRs and PWRs. General trends in core analysis software development are: improved accuracy, automated engineering functions; three-dimensional transient capability; graphical user interfaces. As a demonstration of such software, new features of Studsvik-CMS (Core management system) and examples of applications are discussed in this article. 2 figs., 8 refs

  11. System Reliability Analysis Considering Correlation of Performances

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Saekyeol; Lee, Tae Hee [Hanyang Univ., Seoul (Korea, Republic of); Lim, Woochul [Mando Corporation, Seongnam (Korea, Republic of)

    2017-04-15

    Reliability analysis of a mechanical system has been developed in order to consider the uncertainties in the product design that may occur from the tolerance of design variables, uncertainties of noise, environmental factors, and material properties. In most of the previous studies, the reliability was calculated independently for each performance of the system. However, the conventional methods cannot consider the correlation between the performances of the system that may lead to a difference between the reliability of the entire system and the reliability of the individual performance. In this paper, the joint probability density function (PDF) of the performances is modeled using a copula which takes into account the correlation between performances of the system. The system reliability is proposed as the integral of joint PDF of performances and is compared with the individual reliability of each performance by mathematical examples and two-bar truss example.

  12. System Reliability Analysis Considering Correlation of Performances

    International Nuclear Information System (INIS)

    Kim, Saekyeol; Lee, Tae Hee; Lim, Woochul

    2017-01-01

    Reliability analysis of a mechanical system has been developed in order to consider the uncertainties in the product design that may occur from the tolerance of design variables, uncertainties of noise, environmental factors, and material properties. In most of the previous studies, the reliability was calculated independently for each performance of the system. However, the conventional methods cannot consider the correlation between the performances of the system that may lead to a difference between the reliability of the entire system and the reliability of the individual performance. In this paper, the joint probability density function (PDF) of the performances is modeled using a copula which takes into account the correlation between performances of the system. The system reliability is proposed as the integral of joint PDF of performances and is compared with the individual reliability of each performance by mathematical examples and two-bar truss example.

  13. Reliability analysis using network simulation

    International Nuclear Information System (INIS)

    Engi, D.

    1985-01-01

    The models that can be used to provide estimates of the reliability of nuclear power systems operate at many different levels of sophistication. The least-sophisticated models treat failure processes that entail only time-independent phenomena (such as demand failure). More advanced models treat processes that also include time-dependent phenomena such as run failure and possibly repair. However, many of these dynamic models are deficient in some respects because they either disregard the time-dependent phenomena that cannot be expressed in closed-form analytic terms or because they treat these phenomena in quasi-static terms. The next level of modeling requires a dynamic approach that incorporates not only procedures for treating all significant time-dependent phenomena but also procedures for treating these phenomena when they are conditionally linked or characterized by arbitrarily selected probability distributions. The level of sophistication that is required is provided by a dynamic, Monte Carlo modeling approach. A computer code that uses a dynamic, Monte Carlo modeling approach is Q-GERT (Graphical Evaluation and Review Technique - with Queueing), and the present study had demonstrated the feasibility of using Q-GERT for modeling time-dependent, unconditionally and conditionally linked phenomena that are characterized by arbitrarily selected probability distributions

  14. Analysis of information security reliability: A tutorial

    International Nuclear Information System (INIS)

    Kondakci, Suleyman

    2015-01-01

    This article presents a concise reliability analysis of network security abstracted from stochastic modeling, reliability, and queuing theories. Network security analysis is composed of threats, their impacts, and recovery of the failed systems. A unique framework with a collection of the key reliability models is presented here to guide the determination of the system reliability based on the strength of malicious acts and performance of the recovery processes. A unique model, called Attack-obstacle model, is also proposed here for analyzing systems with immunity growth features. Most computer science curricula do not contain courses in reliability modeling applicable to different areas of computer engineering. Hence, the topic of reliability analysis is often too diffuse to most computer engineers and researchers dealing with network security. This work is thus aimed at shedding some light on this issue, which can be useful in identifying models, their assumptions and practical parameters for estimating the reliability of threatened systems and for assessing the performance of recovery facilities. It can also be useful for the classification of processes and states regarding the reliability of information systems. Systems with stochastic behaviors undergoing queue operations and random state transitions can also benefit from the approaches presented here. - Highlights: • A concise survey and tutorial in model-based reliability analysis applicable to information security. • A framework of key modeling approaches for assessing reliability of networked systems. • The framework facilitates quantitative risk assessment tasks guided by stochastic modeling and queuing theory. • Evaluation of approaches and models for modeling threats, failures, impacts, and recovery analysis of information systems

  15. State of the art report on aging reliability analysis

    International Nuclear Information System (INIS)

    Choi, Sun Yeong; Yang, Joon Eon; Han, Sang Hoon; Ha, Jae Joo

    2002-03-01

    The goal of this report is to describe the state of the art on aging analysis methods to calculate the effects of component aging quantitatively. In this report, we described some aging analysis methods which calculate the increase of Core Damage Frequency (CDF) due to aging by including the influence of aging into PSA. We also described several research topics required for aging analysis for components of domestic NPPs. We have described a statistical model and reliability physics model which calculate the effect of aging quantitatively by using PSA method. It is expected that the practical use of the reliability-physics model will be increased though the process with the reliability-physics model is more complicated than statistical model

  16. Reliability analysis of the reactor protection system with fault diagnosis

    International Nuclear Information System (INIS)

    Lee, D.Y.; Han, J.B.; Lyou, J.

    2004-01-01

    The main function of a reactor protection system (RPS) is to maintain the reactor core integrity and reactor coolant system pressure boundary. The RPS consists of the 2-out-of-m redundant architecture to assure a reliable operation. The system reliability of the RPS is a very important factor for the probability safety assessment (PSA) evaluation in the nuclear field. To evaluate the system failure rate of the k-out-of-m redundant system is not so easy with the deterministic method. In this paper, the reliability analysis method using the binomial process is suggested to calculate the failure rate of the RPS system with a fault diagnosis function. The suggested method is compared with the result of the Markov process to verify the validation of the suggested method, and applied to the several kinds of RPS architectures for a comparative evaluation of the reliability. (orig.)

  17. Space Mission Human Reliability Analysis (HRA) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The purpose of this project is to extend current ground-based Human Reliability Analysis (HRA) techniques to a long-duration, space-based tool to more effectively...

  18. Reliability analysis of stiff versus flexible piping

    International Nuclear Information System (INIS)

    Lu, S.C.

    1985-01-01

    The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. The authors then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. They concluded that these changes substantially reduce calculated piping responses and allow piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the original stiffer design. An investigation of the malfunction of pipe whip restraints confirmed that the malfunction introduced higher thermal stresses and tended to reduce the overall piping reliability. Finally, support and component reliabilities were evaluated based on available fragility data. Results indicated that the support reliability usually exhibits a moderate decrease as the piping flexibility increases. Most on-line pumps and valves showed an insignificant reduction in reliability for a more flexible piping design

  19. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  20. Reliability and validity of risk analysis

    International Nuclear Information System (INIS)

    Aven, Terje; Heide, Bjornar

    2009-01-01

    In this paper we investigate to what extent risk analysis meets the scientific quality requirements of reliability and validity. We distinguish between two types of approaches within risk analysis, relative frequency-based approaches and Bayesian approaches. The former category includes both traditional statistical inference methods and the so-called probability of frequency approach. Depending on the risk analysis approach, the aim of the analysis is different, the results are presented in different ways and consequently the meaning of the concepts reliability and validity are not the same.

  1. Multi-Disciplinary System Reliability Analysis

    Science.gov (United States)

    Mahadevan, Sankaran; Han, Song

    1997-01-01

    The objective of this study is to develop a new methodology for estimating the reliability of engineering systems that encompass multiple disciplines. The methodology is formulated in the context of the NESSUS probabilistic structural analysis code developed under the leadership of NASA Lewis Research Center. The NESSUS code has been successfully applied to the reliability estimation of a variety of structural engineering systems. This study examines whether the features of NESSUS could be used to investigate the reliability of systems in other disciplines such as heat transfer, fluid mechanics, electrical circuits etc., without considerable programming effort specific to each discipline. In this study, the mechanical equivalence between system behavior models in different disciplines are investigated to achieve this objective. A new methodology is presented for the analysis of heat transfer, fluid flow, and electrical circuit problems using the structural analysis routines within NESSUS, by utilizing the equivalence between the computational quantities in different disciplines. This technique is integrated with the fast probability integration and system reliability techniques within the NESSUS code, to successfully compute the system reliability of multi-disciplinary systems. Traditional as well as progressive failure analysis methods for system reliability estimation are demonstrated, through a numerical example of a heat exchanger system involving failure modes in structural, heat transfer and fluid flow disciplines.

  2. Reliability analysis techniques for the design engineer

    International Nuclear Information System (INIS)

    Corran, E.R.; Witt, H.H.

    1982-01-01

    This paper describes a fault tree analysis package that eliminates most of the housekeeping tasks involved in proceeding from the initial construction of a fault tree to the final stage of presenting a reliability analysis in a safety report. It is suitable for designers with relatively little training in reliability analysis and computer operation. Users can rapidly investigate the reliability implications of various options at the design stage and evolve a system which meets specified reliability objectives. Later independent review is thus unlikely to reveal major shortcomings necessitating modification and project delays. The package operates interactively, allowing the user to concentrate on the creative task of developing the system fault tree, which may be modified and displayed graphically. For preliminary analysis, system data can be derived automatically from a generic data bank. As the analysis proceeds, improved estimates of critical failure rates and test and maintenance schedules can be inserted. The technique is applied to the reliability analysis of the recently upgraded HIFAR Containment Isolation System. (author)

  3. Reliability analysis of Angra I safety systems

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Soto, J.B.; Maciel, C.C.; Gibelli, S.M.O.; Fleming, P.V.; Arrieta, L.A.

    1980-07-01

    An extensive reliability analysis of some safety systems of Angra I, are presented. The fault tree technique, which has been successfully used in most reliability studies of nuclear safety systems performed to date is employed. Results of a quantitative determination of the unvailability of the accumulator and the containment spray injection systems are presented. These results are also compared to those reported in WASH-1400. (E.G.) [pt

  4. Reliability analysis techniques for the design engineer

    International Nuclear Information System (INIS)

    Corran, E.R.; Witt, H.H.

    1980-01-01

    A fault tree analysis package is described that eliminates most of the housekeeping tasks involved in proceeding from the initial construction of a fault tree to the final stage of presenting a reliability analysis in a safety report. It is suitable for designers with relatively little training in reliability analysis and computer operation. Users can rapidly investigate the reliability implications of various options at the design stage, and evolve a system which meets specified reliability objectives. Later independent review is thus unlikely to reveal major shortcomings necessitating modification and projects delays. The package operates interactively allowing the user to concentrate on the creative task of developing the system fault tree, which may be modified and displayed graphically. For preliminary analysis system data can be derived automatically from a generic data bank. As the analysis procedes improved estimates of critical failure rates and test and maintenance schedules can be inserted. The computations are standard, - identification of minimal cut-sets, estimation of reliability parameters, and ranking of the effect of the individual component failure modes and system failure modes on these parameters. The user can vary the fault trees and data on-line, and print selected data for preferred systems in a form suitable for inclusion in safety reports. A case history is given - that of HIFAR containment isolation system. (author)

  5. Fault tree analysis on BWR core spray system

    International Nuclear Information System (INIS)

    Watanabe, Norio

    1982-06-01

    Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10 - 3 /demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

  6. An analysis of the uniform core experiment

    Energy Technology Data Exchange (ETDEWEB)

    Waterson, R H

    1973-10-15

    This report describes an analysis of the Uniform Core of HITREX using the WIMS E codes, and presents the results of theory/experiment comparisons. The overall picture is one of good agreement for core reaction rate distributions, but theory umderestimating k{sub eff} by about 1.5% {delta}k/k.

  7. Reliability Analysis of Elasto-Plastic Structures

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle; Sørensen, John Dalsgaard

    1984-01-01

    . Failure of this type of system is defined either as formation of a mechanism or by failure of a prescribed number of elements. In the first case failure is independent of the order in which the elements fail, but this is not so by the second definition. The reliability analysis consists of two parts...... are described and the two definitions of failure can be used by the first formulation, but only the failure definition based on formation of a mechanism by the second formulation. The second part of the reliability analysis is an estimate of the failure probability for the structure on the basis...

  8. Reliability analysis and assessment of structural systems

    International Nuclear Information System (INIS)

    Yao, J.T.P.; Anderson, C.A.

    1977-01-01

    The study of structural reliability deals with the probability of having satisfactory performance of the structure under consideration within any specific time period. To pursue this study, it is necessary to apply available knowledge and methodology in structural analysis (including dynamics) and design, behavior of materials and structures, experimental mechanics, and the theory of probability and statistics. In addition, various severe loading phenomena such as strong motion earthquakes and wind storms are important considerations. For three decades now, much work has been done on reliability analysis of structures, and during this past decade, certain so-called 'Level I' reliability-based design codes have been proposed and are in various stages of implementation. These contributions will be critically reviewed and summarized in this paper. Because of the undesirable consequences resulting from the failure of nuclear structures, it is important and desirable to consider the structural reliability in the analysis and design of these structures. Moreover, after these nuclear structures are constructed, it is desirable for engineers to be able to assess the structural reliability periodically as well as immediately following the occurrence of severe loading conditions such as a strong-motion earthquake. During this past decade, increasing use has been made of techniques of system identification in structural engineering. On the basis of non-destructive test results, various methods have been developed to obtain an adequate mathematical model (such as the equations of motion with more realistic parameters) to represent the structural system

  9. Culture Representation in Human Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Gertman; Julie Marble; Steven Novack

    2006-12-01

    Understanding human-system response is critical to being able to plan and predict mission success in the modern battlespace. Commonly, human reliability analysis has been used to predict failures of human performance in complex, critical systems. However, most human reliability methods fail to take culture into account. This paper takes an easily understood state of the art human reliability analysis method and extends that method to account for the influence of culture, including acceptance of new technology, upon performance. The cultural parameters used to modify the human reliability analysis were determined from two standard industry approaches to cultural assessment: Hofstede’s (1991) cultural factors and Davis’ (1989) technology acceptance model (TAM). The result is called the Culture Adjustment Method (CAM). An example is presented that (1) reviews human reliability assessment with and without cultural attributes for a Supervisory Control and Data Acquisition (SCADA) system attack, (2) demonstrates how country specific information can be used to increase the realism of HRA modeling, and (3) discusses the differences in human error probability estimates arising from cultural differences.

  10. Reliability of core test – Critical assessment and proposed new approach

    OpenAIRE

    Shafik Khoury; Ali Abdel-Hakam Aliabdo; Ahmed Ghazy

    2014-01-01

    Core test is commonly required in the area of concrete industry to evaluate the concrete strength and sometimes it becomes the unique tool for safety assessment of existing concrete structures. Core test is therefore introduced in most codes. An extensive literature survey on different international codes’ provisions; including the Egyptian, British, European and ACI Codes, for core analysis is presented. All studied codes’ provisions seem to be unreliable for predicting the in-situ concrete ...

  11. Reliability Analysis of a Steel Frame

    Directory of Open Access Journals (Sweden)

    M. Sýkora

    2002-01-01

    Full Text Available A steel frame with haunches is designed according to Eurocodes. The frame is exposed to self-weight, snow, and wind actions. Lateral-torsional buckling appears to represent the most critical criterion, which is considered as a basis for the limit state function. In the reliability analysis, the probabilistic models proposed by the Joint Committee for Structural Safety (JCSS are used for basic variables. The uncertainty model coefficients take into account the inaccuracy of the resistance model for the haunched girder and the inaccuracy of the action effect model. The time invariant reliability analysis is based on Turkstra's rule for combinations of snow and wind actions. The time variant analysis describes snow and wind actions by jump processes with intermittencies. Assuming a 50-year lifetime, the obtained values of the reliability index b vary within the range from 3.95 up to 5.56. The cross-profile IPE 330 designed according to Eurocodes seems to be adequate. It appears that the time invariant reliability analysis based on Turkstra's rule provides considerably lower values of b than those obtained by the time variant analysis.

  12. Representative Sampling for reliable data analysis

    DEFF Research Database (Denmark)

    Petersen, Lars; Esbensen, Kim Harry

    2005-01-01

    regime in order to secure the necessary reliability of: samples (which must be representative, from the primary sampling onwards), analysis (which will not mean anything outside the miniscule analytical volume without representativity ruling all mass reductions involved, also in the laboratory) and data...

  13. Reliability analysis of prestressed concrete containment structures

    International Nuclear Information System (INIS)

    Jiang, J.; Zhao, Y.; Sun, J.

    1993-01-01

    The reliability analysis of prestressed concrete containment structures subjected to combinations of static and dynamic loads with consideration of uncertainties of structural and load parameters is presented. Limit state probabilities for given parameters are calculated using the procedure developed at BNL, while that with consideration of parameter uncertainties are calculated by a fast integration for time variant structural reliability. The limit state surface of the prestressed concrete containment is constructed directly incorporating the prestress. The sensitivities of the Choleskey decomposition matrix and the natural vibration character are calculated by simplified procedures. (author)

  14. Prime implicants in dynamic reliability analysis

    International Nuclear Information System (INIS)

    Tyrväinen, Tero

    2016-01-01

    This paper develops an improved definition of a prime implicant for the needs of dynamic reliability analysis. Reliability analyses often aim to identify minimal cut sets or prime implicants, which are minimal conditions that cause an undesired top event, such as a system's failure. Dynamic reliability analysis methods take the time-dependent behaviour of a system into account. This means that the state of a component can change in the analysed time frame and prime implicants can include the failure of a component at different time points. There can also be dynamic constraints on a component's behaviour. For example, a component can be non-repairable in the given time frame. If a non-repairable component needs to be failed at a certain time point to cause the top event, we consider that the condition that it is failed at the latest possible time point is minimal, and the condition in which it fails earlier non-minimal. The traditional definition of a prime implicant does not account for this type of time-related minimality. In this paper, a new definition is introduced and illustrated using a dynamic flowgraph methodology model. - Highlights: • A new definition of a prime implicant is developed for dynamic reliability analysis. • The new definition takes time-related minimality into account. • The new definition is needed in dynamic flowgraph methodology. • Results can be represented by a smaller number of prime implicants.

  15. Reliability and risk analysis methods research plan

    International Nuclear Information System (INIS)

    1984-10-01

    This document presents a plan for reliability and risk analysis methods research to be performed mainly by the Reactor Risk Branch (RRB), Division of Risk Analysis and Operations (DRAO), Office of Nuclear Regulatory Research. It includes those activities of other DRAO branches which are very closely related to those of the RRB. Related or interfacing programs of other divisions, offices and organizations are merely indicated. The primary use of this document is envisioned as an NRC working document, covering about a 3-year period, to foster better coordination in reliability and risk analysis methods development between the offices of Nuclear Regulatory Research and Nuclear Reactor Regulation. It will also serve as an information source for contractors and others to more clearly understand the objectives, needs, programmatic activities and interfaces together with the overall logical structure of the program

  16. Human reliability analysis of control room operators

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Isaac J.A.L.; Carvalho, Paulo Victor R.; Grecco, Claudio H.S. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    2005-07-01

    Human reliability is the probability that a person correctly performs some system required action in a required time period and performs no extraneous action that can degrade the system Human reliability analysis (HRA) is the analysis, prediction and evaluation of work-oriented human performance using some indices as human error likelihood and probability of task accomplishment. Significant progress has been made in the HRA field during the last years, mainly in nuclear area. Some first-generation HRA methods were developed, as THERP (Technique for human error rate prediction). Now, an array of called second-generation methods are emerging as alternatives, for instance ATHEANA (A Technique for human event analysis). The ergonomics approach has as tool the ergonomic work analysis. It focus on the study of operator's activities in physical and mental form, considering at the same time the observed characteristics of operator and the elements of the work environment as they are presented to and perceived by the operators. The aim of this paper is to propose a methodology to analyze the human reliability of the operators of industrial plant control room, using a framework that includes the approach used by ATHEANA, THERP and the work ergonomics analysis. (author)

  17. Modified Core Wash Cytology: A reliable same day biopsy result for breast clinics.

    Science.gov (United States)

    Bulte, J P; Wauters, C A P; Duijm, L E M; de Wilt, J H W; Strobbe, L J A

    2016-12-01

    Fine Needle Aspiration Biopsy (FNAB), Core Needle biopsy (CNB) and hybrid techniques including Core Wash Cytology (CWC) are available for same-day diagnosis in breast lesions. In CWC a washing of the biopsy core is processed for a provisional cytological diagnosis, after which the core is processed like a regular CNB. This study focuses on the reliability of CWC in daily practice. All consecutive CWC procedures performed in a referral breast centre between May 2009 and May 2012 were reviewed, correlating CWC results with the CNB result, definitive diagnosis after surgical resection and/or follow-up. Symptomatic as well as screen-detected lesions, undergoing CNB were included. 1253 CWC procedures were performed. Definitive histology showed 849 (68%) malignant and 404 (32%) benign lesions. 80% of CWC procedures yielded a conclusive diagnosis: this percentage was higher amongst malignant lesions and lower for benign lesions: 89% and 62% respectively. Sensitivity and specificity of a conclusive CWC result were respectively 98.3% and 90.4%. The eventual incidence of malignancy in the cytological 'atypical' group (5%) was similar to the cytological 'benign' group (6%). CWC can be used to make a reliable provisional diagnosis of breast lesions within the hour. The high probability of conclusive results in malignant lesions makes CWC well suited for high risk populations. Copyright © 2016 Elsevier Ltd, BASO ~ the Association for Cancer Surgery, and the European Society of Surgical Oncology. All rights reserved.

  18. Using the graphs models for evaluating in-core monitoring systems reliability by the method of imiting simulaton

    International Nuclear Information System (INIS)

    Golovanov, M.N.; Zyuzin, N.N.; Levin, G.L.; Chesnokov, A.N.

    1987-01-01

    An approach for estimation of reliability factors of complex reserved systems at early stages of development using the method of imitating simulation is considered. Different types of models, their merits and lacks are given. Features of in-core monitoring systems and advosability of graph model and graph theory element application for estimating reliability of such systems are shown. The results of investigation of the reliability factors of the reactor monitoring, control and core local protection subsystem are shown

  19. Sensitivity analysis in a structural reliability context

    International Nuclear Information System (INIS)

    Lemaitre, Paul

    2014-01-01

    This thesis' subject is sensitivity analysis in a structural reliability context. The general framework is the study of a deterministic numerical model that allows to reproduce a complex physical phenomenon. The aim of a reliability study is to estimate the failure probability of the system from the numerical model and the uncertainties of the inputs. In this context, the quantification of the impact of the uncertainty of each input parameter on the output might be of interest. This step is called sensitivity analysis. Many scientific works deal with this topic but not in the reliability scope. This thesis' aim is to test existing sensitivity analysis methods, and to propose more efficient original methods. A bibliographical step on sensitivity analysis on one hand and on the estimation of small failure probabilities on the other hand is first proposed. This step raises the need to develop appropriate techniques. Two variables ranking methods are then explored. The first one proposes to make use of binary classifiers (random forests). The second one measures the departure, at each step of a subset method, between each input original density and the density given the subset reached. A more general and original methodology reflecting the impact of the input density modification on the failure probability is then explored. The proposed methods are then applied on the CWNR case, which motivates this thesis. (author)

  20. Human reliability analysis using event trees

    International Nuclear Information System (INIS)

    Heslinga, G.

    1983-01-01

    The shut-down procedure of a technologically complex installation as a nuclear power plant consists of a lot of human actions, some of which have to be performed several times. The procedure is regarded as a chain of modules of specific actions, some of which are analyzed separately. The analysis is carried out by making a Human Reliability Analysis event tree (HRA event tree) of each action, breaking down each action into small elementary steps. The application of event trees in human reliability analysis implies more difficulties than in the case of technical systems where event trees were mainly used until now. The most important reason is that the operator is able to recover a wrong performance; memory influences play a significant role. In this study these difficulties are dealt with theoretically. The following conclusions can be drawn: (1) in principle event trees may be used in human reliability analysis; (2) although in practice the operator will recover his fault partly, theoretically this can be described as starting the whole event tree again; (3) compact formulas have been derived, by which the probability of reaching a specific failure consequence on passing through the HRA event tree after several times of recovery is to be calculated. (orig.)

  1. Structural reliability analysis and seismic risk assessment

    International Nuclear Information System (INIS)

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

    1984-01-01

    This paper presents a reliability analysis method for safety evaluation of nuclear structures. By utilizing this method, it is possible to estimate the limit state probability in the lifetime of structures and to generate analytically the fragility curves for PRA studies. The earthquake ground acceleration, in this approach, is represented by a segment of stationary Gaussian process with a zero mean and a Kanai-Tajimi Spectrum. All possible seismic hazard at a site represented by a hazard curve is also taken into consideration. Furthermore, the limit state of a structure is analytically defined and the corresponding limit state surface is then established. Finally, the fragility curve is generated and the limit state probability is evaluated. In this paper, using a realistic reinforced concrete containment as an example, results of the reliability analysis of the containment subjected to dead load, live load and ground earthquake acceleration are presented and a fragility curve for PRA studies is also constructed

  2. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

  3. Reliability analysis in interdependent smart grid systems

    Science.gov (United States)

    Peng, Hao; Kan, Zhe; Zhao, Dandan; Han, Jianmin; Lu, Jianfeng; Hu, Zhaolong

    2018-06-01

    Complex network theory is a useful way to study many real complex systems. In this paper, a reliability analysis model based on complex network theory is introduced in interdependent smart grid systems. In this paper, we focus on understanding the structure of smart grid systems and studying the underlying network model, their interactions, and relationships and how cascading failures occur in the interdependent smart grid systems. We propose a practical model for interdependent smart grid systems using complex theory. Besides, based on percolation theory, we also study the effect of cascading failures effect and reveal detailed mathematical analysis of failure propagation in such systems. We analyze the reliability of our proposed model caused by random attacks or failures by calculating the size of giant functioning components in interdependent smart grid systems. Our simulation results also show that there exists a threshold for the proportion of faulty nodes, beyond which the smart grid systems collapse. Also we determine the critical values for different system parameters. In this way, the reliability analysis model based on complex network theory can be effectively utilized for anti-attack and protection purposes in interdependent smart grid systems.

  4. Reliability analysis of neutron flux monitoring system for PFBR

    International Nuclear Information System (INIS)

    Rajesh, M.G.; Bhatnagar, P.V.; Das, D.; Pithawa, C.K.; Vinod, Gopika; Rao, V.V.S.S.

    2010-01-01

    The Neutron Flux Monitoring System (NFMS) measures reactor power, rate of change of power and reactivity changes in the core in all states of operation and shutdown. The system consists of instrument channels that are designed and built to have high reliability. All channels are required to have a Mean Time Between Failures (MTBF) of 150000 hours minimum. Failure Mode and Effects Analysis (FMEA) and failure rate estimation of NFMS channels has been carried out. FMEA is carried out in compliance with MIL-STD-338B. Reliability estimation of the channels is done according to MIL-HDBK-217FN2. Paper discusses the methodology followed for FMEA and failure rate estimation of two safety channels and results. (author)

  5. Qualitative analysis in reliability and safety studies

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  6. Sensitivity analysis in optimization and reliability problems

    International Nuclear Information System (INIS)

    Castillo, Enrique; Minguez, Roberto; Castillo, Carmen

    2008-01-01

    The paper starts giving the main results that allow a sensitivity analysis to be performed in a general optimization problem, including sensitivities of the objective function, the primal and the dual variables with respect to data. In particular, general results are given for non-linear programming, and closed formulas for linear programming problems are supplied. Next, the methods are applied to a collection of civil engineering reliability problems, which includes a bridge crane, a retaining wall and a composite breakwater. Finally, the sensitivity analysis formulas are extended to calculus of variations problems and a slope stability problem is used to illustrate the methods

  7. Sensitivity analysis in optimization and reliability problems

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, Enrique [Department of Applied Mathematics and Computational Sciences, University of Cantabria, Avda. Castros s/n., 39005 Santander (Spain)], E-mail: castie@unican.es; Minguez, Roberto [Department of Applied Mathematics, University of Castilla-La Mancha, 13071 Ciudad Real (Spain)], E-mail: roberto.minguez@uclm.es; Castillo, Carmen [Department of Civil Engineering, University of Castilla-La Mancha, 13071 Ciudad Real (Spain)], E-mail: mariacarmen.castillo@uclm.es

    2008-12-15

    The paper starts giving the main results that allow a sensitivity analysis to be performed in a general optimization problem, including sensitivities of the objective function, the primal and the dual variables with respect to data. In particular, general results are given for non-linear programming, and closed formulas for linear programming problems are supplied. Next, the methods are applied to a collection of civil engineering reliability problems, which includes a bridge crane, a retaining wall and a composite breakwater. Finally, the sensitivity analysis formulas are extended to calculus of variations problems and a slope stability problem is used to illustrate the methods.

  8. The quantitative failure of human reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, C.T.

    1995-07-01

    This philosophical treatise argues the merits of Human Reliability Analysis (HRA) in the context of the nuclear power industry. Actually, the author attacks historic and current HRA as having failed in informing policy makers who make decisions based on risk that humans contribute to systems performance. He argues for an HRA based on Bayesian (fact-based) inferential statistics, which advocates a systems analysis process that employs cogent heuristics when using opinion, and tempers itself with a rational debate over the weight given subjective and empirical probabilities.

  9. Infusing Reliability Techniques into Software Safety Analysis

    Science.gov (United States)

    Shi, Ying

    2015-01-01

    Software safety analysis for a large software intensive system is always a challenge. Software safety practitioners need to ensure that software related hazards are completely identified, controlled, and tracked. This paper discusses in detail how to incorporate the traditional reliability techniques into the entire software safety analysis process. In addition, this paper addresses how information can be effectively shared between the various practitioners involved in the software safety analyses. The author has successfully applied the approach to several aerospace applications. Examples are provided to illustrate the key steps of the proposed approach.

  10. Reactivity accident analysis in MTR cores

    International Nuclear Information System (INIS)

    Waldman, R.M.; Vertullo, A.C.

    1987-01-01

    The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

  11. Subset simulation for structural reliability sensitivity analysis

    International Nuclear Information System (INIS)

    Song Shufang; Lu Zhenzhou; Qiao Hongwei

    2009-01-01

    Based on two procedures for efficiently generating conditional samples, i.e. Markov chain Monte Carlo (MCMC) simulation and importance sampling (IS), two reliability sensitivity (RS) algorithms are presented. On the basis of reliability analysis of Subset simulation (Subsim), the RS of the failure probability with respect to the distribution parameter of the basic variable is transformed as a set of RS of conditional failure probabilities with respect to the distribution parameter of the basic variable. By use of the conditional samples generated by MCMC simulation and IS, procedures are established to estimate the RS of the conditional failure probabilities. The formulae of the RS estimator, its variance and its coefficient of variation are derived in detail. The results of the illustrations show high efficiency and high precision of the presented algorithms, and it is suitable for highly nonlinear limit state equation and structural system with single and multiple failure modes

  12. Validity and Reliability of Curl-Up Test on Assessing the Core Endurance for Kindergarten Children in Hong Kong

    OpenAIRE

    Lai, CY; Lee, KY; Lams, MHS; Wu, CF; Peake, R; Flint, SW; Li, WHC; Ho, E

    2017-01-01

    Objective: The purpose of this study was to examine the test-retest reliability and the criterion validity of a curlup\\ud test (CUT) as a measure of core stability, core endurance and dynamic stability in kindergarten children. CUT\\ud performance was also compared to half hold lying test (HHLT) and walking time on course (WTC) among without\\ud obstacle, with low obstacle and high obstacle measures of core stability, core endurance and dynamic stability.\\ud Methods: To estimate reliability, 33...

  13. A taxonomy for human reliability analysis

    International Nuclear Information System (INIS)

    Beattie, J.D.; Iwasa-Madge, K.M.

    1984-01-01

    A human interaction taxonomy (classification scheme) was developed to facilitate human reliability analysis in a probabilistic safety evaluation of a nuclear power plant, being performed at Ontario Hydro. A human interaction occurs, by definition, when operators or maintainers manipulate, or respond to indication from, a plant component or system. The taxonomy aids the fault tree analyst by acting as a heuristic device. It helps define the range and type of human errors to be identified in the construction of fault trees, while keeping the identification by different analysts consistent. It decreases the workload associated with preliminary quantification of the large number of identified interactions by including a category called 'simple interactions'. Fault tree analysts quantify these according to a procedure developed by a team of human reliability specialists. The interactions which do not fit into this category are called 'complex' and are quantified by the human reliability team. The taxonomy is currently being used in fault tree construction in a probabilistic safety evaluation. As far as can be determined at this early stage, the potential benefits of consistency and completeness in identifying human interactions and streamlining the initial quantification are being realized

  14. System reliability analysis with natural language and expert's subjectivity

    International Nuclear Information System (INIS)

    Onisawa, T.

    1996-01-01

    This paper introduces natural language expressions and expert's subjectivity to system reliability analysis. To this end, this paper defines a subjective measure of reliability and presents the method of the system reliability analysis using the measure. The subjective measure of reliability corresponds to natural language expressions of reliability estimation, which is represented by a fuzzy set defined on [0,1]. The presented method deals with the dependence among subsystems and employs parametrized operations of subjective measures of reliability which can reflect expert 's subjectivity towards the analyzed system. The analysis results are also expressed by linguistic terms. Finally this paper gives an example of the system reliability analysis by the presented method

  15. Reliability analysis of containment isolation systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Counts, C.A.

    1984-06-01

    The Pacific Northwest Laboratory (PNL) is reviewing available information on containment systems design, operating experience, and related research as part of a project being conducted by the Division of Systems Integration, US Nuclear Regulatory Commission. The basic objective of this work is to collect and consolidate data relevant to assessing the functional performance of containment isolation systems and to use this data to the extent possible to characterize containment isolation system reliability for selected reference designs. This paper summarizes the results from initial efforts which focused on collection of data from available documents and briefly describes detailed review and analysis efforts which commenced recently. 5 references

  16. Reliability analysis of containment isolation systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Ames, K.R.; Gallucci, R.H.

    1985-06-01

    This report summarizes the results of the Reliability Analysis of Containment Isolation System Project. Work was performed in five basic areas: design review, operating experience review, related research review, generic analysis and plant specific analysis. Licensee Event Reports (LERs) and Integrated Leak Rate Test (ILRT) reports provided the major sources of containment performance information used in this study. Data extracted from LERs were assembled into a computer data base. Qualitative and quantitative information developed for containment performance under normal operating conditions and design basis accidents indicate that there is room for improvement. A rough estimate of overall containment unavailability for relatively small leaks which violate plant technical specifications is 0.3. An estimate of containment unavailability due to large leakage events is in the range of 0.001 to 0.01. These estimates are dependent on several assumptions (particularly on event duration times) which are documented in the report

  17. Statistical hot spot analysis of reactor cores

    International Nuclear Information System (INIS)

    Schaefer, H.

    1974-05-01

    This report is an introduction into statistical hot spot analysis. After the definition of the term 'hot spot' a statistical analysis is outlined. The mathematical method is presented, especially the formula concerning the probability of no hot spots in a reactor core is evaluated. A discussion with the boundary conditions of a statistical hot spot analysis is given (technological limits, nominal situation, uncertainties). The application of the hot spot analysis to the linear power of pellets and the temperature rise in cooling channels is demonstrated with respect to the test zone of KNK II. Basic values, such as probability of no hot spots, hot spot potential, expected hot spot diagram and cumulative distribution function of hot spots, are discussed. It is shown, that the risk of hot channels can be dispersed equally over all subassemblies by an adequate choice of the nominal temperature distribution in the core

  18. Advancing Usability Evaluation through Human Reliability Analysis

    International Nuclear Information System (INIS)

    Ronald L. Boring; David I. Gertman

    2005-01-01

    This paper introduces a novel augmentation to the current heuristic usability evaluation methodology. The SPAR-H human reliability analysis method was developed for categorizing human performance in nuclear power plants. Despite the specialized use of SPAR-H for safety critical scenarios, the method also holds promise for use in commercial off-the-shelf software usability evaluations. The SPAR-H method shares task analysis underpinnings with human-computer interaction, and it can be easily adapted to incorporate usability heuristics as performance shaping factors. By assigning probabilistic modifiers to heuristics, it is possible to arrive at the usability error probability (UEP). This UEP is not a literal probability of error but nonetheless provides a quantitative basis to heuristic evaluation. When combined with a consequence matrix for usability errors, this method affords ready prioritization of usability issues

  19. Integrated Reliability and Risk Analysis System (IRRAS)

    International Nuclear Information System (INIS)

    Russell, K.D.; McKay, M.K.; Sattison, M.B.; Skinner, N.L.; Wood, S.T.; Rasmuson, D.M.

    1992-01-01

    The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the user the ability to create and analyze fault trees and accident sequences using a microcomputer. This program provides functions that range from graphical fault tree construction to cut set generation and quantification. Version 1.0 of the IRRAS program was released in February of 1987. Since that time, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 4.0 and is the subject of this Reference Manual. Version 4.0 of IRRAS provides the same capabilities as Version 1.0 and adds a relational data base facility for managing the data, improved functionality, and improved algorithm performance

  20. Relative and Absolute Reliability of the Professionalism in Physical Therapy Core Values Self-Assessment Tool.

    Science.gov (United States)

    Furgal, Karen E; Norris, Elizabeth S; Young, Sonia N; Wallmann, Harvey W

    2018-01-01

    Development of professional behaviors in Doctor of Physical Therapy (DPT) students is an important part of professional education. The American Physical Therapy Association (APTA) has developed the Professionalism in Physical Therapy Core Values Self-Assessment (PPTCV-SA) tool to increase awareness of personal values in practice. The PPTCV-SA has been used to measure growth in professionalism following a clinical or educational experience. There are few studies reporting psychometric properties of the PPTCV-SA. The purpose of this study was to establish properties of relative reliability (intraclass correlation coefficient, iCC) and absolute reliability (standard error of measurement, SEM; minimal detectable change, MDC) of the PPTCV-SA. in this project, 29 first-year students in a DPT program were administered the PPTCVA-SA on two occasions, 2 weeks apart. Paired t-tests were used to examine stability in PPTCV-SA scores on the two occasions. iCCs were calculated as a measure of relative reliability and for use in the calculation of the absolute reliability measures of SEM and MDC. Results of paired t-tests indicated differences in the subscale scores between times 1 and 2 were non-significant, except for three subscales: Altruism (p=0.01), Excellence (p=0.05), and Social Responsibility (p=0.02). iCCs for test-retest reliability were moderate-to-good for all subscales, with SEMs ranging from 0.30 to 0.62, and MDC95 ranging from 0.83 to 1.71. These results can guide educators and researchers when determining the likelihood of true change in professionalism following a professional development activity.

  1. Model-based human reliability analysis: prospects and requirements

    International Nuclear Information System (INIS)

    Mosleh, A.; Chang, Y.H.

    2004-01-01

    Major limitations of the conventional methods for human reliability analysis (HRA), particularly those developed for operator response analysis in probabilistic safety assessments (PSA) of nuclear power plants, are summarized as a motivation for the need and a basis for developing requirements for the next generation HRA methods. It is argued that a model-based approach that provides explicit cognitive causal links between operator behaviors and directly or indirectly measurable causal factors should be at the core of the advanced methods. An example of such causal model is briefly reviewed, where due to the model complexity and input requirements can only be currently implemented in a dynamic PSA environment. The computer simulation code developed for this purpose is also described briefly, together with current limitations in the models, data, and the computer implementation

  2. Reliability Analysis Techniques for Communication Networks in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lim, T. J.; Jang, S. C.; Kang, H. G.; Kim, M. C.; Eom, H. S.; Lee, H. J.

    2006-09-01

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

  3. Approaches to Demonstrating the Reliability and Validity of Core Diagnostic Criteria for Chronic Pain.

    Science.gov (United States)

    Bruehl, Stephen; Ohrbach, Richard; Sharma, Sonia; Widerstrom-Noga, Eva; Dworkin, Robert H; Fillingim, Roger B; Turk, Dennis C

    2016-09-01

    The Analgesic, Anesthetic, and Addiction Clinical Trial Translations, Innovations, Opportunities, and Networks-American Pain Society Pain Taxonomy (AAPT) is designed to be an evidence-based multidimensional chronic pain classification system that will facilitate more comprehensive and consistent chronic pain diagnoses, and thereby enhance research, clinical communication, and ultimately patient care. Core diagnostic criteria (dimension 1) for individual chronic pain conditions included in the initial version of AAPT will be the focus of subsequent empirical research to evaluate and provide evidence for their reliability and validity. Challenges to validating diagnostic criteria in the absence of clear and identifiable pathophysiological mechanisms are described. Based in part on previous experience regarding the development of evidence-based diagnostic criteria for psychiatric disorders, headache, and specific chronic pain conditions (fibromyalgia, complex regional pain syndrome, temporomandibular disorders, pain associated with spinal cord injuries), several potential approaches for documentation of the reliability and validity of the AAPT diagnostic criteria are summarized. The AAPT is designed to be an evidence-based multidimensional chronic pain classification system. Conceptual and methodological issues related to demonstrating the reliability and validity of the proposed AAPT chronic pain diagnostic criteria are discussed. Copyright © 2016 American Pain Society. Published by Elsevier Inc. All rights reserved.

  4. Stochastic reliability analysis using Fokker Planck equations

    International Nuclear Information System (INIS)

    Hari Prasad, M.; Rami Reddy, G.; Srividya, A.; Verma, A.K.

    2011-01-01

    The Fokker-Planck equation describes the time evolution of the probability density function of the velocity of a particle, and can be generalized to other observables as well. It is also known as the Kolmogorov forward equation (diffusion). Hence, for any process, which evolves with time, the probability density function as a function of time can be represented with Fokker-Planck equation. In stochastic reliability analysis one is more interested in finding out the reliability or failure probability of the components or structures as a function of time rather than instantaneous failure probabilities. In this analysis the variables are represented with random processes instead of random variables. A random processes can be either stationary or non stationary. If the random process is stationary then the failure probability doesn't change with time where as in the case of non stationary processes the failure probability changes with time. In the present paper Fokker Planck equations have been used to find out the probability density function of the non stationary random processes. In this paper a flow chart has been provided which describes step by step process for carrying out stochastic reliability analysis using Fokker-Planck equations. As a first step one has to identify the failure function as a function of random processes. Then one has to solve the Fokker-Planck equation for each random process. In this paper the Fokker-Planck equation has been solved by using Finite difference method. As a result one gets the probability density values of the random process in the sample space as well as time space. Later at each time step appropriate probability distribution has to be identified based on the available probability density values. For checking the better fitness of the data Kolmogorov-Smirnov Goodness of fit test has been performed. In this way one can find out the distribution of the random process at each time step. Once one has the probability distribution

  5. Structural Reliability Analysis of Wind Turbines: A Review

    Directory of Open Access Journals (Sweden)

    Zhiyu Jiang

    2017-12-01

    Full Text Available The paper presents a detailed review of the state-of-the-art research activities on structural reliability analysis of wind turbines between the 1990s and 2017. We describe the reliability methods including the first- and second-order reliability methods and the simulation reliability methods and show the procedure for and application areas of structural reliability analysis of wind turbines. Further, we critically review the various structural reliability studies on rotor blades, bottom-fixed support structures, floating systems and mechanical and electrical components. Finally, future applications of structural reliability methods to wind turbine designs are discussed.

  6. Research review and development trends of human reliability analysis techniques

    International Nuclear Information System (INIS)

    Li Pengcheng; Chen Guohua; Zhang Li; Dai Licao

    2011-01-01

    Human reliability analysis (HRA) methods are reviewed. The theoretical basis of human reliability analysis, human error mechanism, the key elements of HRA methods as well as the existing HRA methods are respectively introduced and assessed. Their shortcomings,the current research hotspot and difficult problems are identified. Finally, it takes a close look at the trends of human reliability analysis methods. (authors)

  7. Reliability Analysis of Tubular Joints in Offshore Structures

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle; Sørensen, John Dalsgaard

    1987-01-01

    Reliability analysis of single tubular joints and offshore platforms with tubular joints is" presented. The failure modes considered are yielding, punching, buckling and fatigue failure. Element reliability as well as systems reliability approaches are used and illustrated by several examples....... Finally, optimal design of tubular.joints with reliability constraints is discussed and illustrated by an example....

  8. Human Reliability Analysis For Computerized Procedures

    International Nuclear Information System (INIS)

    Boring, Ronald L.; Gertman, David I.; Le Blanc, Katya

    2011-01-01

    This paper provides a characterization of human reliability analysis (HRA) issues for computerized procedures in nuclear power plant control rooms. It is beyond the scope of this paper to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper provides a review of HRA as applied to traditional paper-based procedures, followed by a discussion of what specific factors should additionally be considered in HRAs for computerized procedures. Performance shaping factors and failure modes unique to computerized procedures are highlighted. Since there is no definitive guide to HRA for paper-based procedures, this paper also serves to clarify the existing guidance on paper-based procedures before delving into the unique aspects of computerized procedures.

  9. Reliability Analysis of Structural Timber Systems

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard; Hoffmeyer, P.

    2000-01-01

    Structural systems like timber trussed rafters and roof elements made of timber can be expected to have some degree of redundancy and nonlinear/plastic behaviour when the loading consists of for example snow or imposed load. In this paper this system effect is modelled and the statistic...... of variation. In the paper a stochastic model is described for the strength of a single piece of timber taking into account the stochastic variation of the strength and stiffness with length. Also stochastic models for different types of loads are formulated. First, simple representative systems with different...... types of redundancy and non-linearity are considered. The statistical characteristics of the load bearing capacity are determined by reliability analysis. Next, more complex systems are considered modelling the mechanical behaviour of timber roof elements I stressed skin panels made of timber. Using...

  10. Human reliability analysis of dependent events

    International Nuclear Information System (INIS)

    Swain, A.D.; Guttmann, H.E.

    1977-01-01

    In the human reliability analysis in WASH-1400, the continuous variable of degree of interaction among human events was approximated by selecting four points on this continuum to represent the entire continuum. The four points selected were identified as zero coupling (i.e., zero dependence), complete coupling (i.e., complete dependence), and two intermediate points--loose coupling (a moderate level of dependence) and tight coupling (a high level of dependence). The paper expands the WASH-1400 treatment of common mode failure due to the interaction of human activities. Mathematical expressions for the above four levels of dependence are derived for parallel and series systems. The psychological meaning of each level of dependence is illustrated by examples, with probability tree diagrams to illustrate the use of conditional probabilities resulting from the interaction of human actions in nuclear power plant tasks

  11. Reliability analysis of steel-containment strength

    International Nuclear Information System (INIS)

    Greimann, L.G.; Fanous, F.; Wold-Tinsae, A.; Ketalaar, D.; Lin, T.; Bluhm, D.

    1982-06-01

    A best estimate and uncertainty assessment of the resistance of the St. Lucie, Cherokee, Perry, WPPSS and Browns Ferry containment vessels was performed. The Monte Carlo simulation technique and second moment approach were compared as a means of calculating the probability distribution of the containment resistance. A uniform static internal pressure was used and strain ductility was taken as the failure criterion. Approximate methods were developed and calibrated with finite element analysis. Both approximate and finite element analyses were performed on the axisymmetric containment structure. An uncertainty assessment of the containment strength was then performed by the second moment reliability method. Based upon the approximate methods, the cumulative distribution for the resistance of each of the five containments (shell modes only) is presented

  12. Standardizing the practice of human reliability analysis

    International Nuclear Information System (INIS)

    Hallbert, B.P.

    1993-01-01

    The practice of human reliability analysis (HRA) within the nuclear industry varies greatly in terms of posited mechanisms that shape human performance, methods of characterizing and analytically modeling human behavior, and the techniques that are employed to estimate the frequency with which human error occurs. This variation has been a source of contention among HRA practitioners regarding the validity of results obtained from different HRA methods. It has also resulted in attempts to develop standard methods and procedures for conducting HRAs. For many of the same reasons, the practice of HRA has not been standardized or has been standardized only to the extent that individual analysts have developed heuristics and consistent approaches in their practice of HRA. From the standpoint of consumers and regulators, this has resulted in a lack of clear acceptance criteria for the assumptions, modeling, and quantification of human errors in probabilistic risk assessments

  13. A reliability analysis tool for SpaceWire network

    Science.gov (United States)

    Zhou, Qiang; Zhu, Longjiang; Fei, Haidong; Wang, Xingyou

    2017-04-01

    A SpaceWire is a standard for on-board satellite networks as the basis for future data-handling architectures. It is becoming more and more popular in space applications due to its technical advantages, including reliability, low power and fault protection, etc. High reliability is the vital issue for spacecraft. Therefore, it is very important to analyze and improve the reliability performance of the SpaceWire network. This paper deals with the problem of reliability modeling and analysis with SpaceWire network. According to the function division of distributed network, a reliability analysis method based on a task is proposed, the reliability analysis of every task can lead to the system reliability matrix, the reliability result of the network system can be deduced by integrating these entire reliability indexes in the matrix. With the method, we develop a reliability analysis tool for SpaceWire Network based on VC, where the computation schemes for reliability matrix and the multi-path-task reliability are also implemented. By using this tool, we analyze several cases on typical architectures. And the analytic results indicate that redundancy architecture has better reliability performance than basic one. In practical, the dual redundancy scheme has been adopted for some key unit, to improve the reliability index of the system or task. Finally, this reliability analysis tool will has a directive influence on both task division and topology selection in the phase of SpaceWire network system design.

  14. Human Reliability Analysis for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring; David I. Gertman

    2012-06-01

    Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

  15. Optimizing the design and operation of reactor emergency systems using reliability analysis techniques

    International Nuclear Information System (INIS)

    Snaith, E.R.

    1975-01-01

    Following a reactor trip various reactor emergency systems, e.g. essential power supplies, emergency core cooling and boiler feed water arrangements are required to operate with a high degree of reliability. These systems must therefore be critically assessed to confirm their capability of operation and determine their reliability of performance. The use of probability analysis techniques enables the potential operating reliability of the systems to be calculated and this can then be compared with the overall reliability requirements. However, a system reliability analysis does much more than calculate an overall reliability value for the system. It establishes the reliability of all parts of the system and thus identifies the most sensitive areas of unreliability. This indicates the areas where any required improvements should be made and enables the overall systems' designs and modes of operation to be optimized, to meet the system and hence the overall reactor safety criteria. This paper gives specific examples of sensitive areas of unreliability that were identified as a result of a reliability analysis that was carried out on a reactor emergency core cooling system. Details are given of modifications to design and operation that were implemented with a resulting improvement in reliability of various reactor sub-systems. The report concludes that an initial calculation of system reliability should represent only the beginning of continuing process of system assessment. Data on equipment and system performance, particularly in those areas shown to be sensitive in their effect on the overall nuclear power plant reliability, should be collected and processed to give reliability data. These data should then be applied in further probabilistic analyses and the results correlated with the original analysis. This will demonstrate whether the required and the originally predicted system reliability is likely to be achieved, in the light of the actual history to date of

  16. Task Decomposition in Human Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Boring, Ronald Laurids [Idaho National Laboratory; Joe, Jeffrey Clark [Idaho National Laboratory

    2014-06-01

    In the probabilistic safety assessments (PSAs) used in the nuclear industry, human failure events (HFEs) are determined as a subset of hardware failures, namely those hardware failures that could be triggered by human action or inaction. This approach is top-down, starting with hardware faults and deducing human contributions to those faults. Elsewhere, more traditionally human factors driven approaches would tend to look at opportunities for human errors first in a task analysis and then identify which of those errors is risk significant. The intersection of top-down and bottom-up approaches to defining HFEs has not been carefully studied. Ideally, both approaches should arrive at the same set of HFEs. This question remains central as human reliability analysis (HRA) methods are generalized to new domains like oil and gas. The HFEs used in nuclear PSAs tend to be top-down— defined as a subset of the PSA—whereas the HFEs used in petroleum quantitative risk assessments (QRAs) are more likely to be bottom-up—derived from a task analysis conducted by human factors experts. The marriage of these approaches is necessary in order to ensure that HRA methods developed for top-down HFEs are also sufficient for bottom-up applications.

  17. Finite element reliability analysis of fatigue life

    International Nuclear Information System (INIS)

    Harkness, H.H.; Belytschko, T.; Liu, W.K.

    1992-01-01

    Fatigue reliability is addressed by the first-order reliability method combined with a finite element method. Two-dimensional finite element models of components with cracks in mode I are considered with crack growth treated by the Paris law. Probability density functions of the variables affecting fatigue are proposed to reflect a setting where nondestructive evaluation is used, and the Rosenblatt transformation is employed to treat non-Gaussian random variables. Comparisons of the first-order reliability results and Monte Carlo simulations suggest that the accuracy of the first-order reliability method is quite good in this setting. Results show that the upper portion of the initial crack length probability density function is crucial to reliability, which suggests that if nondestructive evaluation is used, the probability of detection curve plays a key role in reliability. (orig.)

  18. Modeling human reliability analysis using MIDAS

    International Nuclear Information System (INIS)

    Boring, R. L.

    2006-01-01

    This paper documents current efforts to infuse human reliability analysis (HRA) into human performance simulation. The Idaho National Laboratory is teamed with NASA Ames Research Center to bridge the SPAR-H HRA method with NASA's Man-machine Integration Design and Analysis System (MIDAS) for use in simulating and modeling the human contribution to risk in nuclear power plant control room operations. It is anticipated that the union of MIDAS and SPAR-H will pave the path for cost-effective, timely, and valid simulated control room operators for studying current and next generation control room configurations. This paper highlights considerations for creating the dynamic HRA framework necessary for simulation, including event dependency and granularity. This paper also highlights how the SPAR-H performance shaping factors can be modeled in MIDAS across static, dynamic, and initiator conditions common to control room scenarios. This paper concludes with a discussion of the relationship of the workload factors currently in MIDAS and the performance shaping factors in SPAR-H. (authors)

  19. Reliability Analysis and Optimal Design of Monolithic Vertical Wall Breakwaters

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard; Burcharth, Hans F.; Christiani, E.

    1994-01-01

    Reliability analysis and reliability-based design of monolithic vertical wall breakwaters are considered. Probabilistic models of the most important failure modes, sliding failure, failure of the foundation and overturning failure are described . Relevant design variables are identified...

  20. Probabilistic risk assessment course documentation. Volume 3. System reliability and analysis techniques, Session A - reliability

    International Nuclear Information System (INIS)

    Lofgren, E.V.

    1985-08-01

    This course in System Reliability and Analysis Techniques focuses on the quantitative estimation of reliability at the systems level. Various methods are reviewed, but the structure provided by the fault tree method is used as the basis for system reliability estimates. The principles of fault tree analysis are briefly reviewed. Contributors to system unreliability and unavailability are reviewed, models are given for quantitative evaluation, and the requirements for both generic and plant-specific data are discussed. Also covered are issues of quantifying component faults that relate to the systems context in which the components are embedded. All reliability terms are carefully defined. 44 figs., 22 tabs

  1. HUMAN RELIABILITY ANALYSIS DENGAN PENDEKATAN COGNITIVE RELIABILITY AND ERROR ANALYSIS METHOD (CREAM

    Directory of Open Access Journals (Sweden)

    Zahirah Alifia Maulida

    2015-01-01

    Full Text Available Kecelakaan kerja pada bidang grinding dan welding menempati urutan tertinggi selama lima tahun terakhir di PT. X. Kecelakaan ini disebabkan oleh human error. Human error terjadi karena pengaruh lingkungan kerja fisik dan non fisik.Penelitian kali menggunakan skenario untuk memprediksi serta mengurangi kemungkinan terjadinya error pada manusia dengan pendekatan CREAM (Cognitive Reliability and Error Analysis Method. CREAM adalah salah satu metode human reliability analysis yang berfungsi untuk mendapatkan nilai Cognitive Failure Probability (CFP yang dapat dilakukan dengan dua cara yaitu basic method dan extended method. Pada basic method hanya akan didapatkan nilai failure probabailty secara umum, sedangkan untuk extended method akan didapatkan CFP untuk setiap task. Hasil penelitian menunjukkan faktor- faktor yang mempengaruhi timbulnya error pada pekerjaan grinding dan welding adalah kecukupan organisasi, kecukupan dari Man Machine Interface (MMI & dukungan operasional, ketersediaan prosedur/ perencanaan, serta kecukupan pelatihan dan pengalaman. Aspek kognitif pada pekerjaan grinding yang memiliki nilai error paling tinggi adalah planning dengan nilai CFP 0.3 dan pada pekerjaan welding yaitu aspek kognitif execution dengan nilai CFP 0.18. Sebagai upaya untuk mengurangi nilai error kognitif pada pekerjaan grinding dan welding rekomendasi yang diberikan adalah memberikan training secara rutin, work instrucstion yang lebih rinci dan memberikan sosialisasi alat. Kata kunci: CREAM (cognitive reliability and error analysis method, HRA (human reliability analysis, cognitive error Abstract The accidents in grinding and welding sectors were the highest cases over the last five years in PT. X and it caused by human error. Human error occurs due to the influence of working environment both physically and non-physically. This study will implement an approaching scenario called CREAM (Cognitive Reliability and Error Analysis Method. CREAM is one of human

  2. STARS software tool for analysis of reliability and safety

    International Nuclear Information System (INIS)

    Poucet, A.; Guagnini, E.

    1989-01-01

    This paper reports on the STARS (Software Tool for the Analysis of Reliability and Safety) project aims at developing an integrated set of Computer Aided Reliability Analysis tools for the various tasks involved in systems safety and reliability analysis including hazard identification, qualitative analysis, logic model construction and evaluation. The expert system technology offers the most promising perspective for developing a Computer Aided Reliability Analysis tool. Combined with graphics and analysis capabilities, it can provide a natural engineering oriented environment for computer assisted reliability and safety modelling and analysis. For hazard identification and fault tree construction, a frame/rule based expert system is used, in which the deductive (goal driven) reasoning and the heuristic, applied during manual fault tree construction, is modelled. Expert system can explain their reasoning so that the analyst can become aware of the why and the how results are being obtained. Hence, the learning aspect involved in manual reliability and safety analysis can be maintained and improved

  3. Space Mission Human Reliability Analysis (HRA) Project

    Science.gov (United States)

    Boyer, Roger

    2014-01-01

    The purpose of the Space Mission Human Reliability Analysis (HRA) Project is to extend current ground-based HRA risk prediction techniques to a long-duration, space-based tool. Ground-based HRA methodology has been shown to be a reasonable tool for short-duration space missions, such as Space Shuttle and lunar fly-bys. However, longer-duration deep-space missions, such as asteroid and Mars missions, will require the crew to be in space for as long as 400 to 900 day missions with periods of extended autonomy and self-sufficiency. Current indications show higher risk due to fatigue, physiological effects due to extended low gravity environments, and others, may impact HRA predictions. For this project, Safety & Mission Assurance (S&MA) will work with Human Health & Performance (HH&P) to establish what is currently used to assess human reliabiilty for human space programs, identify human performance factors that may be sensitive to long duration space flight, collect available historical data, and update current tools to account for performance shaping factors believed to be important to such missions. This effort will also contribute data to the Human Performance Data Repository and influence the Space Human Factors Engineering research risks and gaps (part of the HRP Program). An accurate risk predictor mitigates Loss of Crew (LOC) and Loss of Mission (LOM).The end result will be an updated HRA model that can effectively predict risk on long-duration missions.

  4. Individual Differences in Human Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey C. Joe; Ronald L. Boring

    2014-06-01

    While human reliability analysis (HRA) methods include uncertainty in quantification, the nominal model of human error in HRA typically assumes that operator performance does not vary significantly when they are given the same initiating event, indicators, procedures, and training, and that any differences in operator performance are simply aleatory (i.e., random). While this assumption generally holds true when performing routine actions, variability in operator response has been observed in multiple studies, especially in complex situations that go beyond training and procedures. As such, complexity can lead to differences in operator performance (e.g., operator understanding and decision-making). Furthermore, psychological research has shown that there are a number of known antecedents (i.e., attributable causes) that consistently contribute to observable and systematically measurable (i.e., not random) differences in behavior. This paper reviews examples of individual differences taken from operational experience and the psychological literature. The impact of these differences in human behavior and their implications for HRA are then discussed. We propose that individual differences should not be treated as aleatory, but rather as epistemic. Ultimately, by understanding the sources of individual differences, it is possible to remove some epistemic uncertainty from analyses.

  5. Evaluating abdominal core muscle fatigue: Assessment of the validity and reliability of the prone bridging test.

    Science.gov (United States)

    De Blaiser, C; De Ridder, R; Willems, T; Danneels, L; Vanden Bossche, L; Palmans, T; Roosen, P

    2018-02-01

    The aims of this study were to research the amplitude and median frequency characteristics of selected abdominal, back, and hip muscles of healthy subjects during a prone bridging endurance test, based on surface electromyography (sEMG), (a) to determine if the prone bridging test is a valid field test to measure abdominal muscle fatigue, and (b) to evaluate if the current method of administrating the prone bridging test is reliable. Thirty healthy subjects participated in this experiment. The sEMG activity of seven abdominal, back, and hip muscles was bilaterally measured. Normalized median frequencies were computed from the EMG power spectra. The prone bridging tests were repeated on separate days to evaluate inter and intratester reliability. Significant differences in normalized median frequency slope (NMF slope ) values between several abdominal, back, and hip muscles could be demonstrated. Moderate-to-high correlation coefficients were shown between NMF slope values and endurance time. Multiple backward linear regression revealed that the test endurance time could only be significantly predicted by the NMF slope of the rectus abdominis. Statistical analysis showed excellent reliability (ICC=0.87-0.89). The findings of this study support the validity and reliability of the prone bridging test for evaluating abdominal muscle fatigue. © 2017 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  6. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Gensler, A.; Kühnel, K.; Kuch, S.

    2015-01-01

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

  7. Advances in human reliability analysis in Mexico

    International Nuclear Information System (INIS)

    Nelson, Pamela F.; Gonzalez C, M.; Ruiz S, T.; Guillen M, D.; Contreras V, A.

    2010-10-01

    Human Reliability Analysis (HRA) is a very important part of Probabilistic Risk Analysis (PRA), and constant work is dedicated to improving methods, guidance and data in order to approach realism in the results as well as looking for ways to use these to reduce accident frequency at plants. Further, in order to advance in these areas, several HRA studies are being performed globally. Mexico has participated in the International HRA Empirical study with the objective of -benchmarking- HRA methods by comparing HRA predictions to actual crew performance in a simulator, as well as in the empirical study on a US nuclear power plant currently in progress. The focus of the first study was the development of an understanding of how methods are applied by various analysts, and characterize the methods for their capability to guide the analysts to identify potential human failures, and associated causes and performance shaping factors. The HRA benchmarking study has been performed by using the Halden simulator, 14 European crews, and 15 HRA equipment s (NRC, EPRI, and foreign HRA equipment s using different HRA methods). This effort in Mexico is reflected through the work being performed on updating the Laguna Verde PRA to comply with the ASME PRA standard. In order to be considered an HRA with technical adequacy, that is, be considered as a capability category II, for risk-informed applications, the methodology used for the HRA in the original PRA is not considered sufficiently detailed, and the methodology had to upgraded. The HCR/CBDT/THERP method was chosen, since this is used in many nuclear plants with similar design. The HRA update includes identification and evaluation of human errors that can occur during testing and maintenance, as well as human errors that can occur during an accident using the Emergency Operating Procedures. The review of procedures for maintenance, surveillance and operation is a necessary step in HRA and provides insight into the possible

  8. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  9. Weibull distribution in reliability data analysis in nuclear power plant

    International Nuclear Information System (INIS)

    Ma Yingfei; Zhang Zhijian; Zhang Min; Zheng Gangyang

    2015-01-01

    Reliability is an important issue affecting each stage of the life cycle ranging from birth to death of a product or a system. The reliability engineering includes the equipment failure data processing, quantitative assessment of system reliability and maintenance, etc. Reliability data refers to the variety of data that describe the reliability of system or component during its operation. These data may be in the form of numbers, graphics, symbols, texts and curves. Quantitative reliability assessment is the task of the reliability data analysis. It provides the information related to preventing, detect, and correct the defects of the reliability design. Reliability data analysis under proceed with the various stages of product life cycle and reliability activities. Reliability data of Systems Structures and Components (SSCs) in Nuclear Power Plants is the key factor of probabilistic safety assessment (PSA); reliability centered maintenance and life cycle management. The Weibull distribution is widely used in reliability engineering, failure analysis, industrial engineering to represent manufacturing and delivery times. It is commonly used to model time to fail, time to repair and material strength. In this paper, an improved Weibull distribution is introduced to analyze the reliability data of the SSCs in Nuclear Power Plants. An example is given in the paper to present the result of the new method. The Weibull distribution of mechanical equipment for reliability data fitting ability is very strong in nuclear power plant. It's a widely used mathematical model for reliability analysis. The current commonly used methods are two-parameter and three-parameter Weibull distribution. Through comparison and analysis, the three-parameter Weibull distribution fits the data better. It can reflect the reliability characteristics of the equipment and it is more realistic to the actual situation. (author)

  10. RELIABILITY ANALYSIS OF POWER DISTRIBUTION SYSTEMS

    Directory of Open Access Journals (Sweden)

    Popescu V.S.

    2012-04-01

    Full Text Available Power distribution systems are basic parts of power systems and reliability of these systems at present is a key issue for power engineering development and requires special attention. Operation of distribution systems is accompanied by a number of factors that produce random data a large number of unplanned interruptions. Research has shown that the predominant factors that have a significant influence on the reliability of distribution systems are: weather conditions (39.7%, defects in equipment(25% and unknown random factors (20.1%. In the article is studied the influence of random behavior and are presented estimations of reliability of predominantly rural electrical distribution systems.

  11. Reliability

    OpenAIRE

    Condon, David; Revelle, William

    2017-01-01

    Separating the signal in a test from the irrelevant noise is a challenge for all measurement. Low test reliability limits test validity, attenuates important relationships, and can lead to regression artifacts. Multiple approaches to the assessment and improvement of reliability are discussed. The advantages and disadvantages of several different approaches to reliability are considered. Practical advice on how to assess reliability using open source software is provided.

  12. Reliability in perceptual analysis of voice quality.

    Science.gov (United States)

    Bele, Irene Velsvik

    2005-12-01

    This study focuses on speaking voice quality in male teachers (n = 35) and male actors (n = 36), who represent untrained and trained voice users, because we wanted to investigate normal and supranormal voices. In this study, both substantial and methodologic aspects were considered. It includes a method for perceptual voice evaluation, and a basic issue was rater reliability. A listening group of 10 listeners, 7 experienced speech-language therapists, and 3 speech-language therapist students evaluated the voices by 15 vocal characteristics using VA scales. Two sets of voice signals were investigated: text reading (2 loudness levels) and sustained vowel (3 levels). The results indicated a high interrater reliability for most perceptual characteristics. Connected speech was evaluated more reliably, especially at the normal level, but both types of voice signals were evaluated reliably, although the reliability for connected speech was somewhat higher than for vowels. Experienced listeners tended to be more consistent in their ratings than did the student raters. Some vocal characteristics achieved acceptable reliability even with a smaller panel of listeners. The perceptual characteristics grouped in 4 factors reflected perceptual dimensions.

  13. Reliability Analysis of Fatigue Fracture of Wind Turbine Drivetrain Components

    DEFF Research Database (Denmark)

    Berzonskis, Arvydas; Sørensen, John Dalsgaard

    2016-01-01

    in the volume of the casted ductile iron main shaft, on the reliability of the component. The probabilistic reliability analysis conducted is based on fracture mechanics models. Additionally, the utilization of the probabilistic reliability for operation and maintenance planning and quality control is discussed....

  14. Reliability of the Core Items in the General Social Survey: Estimates from the Three-Wave Panels, 2006–2014

    Directory of Open Access Journals (Sweden)

    Michael Hout

    2016-11-01

    Full Text Available We used standard and multilevel models to assess the reliability of core items in the General Social Survey panel studies spanning 2006 to 2014. Most of the 293 core items scored well on the measure of reliability: 62 items (21 percent had reliability measures greater than 0.85; another 71 (24 percent had reliability measures between 0.70 and 0.85. Objective items, especially facts about demography and religion, were generally more reliable than subjective items. The economic recession of 2007–2009, the slow recovery afterward, and the election of Barack Obama in 2008 altered the social context in ways that may look like unreliability of items. For example, unemployment status, hours worked, and weeks worked have lower reliability than most work-related items, reflecting the consequences of the recession on the facts of peoples lives. Items regarding racial and gender discrimination and racial stereotypes scored as particularly unreliable, accounting for most of the 15 items with reliability coefficients less than 0.40. Our results allow scholars to more easily take measurement reliability into consideration in their own research, while also highlighting the limitations of these approaches.

  15. Component reliability analysis for development of component reliability DB of Korean standard NPPs

    International Nuclear Information System (INIS)

    Choi, S. Y.; Han, S. H.; Kim, S. H.

    2002-01-01

    The reliability data of Korean NPP that reflects the plant specific characteristics is necessary for PSA and Risk Informed Application. We have performed a project to develop the component reliability DB and calculate the component reliability such as failure rate and unavailability. We have collected the component operation data and failure/repair data of Korean standard NPPs. We have analyzed failure data by developing a data analysis method which incorporates the domestic data situation. And then we have compared the reliability results with the generic data for the foreign NPPs

  16. Mathematical Methods in Survival Analysis, Reliability and Quality of Life

    CERN Document Server

    Huber, Catherine; Mesbah, Mounir

    2008-01-01

    Reliability and survival analysis are important applications of stochastic mathematics (probability, statistics and stochastic processes) that are usually covered separately in spite of the similarity of the involved mathematical theory. This title aims to redress this situation: it includes 21 chapters divided into four parts: Survival analysis, Reliability, Quality of life, and Related topics. Many of these chapters were presented at the European Seminar on Mathematical Methods for Survival Analysis, Reliability and Quality of Life in 2006.

  17. Reliability demonstration test planning using bayesian analysis

    International Nuclear Information System (INIS)

    Chandran, Senthil Kumar; Arul, John A.

    2003-01-01

    In Nuclear Power Plants, the reliability of all the safety systems is very critical from the safety viewpoint and it is very essential that the required reliability requirements be met while satisfying the design constraints. From practical experience, it is found that the reliability of complex systems such as Safety Rod Drive Mechanism is of the order of 10 -4 with an uncertainty factor of 10. To demonstrate the reliability of such systems is prohibitive in terms of cost and time as the number of tests needed is very large. The purpose of this paper is to develop a Bayesian reliability demonstrating testing procedure for exponentially distributed failure times with gamma prior distribution on the failure rate which can be easily and effectively used to demonstrate component/subsystem/system reliability conformance to stated requirements. The important questions addressed in this paper are: With zero failures, how long one should perform the tests and how many components are required to conclude with a given degree of confidence, that the component under test, meets the reliability requirement. The procedure is explained with an example. This procedure can also be extended to demonstrate with more number of failures. The approach presented is applicable for deriving test plans for demonstrating component failure rates of nuclear power plants, as the failure data for similar components are becoming available in existing plants elsewhere. The advantages of this procedure are the criterion upon which the procedure is based is simple and pertinent, the fitting of the prior distribution is an integral part of the procedure and is based on the use of information regarding two percentiles of this distribution and finally, the procedure is straightforward and easy to apply in practice. (author)

  18. Reliability analysis of pipe whip impacts

    International Nuclear Information System (INIS)

    Alzbutas, R.; Dundulis, G.; Kulak, R.F.; Marchertas, P.V.

    2003-01-01

    A probabilistic analysis of a group distribution header (GDH) guillotine break and the damage resulting from the failed GDH impacting against a neighbouring wall was carried out for the Ignalita RBMK-1500 reactor. The NEPTUNE software system was used for the deterministic transient analysis of a GDH guillotine break. Many deterministic analyses were performed using different values of the random variables that were specified by ProFES software. All the deterministic results were transferred to the ProFES system, which then performed probabilistic analyses of piping failure and wall damage. The Monte Carlo Simulation (MCS) method was used to study the sensitivity of the response variables and the effect of uncertainties of material properties and geometry parameters to the probability of limit states. The First Order Reliability Method (FORM) was used to study the probability of failure of the impacted-wall and the support-wall. The Response Surface (RS/MCS) method was used in order to express failure probability as function and to investigate the dependence between impact load and failure probability. The results of the probability analyses for a whipping GDH impacting onto an adjacent wall show that: (i) there is a 0.982 probability that after a GDH guillotine break contact between GDH and wall will occur; (ii) there is a probability of 0.013 that the ultimate tensile strength of concrete at the impact location will be reached, and a through-crack may open; (iii) there is a probability of 0.0126 that the ultimate compressive strength of concrete at the GDH support location will be reached, and the concrete may fail; (iv) at the impact location in the adjacent wall, there is a probability of 0.327 that the ultimate tensile strength of the rebars in the first layer will be reached and the rebars will fail; (v) at the GDH support location, there is a probability of 0.11 that the ultimate stress of the rebars in the first layer will be reached and the rebars will fail

  19. Gas Hydrate Investigations Using Pressure Core Analysis: Current Practice

    Science.gov (United States)

    Schultheiss, P.; Holland, M.; Roberts, J.; Druce, M.

    2006-12-01

    Recently there have been a number of major gas hydrate expeditions, both academic and commercially oriented, that have benefited from advances in the practice of pressure coring and pressure core analysis, especially using the HYACINTH pressure coring systems. We report on the now mature process of pressure core acquisition, pressure core handling and pressure core analysis and the results from the analysis of pressure cores, which have revealed important in situ properties along with some remarkable views of gas hydrate morphologies. Pressure coring success rates have improved as the tools have been modified and adapted for use on different drilling platforms. To ensure that pressure cores remain within the hydrate stability zone, tool deployment, recovery and on-deck handling procedures now mitigate against unwanted temperature rises. Core analysis has been integrated into the core transfer protocol and automated nondestructive measurements, including P-wave velocity, gamma density, and X-ray imaging, are routinely made on cores. Pressure cores can be subjected to controlled depressurization experiments while nondestructive measurements are being made, or cores can be stored at in situ conditions for further analysis and subsampling.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

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

  3. Human Reliability Analysis for Design: Using Reliability Methods for Human Factors Issues

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Laurids Boring

    2010-11-01

    This paper reviews the application of human reliability analysis methods to human factors design issues. An application framework is sketched in which aspects of modeling typically found in human reliability analysis are used in a complementary fashion to the existing human factors phases of design and testing. The paper provides best achievable practices for design, testing, and modeling. Such best achievable practices may be used to evaluate and human system interface in the context of design safety certifications.

  4. Human Reliability Analysis for Design: Using Reliability Methods for Human Factors Issues

    International Nuclear Information System (INIS)

    Boring, Ronald Laurids

    2010-01-01

    This paper reviews the application of human reliability analysis methods to human factors design issues. An application framework is sketched in which aspects of modeling typically found in human reliability analysis are used in a complementary fashion to the existing human factors phases of design and testing. The paper provides best achievable practices for design, testing, and modeling. Such best achievable practices may be used to evaluate and human system interface in the context of design safety certifications.

  5. TIGER reliability analysis in the DSN

    Science.gov (United States)

    Gunn, J. M.

    1982-01-01

    The TIGER algorithm, the inputs to the program and the output are described. TIGER is a computer program designed to simulate a system over a period of time to evaluate system reliability and availability. Results can be used in the Deep Space Network for initial spares provisioning and system evaluation.

  6. Reliability analysis of an offshore structure

    DEFF Research Database (Denmark)

    Sorensen, J. D.; Faber, M. H.; Thoft-Christensen, P.

    1992-01-01

    A jacket type offshore structure from the North Sea is considered. The time variant reliability is estimated for failure defined as brittle fracture and crack through the tubular member walls. The stochastic modelling is described. The hot spot stress spectral moments as function of the stochasti...

  7. Reliability analysis of reactor protection systems

    International Nuclear Information System (INIS)

    Alsan, S.

    1976-07-01

    A theoretical mathematical study of reliability is presented and the concepts subsequently defined applied to the study of nuclear reactor safety systems. The theory is applied to investigations of the operational reliability of the Siloe reactor from the point of view of rod drop. A statistical study conducted between 1964 and 1971 demonstrated that most rod drop incidents arose from circumstances associated with experimental equipment (new set-ups). The reliability of the most suitable safety system for some recently developed experimental equipment is discussed. Calculations indicate that if all experimental equipment were equipped with these new systems, only 1.75 rod drop accidents would be expected to occur per year on average. It is suggested that all experimental equipment should be equipped with these new safety systems and tested every 21 days. The reliability of the new safety system currently being studied for the Siloe reactor was also investigated. The following results were obtained: definite failures must be detected immediately as a result of the disturbances produced; the repair time must not exceed a few hours; the equipment must be tested every week. Under such conditions, the rate of accidental rod drops is about 0.013 on average per year. The level of nondefinite failures is less than 10 -6 per hour and the level of nonprotection 1 hour per year. (author)

  8. Bypassing BDD Construction for Reliability Analysis

    DEFF Research Database (Denmark)

    Williams, Poul Frederick; Nikolskaia, Macha; Rauzy, Antoine

    2000-01-01

    In this note, we propose a Boolean Expression Diagram (BED)-based algorithm to compute the minimal p-cuts of boolean reliability models such as fault trees. BEDs make it possible to bypass the Binary Decision Diagram (BDD) construction, which is the main cost of fault tree assessment....

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

    International Nuclear Information System (INIS)

    Li Pengcheng; Chen Guohua; Zhang Li; Xiao Dongsheng

    2010-01-01

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

  10. Mechanical reliability analysis of tubes intended for hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    Nahal, Mourad; Khelif, Rabia [Badji Mokhtar University, Annaba (Algeria)

    2013-02-15

    Reliability analysis constitutes an essential phase in any study concerning reliability. Many industrialists evaluate and improve the reliability of their products during the development cycle - from design to startup (design, manufacture, and exploitation) - to develop their knowledge on cost/reliability ratio and to control sources of failure. In this study, we obtain results for hardness, tensile, and hydrostatic tests carried out on steel tubes for transporting hydrocarbons followed by statistical analysis. Results obtained allow us to conduct a reliability study based on resistance request. Thus, index of reliability is calculated and the importance of the variables related to the tube is presented. Reliability-based assessment of residual stress effects is applied to underground pipelines under a roadway, with and without active corrosion. Residual stress has been found to greatly increase probability of failure, especially in the early stages of pipe lifetime.

  11. Reliability analysis of RC containment structures under combined loads

    International Nuclear Information System (INIS)

    Hwang, H.; Reich, M.; Kagami, S.

    1984-01-01

    This paper discusses a reliability analysis method and load combination design criteria for reinforced concrete containment structures under combined loads. The probability based reliability analysis method is briefly described. For load combination design criteria, derivations of the load factors for accidental pressure due to a design basis accident and safe shutdown earthquake (SSE) for three target limit state probabilities are presented

  12. Analysis of core uncovery time in Kuosheng station blackout transient with MELCOR

    International Nuclear Information System (INIS)

    Wang, S.J.; Chien, C.S.

    1996-01-01

    The MELCOR code, developed by the Sandia National Laboratories, is capable of simulating severe accident phenomena of nuclear power plants. Core uncovery time is an important parameter in the probabilistic risk assessment. However, many MELCOR users do not generate the initial conditions in a station blackout (SBO) transient analysis. Thus, achieving reliable core uncovery time is difficult. The core uncovery time for the Kuosheng nuclear power plant during an SBO transient is analyzed. First, full-power steady-state conditions are generated with the application of a developed self-initialization algorithm. Then the response of the SBO transient up to core uncovery is simulated. The effects of key parameters including the initialization process and the reactor feed pump (RFP) coastdown time on the core uncovery time are analyzed. The initialization process is the most important parameter that affects the core uncovery time. Because SBO transient analysis, the correct initial conditions must be generated to achieve a reliable core uncovery time. The core uncovery time is also sensitive to the RFP coastdown time. A correct time constant is required

  13. Validation study of core analysis methods for full MOX BWR

    International Nuclear Information System (INIS)

    2013-01-01

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO 2 and MOX fuel rods, (3) analysis of isotopic composition data for UO 2 and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  14. Validation study of core analysis methods for full MOX BWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES has been developing a technical database used in reviewing validation of core analysis methods of LWRs in the coming occasions: (1) confirming the core safety parameters of the initial core (one-third MOX core) through a full MOX core in Oma Nuclear Power Plant, which is under the construction, (2) licensing high-burnup MOX cores in the future and (3) reviewing topical reports on core analysis codes for safety design and evaluation. Based on the technical database, JNES will issue a guide of reviewing the core analysis methods used for safety design and evaluation of LWRs. The database will be also used for validation and improving of core analysis codes developed by JNES. JNES has progressed with the projects: (1) improving a Doppler reactivity analysis model in a Monte Carlo calculation code MVP, (2) sensitivity study of nuclear cross section date on reactivity calculation of experimental cores composed of UO{sub 2} and MOX fuel rods, (3) analysis of isotopic composition data for UO{sub 2} and MOX fuels and (4) the guide of reviewing the core analysis codes and others. (author)

  15. Reliability analysis of PLC safety equipment

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

  16. Reliability analysis of PLC safety equipment

    International Nuclear Information System (INIS)

    Yu, J.; Kim, J. Y.

    2006-06-01

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

  17. Multi-Core Processor Memory Contention Benchmark Analysis Case Study

    Science.gov (United States)

    Simon, Tyler; McGalliard, James

    2009-01-01

    Multi-core processors dominate current mainframe, server, and high performance computing (HPC) systems. This paper provides synthetic kernel and natural benchmark results from an HPC system at the NASA Goddard Space Flight Center that illustrate the performance impacts of multi-core (dual- and quad-core) vs. single core processor systems. Analysis of processor design, application source code, and synthetic and natural test results all indicate that multi-core processors can suffer from significant memory subsystem contention compared to similar single-core processors.

  18. Validity and reliability of the Iranian version of the Pediatric Quality of Life Inventory™ 4.0 (PedsQL™ Generic Core Scales in children

    Directory of Open Access Journals (Sweden)

    Amiri Parisa

    2012-01-01

    Full Text Available Abstract Background This study aimed to investigate the reliability and validity of the Iranian version of the Pediatric Quality of Life Inventory™ 4.0 (PedsQL™ 4.0 Generic Core Scales in children. Methods A standard forward and backward translation procedure was used to translate the US English version of the PedsQL™ 4.0 Generic Core Scales for children into the Iranian language (Persian. The Iranian version of the PedsQL™ 4.0 Generic Core Scales was completed by 503 healthy and 22 chronically ill children aged 8-12 years and their parents. The reliability was evaluated using internal consistency. Known-groups discriminant comparisons were made, and exploratory factor analysis (EFA and confirmatory factor analysis (CFA were conducted. Results The internal consistency, as measured by Cronbach's alpha coefficients, exceeded the minimum reliability standard of 0.70. All monotrait-multimethod correlations were higher than multitrait-multimethod correlations. The intraclass correlation coefficients (ICC between the children self-report and parent proxy-reports showed moderate to high agreement. Exploratory factor analysis extracted six factors from the PedsQL™ 4.0 for both self and proxy reports, accounting for 47.9% and 54.8% of total variance, respectively. The results of the confirmatory factor analysis for 6-factor models for both self-report and proxy-report indicated acceptable fit for the proposed models. Regarding health status, as hypothesized from previous studies, healthy children reported significantly higher health-related quality of life than those with chronic illnesses. Conclusions The findings support the initial reliability and validity of the Iranian version of the PedsQL™ 4.0 as a generic instrument to measure health-related quality of life of children in Iran.

  19. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  20. An application of the fault tree analysis for the power system reliability estimation

    International Nuclear Information System (INIS)

    Volkanovski, A.; Cepin, M.; Mavko, B.

    2007-01-01

    The power system is a complex system with its main function to produce, transfer and provide consumers with electrical energy. Combinations of failures of components in the system can result in a failure of power delivery to certain load points and in some cases in a full blackout of power system. The power system reliability directly affects safe and reliable operation of nuclear power plants because the loss of offsite power is a significant contributor to the core damage frequency in probabilistic safety assessments of nuclear power plants. The method, which is based on the integration of the fault tree analysis with the analysis of the power flows in the power system, was developed and implemented for power system reliability assessment. The main contributors to the power system reliability are identified, both quantitatively and qualitatively. (author)

  1. Advances in methods and applications of reliability and safety analysis

    International Nuclear Information System (INIS)

    Fieandt, J.; Hossi, H.; Laakso, K.; Lyytikaeinen, A.; Niemelae, I.; Pulkkinen, U.; Pulli, T.

    1986-01-01

    The know-how of the reliability and safety design and analysis techniques of Vtt has been established over several years in analyzing the reliability in the Finnish nuclear power plants Loviisa and Olkiluoto. This experience has been later on applied and developed to be used in the process industry, conventional power industry, automation and electronics. VTT develops and transfers methods and tools for reliability and safety analysis to the private and public sectors. The technology transfer takes place in joint development projects with potential users. Several computer-aided methods, such as RELVEC for reliability modelling and analysis, have been developed. The tool developed are today used by major Finnish companies in the fields of automation, nuclear power, shipbuilding and electronics. Development of computer-aided and other methods needed in analysis of operating experience, reliability or safety is further going on in a number of research and development projects

  2. Digital Processor Module Reliability Analysis of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lee, Sang Yong; Jung, Jae Hyun; Kim, Jae Ho; Kim, Sung Hun

    2005-01-01

    The system used in plant, military equipment, satellite, etc. consists of many electronic parts as control module, which requires relatively high reliability than other commercial electronic products. Specially, Nuclear power plant related to the radiation safety requires high safety and reliability, so most parts apply to Military-Standard level. Reliability prediction method provides the rational basis of system designs and also provides the safety significance of system operations. Thus various reliability prediction tools have been developed in recent decades, among of them, the MI-HDBK-217 method has been widely used as a powerful tool for the prediction. In this work, It is explained that reliability analysis work for Digital Processor Module (DPM, control module of SMART) is performed by Parts Stress Method based on MIL-HDBK-217F NOTICE2. We are using the Relex 7.6 of Relex software corporation, because reliability analysis process requires enormous part libraries and data for failure rate calculation

  3. Time-dependent reliability sensitivity analysis of motion mechanisms

    International Nuclear Information System (INIS)

    Wei, Pengfei; Song, Jingwen; Lu, Zhenzhou; Yue, Zhufeng

    2016-01-01

    Reliability sensitivity analysis aims at identifying the source of structure/mechanism failure, and quantifying the effects of each random source or their distribution parameters on failure probability or reliability. In this paper, the time-dependent parametric reliability sensitivity (PRS) analysis as well as the global reliability sensitivity (GRS) analysis is introduced for the motion mechanisms. The PRS indices are defined as the partial derivatives of the time-dependent reliability w.r.t. the distribution parameters of each random input variable, and they quantify the effect of the small change of each distribution parameter on the time-dependent reliability. The GRS indices are defined for quantifying the individual, interaction and total contributions of the uncertainty in each random input variable to the time-dependent reliability. The envelope function method combined with the first order approximation of the motion error function is introduced for efficiently estimating the time-dependent PRS and GRS indices. Both the time-dependent PRS and GRS analysis techniques can be especially useful for reliability-based design. This significance of the proposed methods as well as the effectiveness of the envelope function method for estimating the time-dependent PRS and GRS indices are demonstrated with a four-bar mechanism and a car rack-and-pinion steering linkage. - Highlights: • Time-dependent parametric reliability sensitivity analysis is presented. • Time-dependent global reliability sensitivity analysis is presented for mechanisms. • The proposed method is especially useful for enhancing the kinematic reliability. • An envelope method is introduced for efficiently implementing the proposed methods. • The proposed method is demonstrated by two real planar mechanisms.

  4. Reactivity analysis of core distortion effects in the FFTF

    International Nuclear Information System (INIS)

    Knutson, B.J.

    1982-01-01

    An improved technique for evaluating core distortion reactivity effects was developed using reactivity analyses of two core geometry models (R-Z and HEX). This technique is incorporated into a new processor code called CORDIS. The advantages of this technique over existing reactivity models are that is preserves core heterogeneity, provides a control rod insertion effect model, uses row-dependent axial shape functions, and provides a flexible and cost efficient core distortion reactivity analysis method

  5. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.A.N.

    2015-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author).

  6. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.A.N.

    2014-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author)

  7. Application of Network Analysis Method to VHTR core

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl

    2012-01-01

    A Very High Temperature Reactor (VHTR) is currently envisioned as a promising future reactor concept because of its high-efficiency and capability of generating hydrogen. Prismatic Modular Reactor (PMR) is one of the main VHTR concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However their shape could be changed by neutron damage during the reactor operation and the shape change can makes the gaps between the blocks inducing bypass flow. Most of reactor coolant flows through the coolant channel within the fuel block, but some portion of the reactor coolant bypasses to the interstitial gaps. The vertical gap and horizontal gap are called bypass gap and cross gap, respectively. CFD simulation for the full core of VHTR might be possible but it requires vast computational cost and time. Therefore, fast, flexible and reliable code is required to predict the flow distribution corresponding to the various bypass gap distribution. Consequently in this study, the flow network analysis method is applied to analyze the core flow of VHTR. The applied method was validated by comparing with SNU VHTR multiblock experiment. As a result, the calculated results show good agreements with experimental data although computational time and cost of the developed code was very small

  8. Structural reliability analysis applied to pipeline risk analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gardiner, M. [GL Industrial Services, Loughborough (United Kingdom); Mendes, Renato F.; Donato, Guilherme V.P. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

    2009-07-01

    Quantitative Risk Assessment (QRA) of pipelines requires two main components to be provided. These are models of the consequences that follow from some loss of containment incident, and models for the likelihood of such incidents occurring. This paper describes how PETROBRAS have used Structural Reliability Analysis for the second of these, to provide pipeline- and location-specific predictions of failure frequency for a number of pipeline assets. This paper presents an approach to estimating failure rates for liquid and gas pipelines, using Structural Reliability Analysis (SRA) to analyze the credible basic mechanisms of failure such as corrosion and mechanical damage. SRA is a probabilistic limit state method: for a given failure mechanism it quantifies the uncertainty in parameters to mathematical models of the load-resistance state of a structure and then evaluates the probability of load exceeding resistance. SRA can be used to benefit the pipeline risk management process by optimizing in-line inspection schedules, and as part of the design process for new construction in pipeline rights of way that already contain multiple lines. A case study is presented to show how the SRA approach has recently been used on PETROBRAS pipelines and the benefits obtained from it. (author)

  9. Systems reliability analysis for the national ignition facility

    International Nuclear Information System (INIS)

    Majumdar, K.C.; Annese, C.E.; MacIntyre, A.T.; Sicherman, A.

    1996-01-01

    A Reliability, Availability and Maintainability (RAM) analysis was initiated for the National Ignition Facility (NIF). The NIF is an inertial confinement fusion research facility designed to achieve controlled thermonuclear reaction; the preferred site for the NIF is the Lawrence Livermore National Laboratory (LLNL). The NIF RAM analysis has three purposes: (1) to allocate top level reliability and availability goals for the systems, (2) to develop an operability model for optimum maintainability, and (3) to determine the achievability of the allocated goals of the RAM parameters for the NIF systems and the facility operation as a whole. An allocation model assigns the reliability and availability goals for front line and support systems by a top-down approach; reliability analysis uses a bottom-up approach to determine the system reliability and availability from component level to system level

  10. HTR core physics analysis at NRG

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

    2002-01-01

    Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

  11. Effects of core-to-dentin thickness ratio on the biaxial flexural strength, reliability, and fracture mode of bilayered materials of zirconia core (Y-TZP) and veneer indirect composite resins.

    Science.gov (United States)

    Su, Naichuan; Liao, Yunmao; Zhang, Hai; Yue, Li; Lu, Xiaowen; Shen, Jiefei; Wang, Hang

    2017-01-01

    Indirect composite resins (ICR) are promising alternatives as veneering materials for zirconia frameworks. The effects of core-to-dentin thickness ratio (C/Dtr) on the mechanical property of bilayered veneer ICR/yttria-tetragonal zirconia polycrystalline (Y-TZP) core disks have not been previously studied. The purpose of this in vitro study was to assess the effects of C/Dtr on the biaxial flexural strength, reliability, and fracture mode of bilayered veneer ICR/ Y-TZP core disks. A total of 180 bilayered 0.6-mm-thick composite resin disks in core material and C/Dtr of 2:1, 1:1, and 1:2 were tested with either core material placed up or placed down for piston-on-3-ball biaxial flexural strength. The mean biaxial flexural strength, Weibull modulus, and fracture mode were measured to evaluate the variation trend of the biaxial flexural strength, reliability, and fracture mode of the bilayered disks with various C/Dtr. One-way analysis of variance (ANOVA) and chi-square tests were used to evaluate the variation tendency of fracture mode with the C/Dtr or material placed down during testing (α=.05). Light microscopy was used to identify the fracture mode. The mean biaxial flexural strength and reliability improved with the increase in C/Dtr when specimens were tested with the core material either up and down, and depended on the materials that were placed down during testing. The rates of delamination, Hertzian cone cracks, subcritical radial cracks, and number of fracture fragments partially depended on the C/Dtr and the materials that were placed down during testing. The biaxial flexural strength, reliability, and fracture mode in bilayered structures of Y-TZP core and veneer ICR depend on both the C/Dtr and the material that was placed down during testing. Copyright © 2016 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

  12. Interactive reliability analysis project. FY 80 progress report

    International Nuclear Information System (INIS)

    Rasmuson, D.M.; Shepherd, J.C.

    1981-03-01

    This report summarizes the progress to date in the interactive reliability analysis project. Purpose is to develop and demonstrate a reliability and safety technique that can be incorporated early in the design process. Details are illustrated in a simple example of a reactor safety system

  13. Reliability analysis of grid connected small wind turbine power electronics

    International Nuclear Information System (INIS)

    Arifujjaman, Md.; Iqbal, M.T.; Quaicoe, J.E.

    2009-01-01

    Grid connection of small permanent magnet generator (PMG) based wind turbines requires a power conditioning system comprising a bridge rectifier, a dc-dc converter and a grid-tie inverter. This work presents a reliability analysis and an identification of the least reliable component of the power conditioning system of such grid connection arrangements. Reliability of the configuration is analyzed for the worst case scenario of maximum conversion losses at a particular wind speed. The analysis reveals that the reliability of the power conditioning system of such PMG based wind turbines is fairly low and it reduces to 84% of initial value within one year. The investigation is further enhanced by identifying the least reliable component within the power conditioning system and found that the inverter has the dominant effect on the system reliability, while the dc-dc converter has the least significant effect. The reliability analysis demonstrates that a permanent magnet generator based wind energy conversion system is not the best option from the point of view of power conditioning system reliability. The analysis also reveals that new research is required to determine a robust power electronics configuration for small wind turbine conversion systems.

  14. Reliability analysis of wind embedded power generation system for ...

    African Journals Online (AJOL)

    This paper presents a method for Reliability Analysis of wind energy embedded in power generation system for Indian scenario. This is done by evaluating the reliability index, loss of load expectation, for the power generation system with and without integration of wind energy sources in the overall electric power system.

  15. Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

    International Nuclear Information System (INIS)

    Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

    2000-01-01

    This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

  16. Analysis and assessment of water treatment plant reliability

    Directory of Open Access Journals (Sweden)

    Szpak Dawid

    2017-03-01

    Full Text Available The subject of the publication is the analysis and assessment of the reliability of the surface water treatment plant (WTP. In the study the one parameter method of reliability assessment was used. Based on the flow sheet derived from the water company the reliability scheme of the analysed WTP was prepared. On the basis of the daily WTP work report the availability index Kg for the individual elements included in the WTP, was determined. Then, based on the developed reliability scheme showing the interrelationships between elements, the availability index Kg for the whole WTP was determined. The obtained value of the availability index Kg was compared with the criteria values.

  17. An integrated approach to human reliability analysis -- decision analytic dynamic reliability model

    International Nuclear Information System (INIS)

    Holmberg, J.; Hukki, K.; Norros, L.; Pulkkinen, U.; Pyy, P.

    1999-01-01

    The reliability of human operators in process control is sensitive to the context. In many contemporary human reliability analysis (HRA) methods, this is not sufficiently taken into account. The aim of this article is that integration between probabilistic and psychological approaches in human reliability should be attempted. This is achieved first, by adopting such methods that adequately reflect the essential features of the process control activity, and secondly, by carrying out an interactive HRA process. Description of the activity context, probabilistic modeling, and psychological analysis form an iterative interdisciplinary sequence of analysis in which the results of one sub-task maybe input to another. The analysis of the context is carried out first with the help of a common set of conceptual tools. The resulting descriptions of the context promote the probabilistic modeling, through which new results regarding the probabilistic dynamics can be achieved. These can be incorporated in the context descriptions used as reference in the psychological analysis of actual performance. The results also provide new knowledge of the constraints of activity, by providing information of the premises of the operator's actions. Finally, the stochastic marked point process model gives a tool, by which psychological methodology may be interpreted and utilized for reliability analysis

  18. Reliability analysis applied to structural tests

    Science.gov (United States)

    Diamond, P.; Payne, A. O.

    1972-01-01

    The application of reliability theory to predict, from structural fatigue test data, the risk of failure of a structure under service conditions because its load-carrying capability is progressively reduced by the extension of a fatigue crack, is considered. The procedure is applicable to both safe-life and fail-safe structures and, for a prescribed safety level, it will enable an inspection procedure to be planned or, if inspection is not feasible, it will evaluate the life to replacement. The theory has been further developed to cope with the case of structures with initial cracks, such as can occur in modern high-strength materials which are susceptible to the formation of small flaws during the production process. The method has been applied to a structure of high-strength steel and the results are compared with those obtained by the current life estimation procedures. This has shown that the conventional methods can be unconservative in certain cases, depending on the characteristics of the structure and the design operating conditions. The suitability of the probabilistic approach to the interpretation of the results from full-scale fatigue testing of aircraft structures is discussed and the assumptions involved are examined.

  19. Application of DFM in human reliability analysis

    International Nuclear Information System (INIS)

    Yu Shaojie; Zhao Jun; Tong Jiejuan

    2011-01-01

    Combining with ATHEANA, the possible to identify EFCs and UAs using DFM is studied; and then Steam Generator Tube Rupture (SGTR) accident is modeled and solved. Through inductive analysis, 26 Prime Implicants (PIs) are obtained and the meaning of results is interpreted; and one of PIs is similar to the accident scenario of human failure event in one nuclear power plant. Finally, this paper discusses the methods of quantifying PIs, analysis of Error of commission (EOC) and so on. (authors)

  20. Analysis of operating reliability of WWER-1000 unit

    International Nuclear Information System (INIS)

    Bortlik, J.

    1985-01-01

    The nuclear power unit was divided into 33 technological units. Input data for reliability analysis were surveys of operating results obtained from the IAEA information system and certain indexes of the reliability of technological equipment determined using the Bayes formula. The missing reliability data for technological equipment were used from the basic variant. The fault tree of the WWER-1000 unit was determined for the peak event defined as the impossibility of reaching 100%, 75% and 50% of rated power. The period was observed of the nuclear power plant operation with reduced output owing to defect and the respective time needed for a repair of the equipment. The calculation of the availability of the WWER-1000 unit was made for different variant situations. Certain indexes of the operating reliability of the WWER-1000 unit which are the result of a detailed reliability analysis are tabulated for selected variants. (E.S.)

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

    International Nuclear Information System (INIS)

    Frutuoso e Melo, P.F.F.

    1981-01-01

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

  2. Availability analysis of the AP600 passive core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

    1996-12-01

    The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year; this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). It is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3 figs, 3 tabs.

  3. Simulation Approach to Mission Risk and Reliability Analysis, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — It is proposed to develop and demonstrate an integrated total-system risk and reliability analysis approach that is based on dynamic, probabilistic simulation. This...

  4. Reliability analysis of digital I and C systems at KAERI

    International Nuclear Information System (INIS)

    Kim, Man Cheol

    2013-01-01

    This paper provides an overview of the ongoing research activities on a reliability analysis of digital instrumentation and control (I and C) systems of nuclear power plants (NPPs) performed by the Korea Atomic Energy Research Institute (KAERI). The research activities include the development of a new safety-critical software reliability analysis method by integrating the advantages of existing software reliability analysis methods, a fault coverage estimation method based on fault injection experiments, and a new human reliability analysis method for computer-based main control rooms (MCRs) based on human performance data from the APR-1400 full-scope simulator. The research results are expected to be used to address various issues such as the licensing issues related to digital I and C probabilistic safety assessment (PSA) for advanced digital-based NPPs. (author)

  5. PSA applications and piping reliability analysis: where do we stand?

    International Nuclear Information System (INIS)

    Lydell, B.O.Y.

    1997-01-01

    This reviews a recently proposed framework for piping reliability analysis. The framework was developed to promote critical interpretations of operational data on pipe failures, and to support application-specific-parameter estimation

  6. Reliability analysis of digital safety systems at nuclear power plants

    International Nuclear Information System (INIS)

    Sopira Vladimir; Kovacs, Zoltan

    2015-01-01

    Reliability analysis of digital reactor protection systems built on the basis of TELEPERM XS is described, and experience gained by the Slovak RELKO company during the past 20 years in this domain is highlighted. (orig.)

  7. reliability analysis of a two span floor designed according

    African Journals Online (AJOL)

    user

    deterministic approach, considering both ultimate and serviceability limit states. Reliability analysis of the floor ... loading, strength and stiffness parameters, dimensions .... to show that there is a direct relation between the failure probability (Pf) ...

  8. Preliminaries on core image analysis using fault drilling samples; Core image kaiseki kotohajime (danso kussaku core kaisekirei)

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, T; Ito, H [Geological Survey of Japan, Tsukuba (Japan)

    1996-05-01

    This paper introduces examples of image data analysis on fault drilling samples. The paper describes the following matters: core samples used in the analysis are those obtained from wells drilled piercing the Nojima fault which has moved in the Hygoken-Nanbu Earthquake; the CORESCAN system made by DMT Corporation, Germany, used in acquiring the image data consists of a CCD camera, a light source and core rotation mechanism, and a personal computer, its resolution being about 5 pixels/mm in both axial and circumferential directions, and 24-bit full color; with respect to the opening fractures in core samples collected by using a constant azimuth coring, it was possible to derive values of the opening width, inclination angle, and travel from the image data by using a commercially available software for the personal computer; and comparison of this core image with the BHTV record and the hydrophone VSP record (travel and inclination obtained from the BHTV record agree well with those obtained from the core image). 4 refs., 4 figs.

  9. Human reliability analysis in Loviisa probabilistic safety analysis

    International Nuclear Information System (INIS)

    Illman, L.; Isaksson, J.; Makkonen, L.; Vaurio, J.K.; Vuorio, U.

    1986-01-01

    The human reliability analysis in the Loviisa PSA project is carried out for three major groups of errors in human actions: (A) errors made before an initiating event, (B) errors that initiate a transient and (C) errors made during transients. Recovery possibilities are also included in each group. The methods used or planned for each group are described. A simplified THERP approach is used for group A, with emphasis on test and maintenance error recovery aspects and dependencies between redundancies. For group B, task analyses and human factors assessments are made for startup, shutdown and operational transients, with emphasis on potential common cause initiators. For group C, both misdiagnosis and slow decision making are analyzed, as well as errors made in carrying out necessary or backup actions. New or advanced features of the methodology are described

  10. Safety and reliability analysis based on nonprobabilistic methods

    International Nuclear Information System (INIS)

    Kozin, I.O.; Petersen, K.E.

    1996-01-01

    Imprecise probabilities, being developed during the last two decades, offer a considerably more general theory having many advantages which make it very promising for reliability and safety analysis. The objective of the paper is to argue that imprecise probabilities are more appropriate tool for reliability and safety analysis, that they allow to model the behavior of nuclear industry objects more comprehensively and give a possibility to solve some problems unsolved in the framework of conventional approach. Furthermore, some specific examples are given from which we can see the usefulness of the tool for solving some reliability tasks

  11. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    1984-03-01

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  12. Statis Program Analysis for Reliable, Trusted Apps

    Science.gov (United States)

    2017-02-01

    and prevent errors in their Java programs. The Checker Framework includes compiler plug-ins (“checkers”) that find bugs or verify their absence. It...versions of the Java language. 4.8 DATAFLOW FRAMEWORK The dataflow framework enables more accurate analysis of source code. (Despite their similar...names, the dataflow framework is independent of the (Information) Flow Checker of chapter 2.) In Java code, a given operation may be permitted or

  13. Event course analysis of core disruptive accidents

    International Nuclear Information System (INIS)

    Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

    1995-01-01

    The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

  14. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2004-01-01

    Ice cores are the most direct, continuous, and high resolution archive for Late Quaternary paleoclimate reconstruction. Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate migration strategies for New Zealand. (author). 23 refs., 15 figs., 1 tab

  15. Reliability analysis of digital based I and C system

    Energy Technology Data Exchange (ETDEWEB)

    Kang, I. S.; Cho, B. S.; Choi, M. J. [KOPEC, Yongin (Korea, Republic of)

    1999-10-01

    Rapidly, digital technology is being widely applied in replacing analog component installed in existing plant and designing new nuclear power plant for control and monitoring system in Korea as well as in foreign countries. Even though many merits of digital technology, it is being faced with a new problem of reliability assurance. The studies for solving this problem are being performed vigorously in foreign countries. The reliability of KNGR Engineered Safety Features Component Control System (ESF-CCS), digital based I and C system, was analyzed to verify fulfillment of the ALWR EPRI-URD requirement for reliability analysis and eliminate hazards in design applied new technology. The qualitative analysis using FMEA and quantitative analysis using reliability block diagram were performed. The results of analyses are shown in this paper.

  16. Reliability analysis and utilization of PEMs in space application

    Science.gov (United States)

    Jiang, Xiujie; Wang, Zhihua; Sun, Huixian; Chen, Xiaomin; Zhao, Tianlin; Yu, Guanghua; Zhou, Changyi

    2009-11-01

    More and more plastic encapsulated microcircuits (PEMs) are used in space missions to achieve high performance. Since PEMs are designed for use in terrestrial operating conditions, the successful usage of PEMs in space harsh environment is closely related to reliability issues, which should be considered firstly. However, there is no ready-made methodology for PEMs in space applications. This paper discusses the reliability for the usage of PEMs in space. This reliability analysis can be divided into five categories: radiation test, radiation hardness, screening test, reliability calculation and reliability assessment. One case study is also presented to illuminate the details of the process, in which a PEM part is used in a joint space program Double-Star Project between the European Space Agency (ESA) and China. The influence of environmental constrains including radiation, humidity, temperature and mechanics on the PEM part has been considered. Both Double-Star Project satellites are still running well in space now.

  17. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  18. Discrete event simulation versus conventional system reliability analysis approaches

    DEFF Research Database (Denmark)

    Kozine, Igor

    2010-01-01

    Discrete Event Simulation (DES) environments are rapidly developing and appear to be promising tools for building reliability and risk analysis models of safety-critical systems and human operators. If properly developed, they are an alternative to the conventional human reliability analysis models...... and systems analysis methods such as fault and event trees and Bayesian networks. As one part, the paper describes briefly the author’s experience in applying DES models to the analysis of safety-critical systems in different domains. The other part of the paper is devoted to comparing conventional approaches...

  19. Reliability analysis of dispersion nuclear fuel elements

    Science.gov (United States)

    Ding, Shurong; Jiang, Xin; Huo, Yongzhong; Li, Lin an

    2008-03-01

    Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

  20. Reliability analysis of dispersion nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ding Shurong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: dsr1971@163.com; Jiang Xin [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China); Huo Yongzhong [Department of Mechanics and Engineering Science, Fudan University, Shanghai 200433 (China)], E-mail: yzhuo@fudan.edu.cn; Li Linan [Department of Mechanics, Tianjin University, Tianjin 300072 (China)

    2008-03-15

    Taking a dispersion fuel element as a special particle composite, the representative volume element is chosen to act as the research object. The fuel swelling is simulated through temperature increase. The large strain elastoplastic analysis is carried out for the mechanical behaviors using FEM. The results indicate that the fission swelling is simulated successfully; the thickness increments grow linearly with burnup; with increasing of burnup: (1) the first principal stresses at fuel particles change from tensile ones to compression ones, (2) the maximum Mises stresses at the particles transfer from the centers of fuel particles to the location close to the interfaces between the matrix and the particles, their values increase with burnup; the maximum Mises stresses at the matrix exist in the middle location between the two particles near the mid-plane along the length (or width) direction, and the maximum plastic strains are also at the above region.

  1. s-core network decomposition: A generalization of k-core analysis to weighted networks

    Science.gov (United States)

    Eidsaa, Marius; Almaas, Eivind

    2013-12-01

    A broad range of systems spanning biology, technology, and social phenomena may be represented and analyzed as complex networks. Recent studies of such networks using k-core decomposition have uncovered groups of nodes that play important roles. Here, we present s-core analysis, a generalization of k-core (or k-shell) analysis to complex networks where the links have different strengths or weights. We demonstrate the s-core decomposition approach on two random networks (ER and configuration model with scale-free degree distribution) where the link weights are (i) random, (ii) correlated, and (iii) anticorrelated with the node degrees. Finally, we apply the s-core decomposition approach to the protein-interaction network of the yeast Saccharomyces cerevisiae in the context of two gene-expression experiments: oxidative stress in response to cumene hydroperoxide (CHP), and fermentation stress response (FSR). We find that the innermost s-cores are (i) different from innermost k-cores, (ii) different for the two stress conditions CHP and FSR, and (iii) enriched with proteins whose biological functions give insight into how yeast manages these specific stresses.

  2. Durability reliability analysis for corroding concrete structures under uncertainty

    Science.gov (United States)

    Zhang, Hao

    2018-02-01

    This paper presents a durability reliability analysis of reinforced concrete structures subject to the action of marine chloride. The focus is to provide insight into the role of epistemic uncertainties on durability reliability. The corrosion model involves a number of variables whose probabilistic characteristics cannot be fully determined due to the limited availability of supporting data. All sources of uncertainty, both aleatory and epistemic, should be included in the reliability analysis. Two methods are available to formulate the epistemic uncertainty: the imprecise probability-based method and the purely probabilistic method in which the epistemic uncertainties are modeled as random variables. The paper illustrates how the epistemic uncertainties are modeled and propagated in the two methods, and shows how epistemic uncertainties govern the durability reliability.

  3. Reliability analysis of HVDC grid combined with power flow simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yongtao; Langeland, Tore; Solvik, Johan [DNV AS, Hoevik (Norway); Stewart, Emma [DNV KEMA, Camino Ramon, CA (United States)

    2012-07-01

    Based on a DC grid power flow solver and the proposed GEIR, we carried out reliability analysis for a HVDC grid test system proposed by CIGRE working group B4-58, where the failure statistics are collected from literature survey. The proposed methodology is used to evaluate the impact of converter configuration on the overall reliability performance of the HVDC grid, where the symmetrical monopole configuration is compared with the bipole with metallic return wire configuration. The results quantify the improvement on reliability by using the later alternative. (orig.)

  4. Reliability analysis of neutron transport simulation using Monte Carlo method

    International Nuclear Information System (INIS)

    Souza, Bismarck A. de; Borges, Jose C.

    1995-01-01

    This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs

  5. Analysis of the seismic response of a fast reactor core

    International Nuclear Information System (INIS)

    Martelli, A.; Maresca, G.

    1984-01-01

    This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

  6. User's manual of SECOM2: a computer code for seismic system reliability analysis

    International Nuclear Information System (INIS)

    Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Tamura, Kazuo

    2002-03-01

    This report is the user's manual of seismic system reliability analysis code SECOM2 (Seismic Core Melt Frequency Evaluation Code Ver.2) developed at the Japan Atomic Energy Research Institute for systems reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as: Calculation of component failure probabilities based on the response factor method, Extraction of minimal cut sets (MCSs), Calculation of conditional system failure probabilities for given seismic motion levels at the site of an NPP, Calculation of accident sequence frequencies and the core damage frequency (CDF) with use of the seismic hazard curve, Importance analysis using various indicators, Uncertainty analysis, Calculation of the CDF taking into account the effect of the correlations of responses and capacities of components, and Efficient sensitivity analysis by changing parameters on responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about response and capacity of components and seismic hazard curve for the NPP site as inputs. This report presents the models and methods applied in the SECOM2 code and how to use those functions. (author)

  7. Analysis of sodium valve reliability data at CREDO

    International Nuclear Information System (INIS)

    Bott, T.F.; Haas, P.M.

    1979-01-01

    The Centralized Reliability Data Organization (CREDO) has been established at Oak Ridge National Laboratory (ORNL) by the Department of Energy to provide a centralized source of data for reliability/maintainabilty analysis of advanced reactor systems. The current schedule calls for develoment of the data system at a moderate pace, with the first major distribution of data in late FY-1980. Continuous long-term collection of engineering, operating, and event data has been initiated at EBR-II and FFTF

  8. Reliability analysis and updating of deteriorating systems with subset simulation

    DEFF Research Database (Denmark)

    Schneider, Ronald; Thöns, Sebastian; Straub, Daniel

    2017-01-01

    An efficient approach to reliability analysis of deteriorating structural systems is presented, which considers stochastic dependence among element deterioration. Information on a deteriorating structure obtained through inspection or monitoring is included in the reliability assessment through B...... is an efficient and robust sampling-based algorithm suitable for such analyses. The approach is demonstrated in two case studies considering a steel frame structure and a Daniels system subjected to high-cycle fatigue....

  9. Use of COMCAN III in system design and reliability analysis

    International Nuclear Information System (INIS)

    Rasmuson, D.M.; Shepherd, J.C.; Marshall, N.H.; Fitch, L.R.

    1982-03-01

    This manual describes the COMCAN III computer program and its use. COMCAN III is a tool that can be used by the reliability analyst performing a probabilistic risk assessment or by the designer of a system desiring improved performance and efficiency. COMCAN III can be used to determine minimal cut sets of a fault tree, to calculate system reliability characteristics, and to perform qualitative common cause failure analysis

  10. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2009-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 45 refs., 16 figs., 2 tabs.

  11. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2009-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 27 refs., 18 figs., 2 tabs

  12. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.A.N.

    2012-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 28 refs., 20 figs., 1 tab.

  13. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2008-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. Ice core records provide an annual-scale, 'instrumental-quality' baseline of atmospheric temperature and circulation changes back many thousands of years. (author). 27 refs., 18 figs., 2 tabs

  14. Analysis and study on core power capability with margin method

    International Nuclear Information System (INIS)

    Liu Tongxian; Wu Lei; Yu Yingrui; Zhou Jinman

    2015-01-01

    Core power capability analysis focuses on the power distribution control of reactor within the given mode of operation, for the purpose of defining the allowed normal operating space so that Condition Ⅰ maneuvering flexibility is maintained and Condition Ⅱ occurrences are adequately protected by the reactor protection system. For the traditional core power capability analysis methods, such as synthesis method or advanced three dimension method, usually calculate the key safety parameters of the power distribution, and then verify that these parameters meet the design criteria. For PWR with on-line power distribution monitoring system, core power capability analysis calculates the most power level which just meets the design criteria. On the base of 3D FAC method of Westinghouse, the calculation model of core power capability analysis with margin method is introduced to provide reference for engineers. The core power capability analysis of specific burnup of Sanmen NPP is performed with the margin method. The results demonstrate the rationality of the margin method. The calculation model of the margin method not only helps engineers to master the core power capability analysis for AP1000, but also provides reference for engineers for core power capability analysis of other PWR with on-line power distribution monitoring system. (authors)

  15. Reliability analysis framework for computer-assisted medical decision systems

    International Nuclear Information System (INIS)

    Habas, Piotr A.; Zurada, Jacek M.; Elmaghraby, Adel S.; Tourassi, Georgia D.

    2007-01-01

    We present a technique that enhances computer-assisted decision (CAD) systems with the ability to assess the reliability of each individual decision they make. Reliability assessment is achieved by measuring the accuracy of a CAD system with known cases similar to the one in question. The proposed technique analyzes the feature space neighborhood of the query case to dynamically select an input-dependent set of known cases relevant to the query. This set is used to assess the local (query-specific) accuracy of the CAD system. The estimated local accuracy is utilized as a reliability measure of the CAD response to the query case. The underlying hypothesis of the study is that CAD decisions with higher reliability are more accurate. The above hypothesis was tested using a mammographic database of 1337 regions of interest (ROIs) with biopsy-proven ground truth (681 with masses, 656 with normal parenchyma). Three types of decision models, (i) a back-propagation neural network (BPNN), (ii) a generalized regression neural network (GRNN), and (iii) a support vector machine (SVM), were developed to detect masses based on eight morphological features automatically extracted from each ROI. The performance of all decision models was evaluated using the Receiver Operating Characteristic (ROC) analysis. The study showed that the proposed reliability measure is a strong predictor of the CAD system's case-specific accuracy. Specifically, the ROC area index for CAD predictions with high reliability was significantly better than for those with low reliability values. This result was consistent across all decision models investigated in the study. The proposed case-specific reliability analysis technique could be used to alert the CAD user when an opinion that is unlikely to be reliable is offered. The technique can be easily deployed in the clinical environment because it is applicable with a wide range of classifiers regardless of their structure and it requires neither additional

  16. Reliability analysis and initial requirements for FC systems and stacks

    Science.gov (United States)

    Åström, K.; Fontell, E.; Virtanen, S.

    In the year 2000 Wärtsilä Corporation started an R&D program to develop SOFC systems for CHP applications. The program aims to bring to the market highly efficient, clean and cost competitive fuel cell systems with rated power output in the range of 50-250 kW for distributed generation and marine applications. In the program Wärtsilä focuses on system integration and development. System reliability and availability are key issues determining the competitiveness of the SOFC technology. In Wärtsilä, methods have been implemented for analysing the system in respect to reliability and safety as well as for defining reliability requirements for system components. A fault tree representation is used as the basis for reliability prediction analysis. A dynamic simulation technique has been developed to allow for non-static properties in the fault tree logic modelling. Special emphasis has been placed on reliability analysis of the fuel cell stacks in the system. A method for assessing reliability and critical failure predictability requirements for fuel cell stacks in a system consisting of several stacks has been developed. The method is based on a qualitative model of the stack configuration where each stack can be in a functional, partially failed or critically failed state, each of the states having different failure rates and effects on the system behaviour. The main purpose of the method is to understand the effect of stack reliability, critical failure predictability and operating strategy on the system reliability and availability. An example configuration, consisting of 5 × 5 stacks (series of 5 sets of 5 parallel stacks) is analysed in respect to stack reliability requirements as a function of predictability of critical failures and Weibull shape factor of failure rate distributions.

  17. Neutronic analysis of the ford nuclear reactor leu core

    International Nuclear Information System (INIS)

    Raza, S.S.; Hayat, T.

    1989-08-01

    Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs

  18. Buckling analysis of laminated sandwich beam with soft core

    Directory of Open Access Journals (Sweden)

    Anupam Chakrabarti

    Full Text Available Stability analysis of laminated soft core sandwich beam has been studied by a C0 FE model developed by the authors based on higher order zigzag theory (HOZT. The in-plane displacement variation is considered to be cubic for the face sheets and the core, while transverse displacement is quadratic within the core and constant in the faces beyond the core. The proposed model satisfies the condition of stress continuity at the layer interfaces and the zero stress condition at the top and bottom of the beam for transverse shear. Numerical examples are presented to illustrate the accuracy of the present model.

  19. Test-retest reliability of trunk accelerometric gait analysis

    DEFF Research Database (Denmark)

    Henriksen, Marius; Lund, Hans; Moe-Nilssen, R

    2004-01-01

    The purpose of this study was to determine the test-retest reliability of a trunk accelerometric gait analysis in healthy subjects. Accelerations were measured during walking using a triaxial accelerometer mounted on the lumbar spine of the subjects. Six men and 14 women (mean age 35.2; range 18...... a definite potential in clinical gait analysis....

  20. Reliability Analysis of a Two Dissimilar Unit Cold Standby System ...

    African Journals Online (AJOL)

    (2009) using linear first order differential equation evaluated the reliability and availability characteristics of two-dissimilar-unit cold standby system with three mode for which no cost benefit analysis was considered. El-said (1994) contributed on stochastic analysis of a two-dissimilar-unit standby redundant system.

  1. Recent advances in computational structural reliability analysis methods

    Science.gov (United States)

    Thacker, Ben H.; Wu, Y.-T.; Millwater, Harry R.; Torng, Tony Y.; Riha, David S.

    1993-10-01

    The goal of structural reliability analysis is to determine the probability that the structure will adequately perform its intended function when operating under the given environmental conditions. Thus, the notion of reliability admits the possibility of failure. Given the fact that many different modes of failure are usually possible, achievement of this goal is a formidable task, especially for large, complex structural systems. The traditional (deterministic) design methodology attempts to assure reliability by the application of safety factors and conservative assumptions. However, the safety factor approach lacks a quantitative basis in that the level of reliability is never known and usually results in overly conservative designs because of compounding conservatisms. Furthermore, problem parameters that control the reliability are not identified, nor their importance evaluated. A summary of recent advances in computational structural reliability assessment is presented. A significant level of activity in the research and development community was seen recently, much of which was directed towards the prediction of failure probabilities for single mode failures. The focus is to present some early results and demonstrations of advanced reliability methods applied to structural system problems. This includes structures that can fail as a result of multiple component failures (e.g., a redundant truss), or structural components that may fail due to multiple interacting failure modes (e.g., excessive deflection, resonate vibration, or creep rupture). From these results, some observations and recommendations are made with regard to future research needs.

  2. Method for a reliable activation calculation of core components; Methode zur zuverlaessigen Berechnung von Aktivierungen in Kernbauteilen

    Energy Technology Data Exchange (ETDEWEB)

    Mispagel, T.; Phlippen, P.W.; Rose, J. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany)

    2013-07-01

    During nuclear power plant operation components and materials are exposed to the neutron flux from the reactor core and radionuclides are produced. After removal of the fuel elements the radioactivity of these radionuclides in the reactor pressure vessel and the core internals provide more than 99% of the activity of the power plant. For the transport, the interim storage and the final disposal of these radioactive components the radioactive inventories have to be decoded with respect to radiation and nuclides. The declaration of the nuclide and activity inventories requires a reliable calculation of neutron induced activation of reactor components. These activation calculations describe the pile-up of nuclides due to irradiation and due to the decay of nuclides. For an optimum usage of the activity capacities of the repository Konrad it is necessary to have a qualified calculation procedure that keeps the conservatism as low as possible.

  3. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2006-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. (author). 27 refs., 18 figs., 2 tabs

  4. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2005-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. (author). 27 refs., 18 figs., 3 tabs

  5. Stable isotope analysis in ice core paleoclimatology

    International Nuclear Information System (INIS)

    Bertler, N.

    2007-01-01

    Ice cores from New Zealand and the Antarctic margin provide an excellent means of addressing the lack of longer-term climate observations in the Southern Hemisphere with near instrumental quality. Their study helps us to improve our understanding of regional patterns of climate behaviour in Antarctica and its influence on New Zealand, leading to more realistic regional climate models. Such models are needed to sensibly interpret current Antarctic and New Zealand climate variability and for the development of appropriate mitigation strategies for New Zealand. (author). 27 refs., 18 figs., 2 tabs

  6. Core physics calculation and analysis for SNRE

    International Nuclear Information System (INIS)

    Xie Jiachun; Zhao Shouzhi; Jia Baoshan

    2010-01-01

    Five different precise calculation models have been set up for Small Nuclear Rocket Engine (SNRE) core based on MCNP code, and then the effective multiplication constant, drum control worth and power distribution were calculated. The results from different models indicate that the model in which elements are homogeneous could be used in the reactivity calculation, but a detailed description of elements have to be used in the element internal power distribution calculation. The results of physics parameters show that the basic characteristics of SNRE are reasonable. The drum control worth is sufficient. The power distribution is symmetrical and reasonable. All of the parameters can satisfy the design requirement. (authors)

  7. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose Luis

    1996-01-01

    Atucha II is a 745 MW Argentine Power Nuclear Reactor constructed by ENACE SA, Nuclear Argentine Company for Electrical Power Generation and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed

  8. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose L.

    2000-01-01

    Atucha II is a 745 MW Argentine power nuclear reactor constructed by Nuclear Argentine Company for Electric Power Generation S.A. (ENACE S.A.) and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed. (author)

  9. 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.)

  10. The development of a reliable amateur boxing performance analysis template.

    Science.gov (United States)

    Thomson, Edward; Lamb, Kevin; Nicholas, Ceri

    2013-01-01

    The aim of this study was to devise a valid performance analysis system for the assessment of the movement characteristics associated with competitive amateur boxing and assess its reliability using analysts of varying experience of the sport and performance analysis. Key performance indicators to characterise the demands of an amateur contest (offensive, defensive and feinting) were developed and notated using a computerised notational analysis system. Data were subjected to intra- and inter-observer reliability assessment using median sign tests and calculating the proportion of agreement within predetermined limits of error. For all performance indicators, intra-observer reliability revealed non-significant differences between observations (P > 0.05) and high agreement was established (80-100%) regardless of whether exact or the reference value of ±1 was applied. Inter-observer reliability was less impressive for both analysts (amateur boxer and experienced analyst), with the proportion of agreement ranging from 33-100%. Nonetheless, there was no systematic bias between observations for any indicator (P > 0.05), and the proportion of agreement within the reference range (±1) was 100%. A reliable performance analysis template has been developed for the assessment of amateur boxing performance and is available for use by researchers, coaches and athletes to classify and quantify the movement characteristics of amateur boxing.

  11. Reliability analysis of cluster-based ad-hoc networks

    International Nuclear Information System (INIS)

    Cook, Jason L.; Ramirez-Marquez, Jose Emmanuel

    2008-01-01

    The mobile ad-hoc wireless network (MAWN) is a new and emerging network scheme that is being employed in a variety of applications. The MAWN varies from traditional networks because it is a self-forming and dynamic network. The MAWN is free of infrastructure and, as such, only the mobile nodes comprise the network. Pairs of nodes communicate either directly or through other nodes. To do so, each node acts, in turn, as a source, destination, and relay of messages. The virtue of a MAWN is the flexibility this provides; however, the challenge for reliability analyses is also brought about by this unique feature. The variability and volatility of the MAWN configuration makes typical reliability methods (e.g. reliability block diagram) inappropriate because no single structure or configuration represents all manifestations of a MAWN. For this reason, new methods are being developed to analyze the reliability of this new networking technology. New published methods adapt to this feature by treating the configuration probabilistically or by inclusion of embedded mobility models. This paper joins both methods together and expands upon these works by modifying the problem formulation to address the reliability analysis of a cluster-based MAWN. The cluster-based MAWN is deployed in applications with constraints on networking resources such as bandwidth and energy. This paper presents the problem's formulation, a discussion of applicable reliability metrics for the MAWN, and illustration of a Monte Carlo simulation method through the analysis of several example networks

  12. Fatigue Reliability Analysis of a Mono-Tower Platform

    DEFF Research Database (Denmark)

    Kirkegaard, Poul Henning; Sørensen, John Dalsgaard; Brincker, Rune

    1991-01-01

    In this paper, a fatigue reliability analysis of a Mono-tower platform is presented. The failure mode, fatigue failure in the butt welds, is investigated with two different models. The one with the fatigue strength expressed through SN relations, the other with the fatigue strength expressed thro...... of the natural period, damping ratio, current, stress spectrum and parameters describing the fatigue strength. Further, soil damping is shown to be significant for the Mono-tower.......In this paper, a fatigue reliability analysis of a Mono-tower platform is presented. The failure mode, fatigue failure in the butt welds, is investigated with two different models. The one with the fatigue strength expressed through SN relations, the other with the fatigue strength expressed...... through linear-elastic fracture mechanics (LEFM). In determining the cumulative fatigue damage, Palmgren-Miner's rule is applied. Element reliability, as well as systems reliability, is estimated using first-order reliability methods (FORM). The sensitivity of the systems reliability to various parameters...

  13. IEEE guide for the analysis of human reliability

    International Nuclear Information System (INIS)

    Dougherty, E.M. Jr.

    1987-01-01

    The Institute of Electrical and Electronics Engineers (IEEE) working group 7.4 of the Human Factors and Control Facilities Subcommittee of the Nuclear Power Engineering Committee (NPEC) has released its fifth draft of a Guide for General Principles of Human Action Reliability Analysis for Nuclear Power Generating Stations, for approval of NPEC. A guide is the least mandating in the IEEE hierarchy of standards. The purpose is to enhance the performance of an human reliability analysis (HRA) as a part of a probabilistic risk assessment (PRA), to assure reproducible results, and to standardize documentation. The guide does not recommend or even discuss specific techniques, which are too rapidly evolving today. Considerable maturation in the analysis of human reliability in a PRA context has taken place in recent years. The IEEE guide on this subject is an initial step toward bringing HRA out of the research and development arena into the toolbox of standard engineering practices

  14. Reliability Analysis of Wireless Sensor Networks Using Markovian Model

    Directory of Open Access Journals (Sweden)

    Jin Zhu

    2012-01-01

    Full Text Available This paper investigates reliability analysis of wireless sensor networks whose topology is switching among possible connections which are governed by a Markovian chain. We give the quantized relations between network topology, data acquisition rate, nodes' calculation ability, and network reliability. By applying Lyapunov method, sufficient conditions of network reliability are proposed for such topology switching networks with constant or varying data acquisition rate. With the conditions satisfied, the quantity of data transported over wireless network node will not exceed node capacity such that reliability is ensured. Our theoretical work helps to provide a deeper understanding of real-world wireless sensor networks, which may find its application in the fields of network design and topology control.

  15. Reliability of the Emergency Severity Index: Meta-analysis

    Directory of Open Access Journals (Sweden)

    Amir Mirhaghi

    2015-01-01

    Full Text Available Objectives: Although triage systems based on the Emergency Severity Index (ESI have many advantages in terms of simplicity and clarity, previous research has questioned their reliability in practice. Therefore, the aim of this meta-analysis was to determine the reliability of ESI triage scales. Methods: This metaanalysis was performed in March 2014. Electronic research databases were searched and articles conforming to the Guidelines for Reporting Reliability and Agreement Studies were selected. Two researchers independently examined selected abstracts. Data were extracted in the following categories: version of scale (latest/older, participants (adult/paediatric, raters (nurse, physician or expert, method of reliability (intra/inter-rater, reliability statistics (weighted/unweighted kappa and the origin and publication year of the study. The effect size was obtained by the Z-transformation of reliability coefficients. Data were pooled with random-effects models and a meta-regression was performed based on the method of moments estimator. Results: A total of 19 studies from six countries were included in the analysis. The pooled coefficient for the ESI triage scales was substantial at 0.791 (95% confidence interval: 0.787‒0.795. Agreement was higher with the latest and adult versions of the scale and among expert raters, compared to agreement with older and paediatric versions of the scales and with other groups of raters, respectively. Conclusion: ESI triage scales showed an acceptable level of overall reliability. However, ESI scales require more development in order to see full agreement from all rater groups. Further studies concentrating on other aspects of reliability assessment are needed.

  16. Statistical models and methods for reliability and survival analysis

    CERN Document Server

    Couallier, Vincent; Huber-Carol, Catherine; Mesbah, Mounir; Huber -Carol, Catherine; Limnios, Nikolaos; Gerville-Reache, Leo

    2013-01-01

    Statistical Models and Methods for Reliability and Survival Analysis brings together contributions by specialists in statistical theory as they discuss their applications providing up-to-date developments in methods used in survival analysis, statistical goodness of fit, stochastic processes for system reliability, amongst others. Many of these are related to the work of Professor M. Nikulin in statistics over the past 30 years. The authors gather together various contributions with a broad array of techniques and results, divided into three parts - Statistical Models and Methods, Statistical

  17. Tank 241-BY-105 rotary core sampling and analysis plan

    International Nuclear Information System (INIS)

    Sasaki, L.M.

    1995-01-01

    This Sampling and Analysis Plan (SAP) identifies characterization objectives pertaining to sample collection, laboratory analytical evaluation, and reporting requirements for two rotary-mode core samples from tank 241-BY-105 (BY-105)

  18. Bayesian Inference for NASA Probabilistic Risk and Reliability Analysis

    Science.gov (United States)

    Dezfuli, Homayoon; Kelly, Dana; Smith, Curtis; Vedros, Kurt; Galyean, William

    2009-01-01

    This document, Bayesian Inference for NASA Probabilistic Risk and Reliability Analysis, is intended to provide guidelines for the collection and evaluation of risk and reliability-related data. It is aimed at scientists and engineers familiar with risk and reliability methods and provides a hands-on approach to the investigation and application of a variety of risk and reliability data assessment methods, tools, and techniques. This document provides both: A broad perspective on data analysis collection and evaluation issues. A narrow focus on the methods to implement a comprehensive information repository. The topics addressed herein cover the fundamentals of how data and information are to be used in risk and reliability analysis models and their potential role in decision making. Understanding these topics is essential to attaining a risk informed decision making environment that is being sought by NASA requirements and procedures such as 8000.4 (Agency Risk Management Procedural Requirements), NPR 8705.05 (Probabilistic Risk Assessment Procedures for NASA Programs and Projects), and the System Safety requirements of NPR 8715.3 (NASA General Safety Program Requirements).

  19. Reliability analysis of reactor inspection robot(RIROB)

    International Nuclear Information System (INIS)

    Eom, H. S.; Kim, J. H.; Lee, J. C.; Choi, Y. R.; Moon, S. S.

    2002-05-01

    This report describes the method and the result of the reliability analysis of RIROB developed in Korea Atomic Energy Research Institute. There are many classic techniques and models for the reliability analysis. These techniques and models have been used widely and approved in other industries such as aviation and nuclear industry. Though these techniques and models have been approved in real fields they are still insufficient for the complicated systems such RIROB which are composed of computer, networks, electronic parts, mechanical parts, and software. Particularly the application of these analysis techniques to digital and software parts of complicated systems is immature at this time thus expert judgement plays important role in evaluating the reliability of the systems at these days. In this report we proposed a method which combines diverse evidences relevant to the reliability to evaluate the reliability of complicated systems such as RIROB. The proposed method combines diverse evidences and performs inference in formal and in quantitative way by using the benefits of Bayesian Belief Nets (BBN)

  20. Gas Hydrate-Sediment Morphologies Revealed by Pressure Core Analysis

    Science.gov (United States)

    Holland, M.; Schultheiss, P.; Roberts, J.; Druce, M.

    2006-12-01

    Analysis of HYACINTH pressure cores collected on IODP Expedition 311 and NGHP Expedition 1 showed gas hydrate layers, lenses, and veins contained in fine-grained sediments as well as gas hydrate contained in coarse-grained layers. Pressure cores were recovered from sediments on the Cascadia Margin off the North American West Coast and in the Krishna-Godavari Basin in the Western Bay of Bengal in water depths of 800- 1400 meters. Recovered cores were transferred to laboratory chambers without loss of pressure and nondestructive measurements were made at in situ pressures and controlled temperatures. Gamma density, P-wave velocity, and X-ray images showed evidence of grain-displacing and pore-filling gas hydrate in the cores. Data highlights include X-ray images of fine-grained sediment cores showing wispy subvertical veins of gas hydrate and P-wave velocity excursions corresponding to grain-displacing layers and pore-filling layers of gas hydrate. Most cores were subjected to controlled depressurization experiments, where expelled gas was collected, analyzed for composition, and used to calculate gas hydrate saturation within the core. Selected cores were stored under pressure for postcruise analysis and subsampling.

  1. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  2. Core management and performance analysis for PWR

    International Nuclear Information System (INIS)

    Lee, J.B.; Lee, C.K.; Kim, J.S.; Lee, S.K.; Moon, K.S.; Chun, B.J.; Chang, J.W.; Kim, Y.J.

    1981-01-01

    The KINS (KAERI Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactor fuel management, has been developed and benchmarked against the cycles 1 and 2 of the Kori-1 reactor. The critical boron concentration and three-dimensional power distribution at BOL, HZP condition have been calculated and compared with the operating data. A three-dimensional depletion calculation at HFP condition has been performed for cycle 1 with an interval of 1000 MWD/MTU and compared with the operating data. Similar calculation was also performed for cycle 2 and then compared with the design data of the reactor vendor. At the same time, a prediction of in-core detectors reaction rate was made so as to be compared with the operating data. As the result of comparisons, our calculation as well as the justification of the correlations is shown to be in excellent agreement with the operating data within an allowable limit

  3. Reliability-Based Robustness Analysis for a Croatian Sports Hall

    DEFF Research Database (Denmark)

    Čizmar, Dean; Kirkegaard, Poul Henning; Sørensen, John Dalsgaard

    2011-01-01

    This paper presents a probabilistic approach for structural robustness assessment for a timber structure built a few years ago. The robustness analysis is based on a structural reliability based framework for robustness and a simplified mechanical system modelling of a timber truss system....... A complex timber structure with a large number of failure modes is modelled with only a few dominant failure modes. First, a component based robustness analysis is performed based on the reliability indices of the remaining elements after the removal of selected critical elements. The robustness...... is expressed and evaluated by a robustness index. Next, the robustness is assessed using system reliability indices where the probabilistic failure model is modelled by a series system of parallel systems....

  4. Distribution System Reliability Analysis for Smart Grid Applications

    Science.gov (United States)

    Aljohani, Tawfiq Masad

    Reliability of power systems is a key aspect in modern power system planning, design, and operation. The ascendance of the smart grid concept has provided high hopes of developing an intelligent network that is capable of being a self-healing grid, offering the ability to overcome the interruption problems that face the utility and cost it tens of millions in repair and loss. To address its reliability concerns, the power utilities and interested parties have spent extensive amount of time and effort to analyze and study the reliability of the generation and transmission sectors of the power grid. Only recently has attention shifted to be focused on improving the reliability of the distribution network, the connection joint between the power providers and the consumers where most of the electricity problems occur. In this work, we will examine the effect of the smart grid applications in improving the reliability of the power distribution networks. The test system used in conducting this thesis is the IEEE 34 node test feeder, released in 2003 by the Distribution System Analysis Subcommittee of the IEEE Power Engineering Society. The objective is to analyze the feeder for the optimal placement of the automatic switching devices and quantify their proper installation based on the performance of the distribution system. The measures will be the changes in the reliability system indices including SAIDI, SAIFI, and EUE. The goal is to design and simulate the effect of the installation of the Distributed Generators (DGs) on the utility's distribution system and measure the potential improvement of its reliability. The software used in this work is DISREL, which is intelligent power distribution software that is developed by General Reliability Co.

  5. Structural reliability analysis based on the cokriging technique

    International Nuclear Information System (INIS)

    Zhao Wei; Wang Wei; Dai Hongzhe; Xue Guofeng

    2010-01-01

    Approximation methods are widely used in structural reliability analysis because they are simple to create and provide explicit functional relationships between the responses and variables in stead of the implicit limit state function. Recently, the kriging method which is a semi-parameter interpolation technique that can be used for deterministic optimization and structural reliability has gained popularity. However, to fully exploit the kriging method, especially in high-dimensional problems, a large number of sample points should be generated to fill the design space and this can be very expensive and even impractical in practical engineering analysis. Therefore, in this paper, a new method-the cokriging method, which is an extension of kriging, is proposed to calculate the structural reliability. cokriging approximation incorporates secondary information such as the values of the gradients of the function being approximated. This paper explores the use of the cokriging method for structural reliability problems by comparing it with the Kriging method based on some numerical examples. The results indicate that the cokriging procedure described in this work can generate approximation models to improve on the accuracy and efficiency for structural reliability problems and is a viable alternative to the kriging.

  6. Reliability analysis - systematic approach based on limited data

    International Nuclear Information System (INIS)

    Bourne, A.J.

    1975-11-01

    The initial approaches required for reliability analysis are outlined. These approaches highlight the system boundaries, examine the conditions under which the system is required to operate, and define the overall performance requirements. The discussion is illustrated by a simple example of an automatic protective system for a nuclear reactor. It is then shown how the initial approach leads to a method of defining the system, establishing performance parameters of interest and determining the general form of reliability models to be used. The overall system model and the availability of reliability data at the system level are next examined. An iterative process is then described whereby the reliability model and data requirements are systematically refined at progressively lower hierarchic levels of the system. At each stage, the approach is illustrated with examples from the protective system previously described. The main advantages of the approach put forward are the systematic process of analysis, the concentration of assessment effort in the critical areas and the maximum use of limited reliability data. (author)

  7. DATMAN: A reliability data analysis program using Bayesian updating

    International Nuclear Information System (INIS)

    Becker, M.; Feltus, M.A.

    1996-01-01

    Preventive maintenance (PM) techniques focus on the prevention of failures, in particular, system components that are important to plant functions. Reliability-centered maintenance (RCM) improves on the PM techniques by introducing a set of guidelines by which to evaluate the system functions. It also minimizes intrusive maintenance, labor, and equipment downtime without sacrificing system performance when its function is essential for plant safety. Both the PM and RCM approaches require that system reliability data be updated as more component failures and operation time are acquired. Systems reliability and the likelihood of component failures can be calculated by Bayesian statistical methods, which can update these data. The DATMAN computer code has been developed at Penn State to simplify the Bayesian analysis by performing tedious calculations needed for RCM reliability analysis. DATMAN reads data for updating, fits a distribution that best fits the data, and calculates component reliability. DATMAN provides a user-friendly interface menu that allows the user to choose from several common prior and posterior distributions, insert new failure data, and visually select the distribution that matches the data most accurately

  8. Human reliability analysis of Lingao Nuclear Power Station

    International Nuclear Information System (INIS)

    Zhang Li; Huang Shudong; Yang Hong; He Aiwu; Huang Xiangrui; Zheng Tao; Su Shengbing; Xi Haiying

    2001-01-01

    The necessity of human reliability analysis (HRA) of Lingao Nuclear Power Station are analyzed, and the method and operation procedures of HRA is briefed. One of the human factors events (HFE) is analyzed in detail and some questions of HRA are discussed. The authors present the analytical results of 61 HFEs, and make a brief introduction of HRA contribution to Lingao Nuclear Power Station

  9. System Reliability Analysis Capability and Surrogate Model Application in RAVEN

    Energy Technology Data Exchange (ETDEWEB)

    Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Huang, Dongli [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Bei [Idaho National Lab. (INL), Idaho Falls, ID (United States); Adbel-Khalik, Hany S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pascucci, Valerio [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    This report collect the effort performed to improve the reliability analysis capabilities of the RAVEN code and explore new opportunity in the usage of surrogate model by extending the current RAVEN capabilities to multi physics surrogate models and construction of surrogate models for high dimensionality fields.

  10. Factorial validation and reliability analysis of the brain fag syndrome ...

    African Journals Online (AJOL)

    Results: Two valid factors emerged with items 1-3 and items 4, 5 & 7 loading on respectively, making the BFSS a twodimensional (multidimensional) scale which measures 2 aspects of brain fag [labeled burning sensation and crawling sensation respectively]. The reliability analysis yielded a Cronbach Alpha coefficient of ...

  11. A study of operational and testing reliability in software reliability analysis

    International Nuclear Information System (INIS)

    Yang, B.; Xie, M.

    2000-01-01

    Software reliability is an important aspect of any complex equipment today. Software reliability is usually estimated based on reliability models such as nonhomogeneous Poisson process (NHPP) models. Software systems are improving in testing phase, while it normally does not change in operational phase. Depending on whether the reliability is to be predicted for testing phase or operation phase, different measure should be used. In this paper, two different reliability concepts, namely, the operational reliability and the testing reliability, are clarified and studied in detail. These concepts have been mixed up or even misused in some existing literature. Using different reliability concept will lead to different reliability values obtained and it will further lead to different reliability-based decisions made. The difference of the estimated reliabilities is studied and the effect on the optimal release time is investigated

  12. Size analysis of single-core magnetic nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Frank, E-mail: f.ludwig@tu-bs.de [Institut für Elektrische Messtechnik und Grundlagen der Elektrotechnik, TU Braunschweig, Braunschweig (Germany); Balceris, Christoph; Viereck, Thilo [Institut für Elektrische Messtechnik und Grundlagen der Elektrotechnik, TU Braunschweig, Braunschweig (Germany); Posth, Oliver; Steinhoff, Uwe [Physikalisch-Technische Bundesanstalt, Berlin (Germany); Gavilan, Helena; Costo, Rocio [Instituto de Ciencia de Materiales de Madrid, ICMM/CSIC, Madrid (Spain); Zeng, Lunjie; Olsson, Eva [Department of Applied Physics, Chalmers University of Technology, Göteborg (Sweden); Jonasson, Christian; Johansson, Christer [ACREO Swedish ICT AB, Göteborg (Sweden)

    2017-04-01

    Single-core iron-oxide nanoparticles with nominal core diameters of 14 nm and 19 nm were analyzed with a variety of non-magnetic and magnetic analysis techniques, including transmission electron microscopy (TEM), dynamic light scattering (DLS), static magnetization vs. magnetic field (M-H) measurements, ac susceptibility (ACS) and magnetorelaxometry (MRX). From the experimental data, distributions of core and hydrodynamic sizes are derived. Except for TEM where a number-weighted distribution is directly obtained, models have to be applied in order to determine size distributions from the measurand. It was found that the mean core diameters determined from TEM, M-H, ACS and MRX measurements agree well although they are based on different models (Langevin function, Brownian and Néel relaxation times). Especially for the sample with large cores, particle interaction effects come into play, causing agglomerates which were detected in DLS, ACS and MRX measurements. We observed that the number and size of agglomerates can be minimized by sufficiently strong diluting the suspension. - Highlights: • Investigation of size parameters of single-core magnetic nanoparticles with nominal core diameters of 14 nm and 19 nm utilizing different magnetic and non-magnetic methods • Hydrodynamic size determined from ac susceptibility measurements is consistent with the DLS findings • Core size agrees determined from static magnetization curves, MRX and ACS data agrees with results from TEM although the estimation is based on different models (Langevin function, Brownian and Néel relaxation times).

  13. ZERBERUS - the code for reliability analysis of crack containing structures

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.

    1992-04-01

    Brief description of the First- and Second Order Reliability Methods, being the theoretical background of the code, is given. The code structure is described in detail, with special emphasis to the new application fields. The numerical example investigates failure probability of steam generator tubing affected by stress corrosion cracking. The changes necessary to accommodate this analysis within the ZERBERUS code are explained. Analysis results are compared to different Monte Carlo techniques. (orig./HP) [de

  14. Accident Sequence Evaluation Program: Human reliability analysis procedure

    Energy Technology Data Exchange (ETDEWEB)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

  15. Accident Sequence Evaluation Program: Human reliability analysis procedure

    International Nuclear Information System (INIS)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs

  16. Maintenance management of railway infrastructures based on reliability analysis

    International Nuclear Information System (INIS)

    Macchi, Marco; Garetti, Marco; Centrone, Domenico; Fumagalli, Luca; Piero Pavirani, Gian

    2012-01-01

    Railway infrastructure maintenance plays a crucial role for rail transport. It aims at guaranteeing safety of operations and availability of railway tracks and related equipment for traffic regulation. Moreover, it is one major cost for rail transport operations. Thus, the increased competition in traffic market is asking for maintenance improvement, aiming at the reduction of maintenance expenditures while keeping the safety of operations. This issue is addressed by the methodology presented in the paper. The first step of the methodology consists of a family-based approach for the equipment reliability analysis; its purpose is the identification of families of railway items which can be given the same reliability targets. The second step builds the reliability model of the railway system for identifying the most critical items, given a required service level for the transportation system. The two methods have been implemented and tested in practical case studies, in the context of Rete Ferroviaria Italiana, the Italian public limited company for railway transportation.

  17. Reliability Analysis of Free Jet Scour Below Dams

    Directory of Open Access Journals (Sweden)

    Chuanqi Li

    2012-12-01

    Full Text Available Current formulas for calculating scour depth below of a free over fall are mostly deterministic in nature and do not adequately consider the uncertainties of various scouring parameters. A reliability-based assessment of scour, taking into account uncertainties of parameters and coefficients involved, should be performed. This paper studies the reliability of a dam foundation under the threat of scour. A model for calculating the reliability of scour and estimating the probability of failure of the dam foundation subjected to scour is presented. The Maximum Entropy Method is applied to construct the probability density function (PDF of the performance function subject to the moment constraints. Monte Carlo simulation (MCS is applied for uncertainty analysis. An example is considered, and there liability of its scour is computed, the influence of various random variables on the probability failure is analyzed.

  18. Regulatory analysis for the resolution of Generic Issue 115, enhancement of the reliability of the Westinghouse Solid State Protection System

    International Nuclear Information System (INIS)

    Basdekas, D.L.

    1989-05-01

    Generic Issue 115 addresses a concern related to the reliability of the Westinghouse reactor protection system for plants using the Westinghouse Solid State Protection System (SSPS). Several options for improving the reliability of the Westinghouse reactor trip function for these plants and their effect on core damage frequency (CDF) and overall risk were evaluated. This regulatory analysis includes a quantitative assessment of the costs and benefits associated with the various options for enhancing the reliability of the Westinghouse SSPS and provides insights for consideration and industry initiatives. No new regulatory requirements are proposed. 25 refs., 11 tabs

  19. Analysis of the Reliability of the "Alternator- Alternator Belt" System

    Directory of Open Access Journals (Sweden)

    Ivan Mavrin

    2012-10-01

    Full Text Available Before starting and also during the exploitation of va1ioussystems, it is vety imp011ant to know how the system and itsparts will behave during operation regarding breakdowns, i.e.failures. It is possible to predict the service behaviour of a systemby determining the functions of reliability, as well as frequencyand intensity of failures.The paper considers the theoretical basics of the functionsof reliability, frequency and intensity of failures for the twomain approaches. One includes 6 equal intetvals and the other13 unequal intetvals for the concrete case taken from practice.The reliability of the "alternator- alternator belt" system installedin the buses, has been analysed, according to the empiricaldata on failures.The empitical data on failures provide empirical functionsof reliability and frequency and intensity of failures, that arepresented in tables and graphically. The first analysis perfO!med by dividing the mean time between failures into 6 equaltime intervals has given the forms of empirical functions of fa ilurefrequency and intensity that approximately cotTespond totypical functions. By dividing the failure phase into 13 unequalintetvals with two failures in each interval, these functions indicateexplicit transitions from early failure inte1val into the randomfailure interval, i.e. into the ageing intetval. Functions thusobtained are more accurate and represent a better solution forthe given case.In order to estimate reliability of these systems with greateraccuracy, a greater number of failures needs to be analysed.

  20. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.

    1982-07-01

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  1. An exact method for solving logical loops in reliability analysis

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi

    2009-01-01

    This paper presents an exact method for solving logical loops in reliability analysis. The systems that include logical loops are usually described by simultaneous Boolean equations. First, present a basic rule of solving simultaneous Boolean equations. Next, show the analysis procedures for three-component system with external supports. Third, more detailed discussions are given for the establishment of logical loop relation. Finally, take up two typical structures which include more than one logical loop. Their analysis results and corresponding GO-FLOW charts are given. The proposed analytical method is applicable to loop structures that can be described by simultaneous Boolean equations, and it is very useful in evaluating the reliability of complex engineering systems.

  2. Reliability analysis of service water system under earthquake

    International Nuclear Information System (INIS)

    Yu Yu; Qian Xiaoming; Lu Xuefeng; Wang Shengfei; Niu Fenglei

    2013-01-01

    Service water system is one of the important safety systems in nuclear power plant, whose failure probability is always gained by system reliability analysis. The probability of equipment failure under the earthquake is the function of the peak acceleration of earthquake motion, while the occurrence of earthquake is of randomicity, thus the traditional fault tree method in current probability safety assessment is not powerful enough to deal with such case of conditional probability problem. An analysis frame was put forward for system reliability evaluation in seismic condition in this paper, in which Monte Carlo simulation was used to deal with conditional probability problem. Annual failure probability of service water system was calculated, and failure probability of 1.46X10 -4 per year was obtained. The analysis result is in accordance with the data which indicate equipment seismic resistance capability, and the rationality of the model is validated. (authors)

  3. A framework for intelligent reliability centered maintenance analysis

    International Nuclear Information System (INIS)

    Cheng Zhonghua; Jia Xisheng; Gao Ping; Wu Su; Wang Jianzhao

    2008-01-01

    To improve the efficiency of reliability-centered maintenance (RCM) analysis, case-based reasoning (CBR), as a kind of artificial intelligence (AI) technology, was successfully introduced into RCM analysis process, and a framework for intelligent RCM analysis (IRCMA) was studied. The idea for IRCMA is based on the fact that the historical records of RCM analysis on similar items can be referenced and used for the current RCM analysis of a new item. Because many common or similar items may exist in the analyzed equipment, the repeated tasks of RCM analysis can be considerably simplified or avoided by revising the similar cases in conducting RCM analysis. Based on the previous theory studies, an intelligent RCM analysis system (IRCMAS) prototype was developed. This research has focused on the description of the definition, basic principles as well as a framework of IRCMA, and discussion of critical techniques in the IRCMA. Finally, IRCMAS prototype is presented based on a case study

  4. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    Gueray, B.S.

    2001-01-01

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  5. A core ontology for business process analysis

    NARCIS (Netherlands)

    Pedrinaci, C.; Domingue, J.; Alves De Medeiros, A.K.; Bechhofer, S.; Hauswirth, M.; Hoffmann, J.; Koubarakis, M.

    2008-01-01

    Business Process Management (BPM) aims at supporting the whole life-cycle necessary to deploy and maintain business processes in organisations. An important step of the BPM life-cycle is the analysis of the processes deployed in companies. However, the degree of automation currently achieved cannot

  6. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  7. Reliability test and failure analysis of high power LED packages

    International Nuclear Information System (INIS)

    Chen Zhaohui; Zhang Qin; Wang Kai; Luo Xiaobing; Liu Sheng

    2011-01-01

    A new type application specific light emitting diode (LED) package (ASLP) with freeform polycarbonate lens for street lighting is developed, whose manufacturing processes are compatible with a typical LED packaging process. The reliability test methods and failure criterions from different vendors are reviewed and compared. It is found that test methods and failure criterions are quite different. The rapid reliability assessment standards are urgently needed for the LED industry. 85 0 C/85 RH with 700 mA is used to test our LED modules with three other vendors for 1000 h, showing no visible degradation in optical performance for our modules, with two other vendors showing significant degradation. Some failure analysis methods such as C-SAM, Nano X-ray CT and optical microscope are used for LED packages. Some failure mechanisms such as delaminations and cracks are detected in the LED packages after the accelerated reliability testing. The finite element simulation method is helpful for the failure analysis and design of the reliability of the LED packaging. One example is used to show one currently used module in industry is vulnerable and may not easily pass the harsh thermal cycle testing. (semiconductor devices)

  8. HTGR core seismic analysis using an array processor

    International Nuclear Information System (INIS)

    Shatoff, H.; Charman, C.M.

    1983-01-01

    A Floating Point Systems array processor performs nonlinear dynamic analysis of the high-temperature gas-cooled reactor (HTGR) core with significant time and cost savings. The graphite HTGR core consists of approximately 8000 blocks of various shapes which are subject to motion and impact during a seismic event. Two-dimensional computer programs (CRUNCH2D, MCOCO) can perform explicit step-by-step dynamic analyses of up to 600 blocks for time-history motions. However, use of two-dimensional codes was limited by the large cost and run times required. Three-dimensional analysis of the entire core, or even a large part of it, had been considered totally impractical. Because of the needs of the HTGR core seismic program, a Floating Point Systems array processor was used to enhance computer performance of the two-dimensional core seismic computer programs, MCOCO and CRUNCH2D. This effort began by converting the computational algorithms used in the codes to a form which takes maximum advantage of the parallel and pipeline processors offered by the architecture of the Floating Point Systems array processor. The subsequent conversion of the vectorized FORTRAN coding to the array processor required a significant programming effort to make the system work on the General Atomic (GA) UNIVAC 1100/82 host. These efforts were quite rewarding, however, since the cost of running the codes has been reduced approximately 50-fold and the time threefold. The core seismic analysis with large two-dimensional models has now become routine and extension to three-dimensional analysis is feasible. These codes simulate the one-fifth-scale full-array HTGR core model. This paper compares the analysis with the test results for sine-sweep motion

  9. TRACE analysis of Phenix core response to an increase of the core inlet sodium temperature

    Energy Technology Data Exchange (ETDEWEB)

    Chenu, A., E-mail: aurelia.chenu@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Adams, R., E-mail: robert.adams@psi.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Eidgenossische Technische Hochschule, Zurich (Switzerland); Chawla, R., E-mail: rakesh.chawla@epfl.ch [Paul Scherrer Inst., Villigen PSI (Switzerland); Ecole Polytechnique Federale (Switzerland)

    2011-07-01

    This work presents the analysis, using the TRACE code, of the Phenix core response to an inlet sodium temperature increase. The considered experiment was performed in the frame of the Phenix End-Of-Life (EOL) test program of the CEA, prior to the final shutdown of the reactor. It corresponds to a transient following a 40°C increase of the core inlet temperature, which leads to a power decrease of 60%. This work focuses on the first phase of the transient, prior to the reactor scram and pump trip. First, the thermal-hydraulic TRACE model of the core developed for the present analysis is described. The kinetic parameters and feedback coefficients for the point kinetic model were first derived from a 3D static neutronic ERANOS model developed in a former study. The calculated kinetic parameters were then optimized, before use, on the basis of the experimental reactivity in order to minimize the error on the power calculation. The different reactivity feedbacks taken into account include various expansion mechanisms that have been specifically implemented in TRACE for analysis of fast-neutron spectrum systems. The point kinetic model has been used to study the sensitivity of the core response to the different feedback effects. The comparison of the calculated results with the experimental data reveals the need to accurately calculate the reactivity feedback coefficients. This is because the reactor response is very sensitive to small reactivity changes. This study has enabled us to study the sensitivity of the power change to the different reactivity feedbacks and define the most important parameters. As such, it furthers the validation of the FAST code system, which is being used to gain a more in-depth understanding of SFR core behavior during accidental transients. (author)

  10. Reliability analysis of 2400 MWth gas-cooled fast reactor natural circulation decay heat removal system

    International Nuclear Information System (INIS)

    Marques, M.; Bassi, C.; Bentivoglio, F.

    2012-01-01

    In support to a PSA (Probability Safety Assessment) performed at the design level on the 2400 MWth Gas-cooled Fast Reactor, the functional reliability of the decay heat removal system (DHR) working in natural circulation has been estimated in two transient situations corresponding to an 'aggravated' Loss of Flow Accident (LOFA) and a Loss of Coolant Accident (LOCA). The reliability analysis was based on the RMPS methodology. Reliability and global sensitivity analyses use uncertainty propagation by Monte Carlo techniques. The DHR system consists of 1) 3 dedicated DHR loops: the choice of 3 loops (3*100% redundancy) is made in assuming that one could be lost due to the accident initiating event (break for example) and that another one must be supposed unavailable (single failure criterion); 2) a metallic guard containment enclosing the primary system (referred as close containment), not pressurized in normal operation, having a free volume such as the fast primary helium expansion gives an equilibrium pressure of 1.0 MPa, in the first part of the transient (few hours). Each dedicated DHR loop designed to work in forced circulation with blowers or in natural circulation, is composed of 1) a primary loop (cross-duct connected to the core vessel), with a driving height of 10 meters between core and DHX mid-plan; 2) a secondary circuit filled with pressurized water at 1.0 MPa (driving height of 5 meters for natural circulation DHR); 3) a ternary pool, initially at 50 C. degrees, whose volume is determined to handle one day heat extraction (after this time delay, additional measures are foreseen to fill up the pool). The results obtained on the reliability of the DHR system and on the most important input parameters are very different from one scenario to the other showing the necessity for the PSA to perform specific reliability analysis of the passive system for each considered scenario. The analysis shows that the DHR system working in natural circulation is

  11. Sequence comparison and phylogenetic analysis of core gene of ...

    African Journals Online (AJOL)

    Phylogenetic analysis suggests that our sequences are clustered with sequences reported from Japan. This is the first phylogenetic analysis of HCV core gene from Pakistani population. Our sequences and sequences from Japan are grouped into same cluster in the phylogenetic tree. Sequence comparison and ...

  12. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin

    2013-01-01

    State-of-the-art in network coding for wireless, meshed networks typically considers two problems separately. First, the problem of providing reliability for a single session. Second, the problem of opportunistic combination of flows by using minimalistic coding, i.e., by XORing packets from diff...

  13. SBWR core thermal hydraulic analysis during startup

    International Nuclear Information System (INIS)

    Lin, J.H.; Huang, R.L.; Sawyer, C.D.

    1993-01-01

    This paper reports on a thermal hydraulic analysis of the SIMPLIFIED BOILING WATER REACTOR (SBWR) during startup. The potential instability during a SBWR startup has drawn the attention of designers, researchers, and engineers. It has not been a concern for a Boiling Water Reactor (BWR) with forced recirculation; however, for SBWR with natural circulation the concern exists. The concern is about the possibility of a geysering mode oscillation during SBWR startup from a cold temperature and a low system pressure with a low natural circulation flow rate. A thermal hydraulic analysis of the SBWR is performed in simulation of the startup using the TRACG computer code. The temperature, pressure, and reactor power profiles of SBWR during the startup are presented. The results are compared with the data of a natural circulation boiling water reactor, the DODEWAARD plant, in which no instabilities have been observed during many startups. It is shown that a SBWR startup which follows proper procedures, geysering and other modes of oscillations can be avoided

  14. Modelling of magnetostriction of transformer magnetic core for vibration analysis

    Science.gov (United States)

    Marks, Janis; Vitolina, Sandra

    2017-12-01

    Magnetostriction is a phenomenon occurring in transformer core in normal operation mode. Yet in time, it can cause the delamination of magnetic core resulting in higher level of vibrations that are measured on the surface of transformer tank during diagnostic tests. The aim of this paper is to create a model for evaluating elastic deformations in magnetic core that can be used for power transformers with intensive vibrations in order to eliminate magnetostriction as a their cause. Description of the developed model in Matlab and COMSOL software is provided including restrictions concerning geometry and properties of materials, and the results of performed research on magnetic core anisotropy are provided. As a case study modelling of magnetostriction for 5-legged 200 MVA power transformer with the rated voltage of 13.8/137kV is conducted, based on which comparative analysis of vibration levels and elastic deformations is performed.

  15. Modelling of magnetostriction of transformer magnetic core for vibration analysis

    Directory of Open Access Journals (Sweden)

    Marks Janis

    2017-12-01

    Full Text Available Magnetostriction is a phenomenon occurring in transformer core in normal operation mode. Yet in time, it can cause the delamination of magnetic core resulting in higher level of vibrations that are measured on the surface of transformer tank during diagnostic tests. The aim of this paper is to create a model for evaluating elastic deformations in magnetic core that can be used for power transformers with intensive vibrations in order to eliminate magnetostriction as a their cause. Description of the developed model in Matlab and COMSOL software is provided including restrictions concerning geometry and properties of materials, and the results of performed research on magnetic core anisotropy are provided. As a case study modelling of magnetostriction for 5-legged 200 MVA power transformer with the rated voltage of 13.8/137kV is conducted, based on which comparative analysis of vibration levels and elastic deformations is performed.

  16. Reliability analysis of maintenance operations for railway tracks

    International Nuclear Information System (INIS)

    Rhayma, N.; Bressolette, Ph.; Breul, P.; Fogli, M.; Saussine, G.

    2013-01-01

    Railway engineering is confronted with problems due to degradation of the railway network that requires important and costly maintenance work. However, because of the lack of knowledge on the geometrical and mechanical parameters of the track, it is difficult to optimize the maintenance management. In this context, this paper presents a new methodology to analyze the behavior of railway tracks. It combines new diagnostic devices which permit to obtain an important amount of data and thus to make statistics on the geometric and mechanical parameters and a non-intrusive stochastic approach which can be coupled with any mechanical model. Numerical results show the possibilities of this methodology for reliability analysis of different maintenance operations. In the future this approach will give important informations to railway managers to optimize maintenance operations using a reliability analysis

  17. Prediction of Hydrophobic Cores of Proteins Using Wavelet Analysis.

    Science.gov (United States)

    Hirakawa; Kuhara

    1997-01-01

    Information concerning the secondary structures, flexibility, epitope and hydrophobic regions of amino acid sequences can be extracted by assigning physicochemical indices to each amino acid residue, and information on structure can be derived using the sliding window averaging technique, which is in wide use for smoothing out raw functions. Wavelet analysis has shown great potential and applicability in many fields, such as astronomy, radar, earthquake prediction, and signal or image processing. This approach is efficient for removing noise from various functions. Here we employed wavelet analysis to smooth out a plot assigned to a hydrophobicity index for amino acid sequences. We then used the resulting function to predict hydrophobic cores in globular proteins. We calculated the prediction accuracy for the hydrophobic cores of 88 representative set of proteins. Use of wavelet analysis made feasible the prediction of hydrophobic cores at 6.13% greater accuracy than the sliding window averaging technique.

  18. Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

    International Nuclear Information System (INIS)

    Khan, L.A.; Nabbi, R.

    1987-02-01

    In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

  19. Reliability analysis based on the losses from failures.

    Science.gov (United States)

    Todinov, M T

    2006-04-01

    The conventional reliability analysis is based on the premise that increasing the reliability of a system will decrease the losses from failures. On the basis of counterexamples, it is demonstrated that this is valid only if all failures are associated with the same losses. In case of failures associated with different losses, a system with larger reliability is not necessarily characterized by smaller losses from failures. Consequently, a theoretical framework and models are proposed for a reliability analysis, linking reliability and the losses from failures. Equations related to the distributions of the potential losses from failure have been derived. It is argued that the classical risk equation only estimates the average value of the potential losses from failure and does not provide insight into the variability associated with the potential losses. Equations have also been derived for determining the potential and the expected losses from failures for nonrepairable and repairable systems with components arranged in series, with arbitrary life distributions. The equations are also valid for systems/components with multiple mutually exclusive failure modes. The expected losses given failure is a linear combination of the expected losses from failure associated with the separate failure modes scaled by the conditional probabilities with which the failure modes initiate failure. On this basis, an efficient method for simplifying complex reliability block diagrams has been developed. Branches of components arranged in series whose failures are mutually exclusive can be reduced to single components with equivalent hazard rate, downtime, and expected costs associated with intervention and repair. A model for estimating the expected losses from early-life failures has also been developed. For a specified time interval, the expected losses from early-life failures are a sum of the products of the expected number of failures in the specified time intervals covering the

  20. Solid Rocket Booster Large Main and Drogue Parachute Reliability Analysis

    Science.gov (United States)

    Clifford, Courtenay B.; Hengel, John E.

    2009-01-01

    The parachutes on the Space Transportation System (STS) Solid Rocket Booster (SRB) are the means for decelerating the SRB and allowing it to impact the water at a nominal vertical velocity of 75 feet per second. Each SRB has one pilot, one drogue, and three main parachutes. About four minutes after SRB separation, the SRB nose cap is jettisoned, deploying the pilot parachute. The pilot chute then deploys the drogue parachute. The drogue chute provides initial deceleration and proper SRB orientation prior to frustum separation. At frustum separation, the drogue pulls the frustum from the SRB and allows the main parachutes that are mounted in the frustum to unpack and inflate. These chutes are retrieved, inspected, cleaned, repaired as needed, and returned to the flight inventory and reused. Over the course of the Shuttle Program, several improvements have been introduced to the SRB main parachutes. A major change was the replacement of the small (115 ft. diameter) main parachutes with the larger (136 ft. diameter) main parachutes. Other modifications were made to the main parachutes, main parachute support structure, and SRB frustum to eliminate failure mechanisms, improve damage tolerance, and improve deployment and inflation characteristics. This reliability analysis is limited to the examination of the SRB Large Main Parachute (LMP) and drogue parachute failure history to assess the reliability of these chutes. From the inventory analysis, 68 Large Main Parachutes were used in 651 deployments, and 7 chute failures occurred in the 651 deployments. Logistic regression was used to analyze the LMP failure history, and it showed that reliability growth has occurred over the period of use resulting in a current chute reliability of R = .9983. This result was then used to determine the reliability of the 3 LMPs on the SRB, when all must function. There are 29 drogue parachutes that were used in 244 deployments, and no in-flight failures have occurred. Since there are no

  1. Inclusion of task dependence in human reliability analysis

    International Nuclear Information System (INIS)

    Su, Xiaoyan; Mahadevan, Sankaran; Xu, Peida; Deng, Yong

    2014-01-01

    Dependence assessment among human errors in human reliability analysis (HRA) is an important issue, which includes the evaluation of the dependence among human tasks and the effect of the dependence on the final human error probability (HEP). This paper represents a computational model to handle dependence in human reliability analysis. The aim of the study is to automatically provide conclusions on the overall degree of dependence and calculate the conditional human error probability (CHEP) once the judgments of the input factors are given. The dependence influencing factors are first identified by the experts and the priorities of these factors are also taken into consideration. Anchors and qualitative labels are provided as guidance for the HRA analyst's judgment of the input factors. The overall degree of dependence between human failure events is calculated based on the input values and the weights of the input factors. Finally, the CHEP is obtained according to a computing formula derived from the technique for human error rate prediction (THERP) method. The proposed method is able to quantify the subjective judgment from the experts and improve the transparency in the HEP evaluation process. Two examples are illustrated to show the effectiveness and the flexibility of the proposed method. - Highlights: • We propose a computational model to handle dependence in human reliability analysis. • The priorities of the dependence influencing factors are taken into consideration. • The overall dependence degree is determined by input judgments and the weights of factors. • The CHEP is obtained according to a computing formula derived from THERP

  2. High-Reliable PLC RTOS Development and RPS Structure Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, H. S.; Song, D. Y.; Sohn, D. S.; Kim, J. H. [Enersys Co., Daejeon (Korea, Republic of)

    2008-04-15

    One of the KNICS objectives is to develop a platform for Nuclear Power Plant(NPP) I and C(Instrumentation and Control) system, especially plant protection system. The developed platform is POSAFE-Q and this work supports the development of POSAFE-Q with the development of high-reliable real-time operating system(RTOS) and programmable logic device(PLD) software. Another KNICS objective is to develop safety I and C systems, such as Reactor Protection System(RPS) and Engineered Safety Feature-Component Control System(ESF-CCS). This work plays an important role in the structure analysis for RPS. Validation and verification(V and V) of the safety critical software is an essential work to make digital plant protection system highly reliable and safe. Generally, the reliability and safety of software based system can be improved by strict quality assurance framework including the software development itself. In other words, through V and V, the reliability and safety of a system can be improved and the development activities like software requirement specification, software design specification, component tests, integration tests, and system tests shall be appropriately documented for V and V.

  3. High-Reliable PLC RTOS Development and RPS Structure Analysis

    International Nuclear Information System (INIS)

    Sohn, H. S.; Song, D. Y.; Sohn, D. S.; Kim, J. H.

    2008-04-01

    One of the KNICS objectives is to develop a platform for Nuclear Power Plant(NPP) I and C(Instrumentation and Control) system, especially plant protection system. The developed platform is POSAFE-Q and this work supports the development of POSAFE-Q with the development of high-reliable real-time operating system(RTOS) and programmable logic device(PLD) software. Another KNICS objective is to develop safety I and C systems, such as Reactor Protection System(RPS) and Engineered Safety Feature-Component Control System(ESF-CCS). This work plays an important role in the structure analysis for RPS. Validation and verification(V and V) of the safety critical software is an essential work to make digital plant protection system highly reliable and safe. Generally, the reliability and safety of software based system can be improved by strict quality assurance framework including the software development itself. In other words, through V and V, the reliability and safety of a system can be improved and the development activities like software requirement specification, software design specification, component tests, integration tests, and system tests shall be appropriately documented for V and V.

  4. Reliability analysis with linguistic data: An evidential network approach

    International Nuclear Information System (INIS)

    Zhang, Xiaoge; Mahadevan, Sankaran; Deng, Xinyang

    2017-01-01

    In practical applications of reliability assessment of a system in-service, information about the condition of a system and its components is often available in text form, e.g., inspection reports. Estimation of the system reliability from such text-based records becomes a challenging problem. In this paper, we propose a four-step framework to deal with this problem. In the first step, we construct an evidential network with the consideration of available knowledge and data. Secondly, we train a Naive Bayes text classification algorithm based on the past records. By using the trained Naive Bayes algorithm to classify the new records, we build interval basic probability assignments (BPA) for each new record available in text form. Thirdly, we combine the interval BPAs of multiple new records using an evidence combination approach based on evidence theory. Finally, we propagate the interval BPA through the evidential network constructed earlier to obtain the system reliability. Two numerical examples are used to demonstrate the efficiency of the proposed method. We illustrate the effectiveness of the proposed method by comparing with Monte Carlo Simulation (MCS) results. - Highlights: • We model reliability analysis with linguistic data using evidential network. • Two examples are used to demonstrate the efficiency of the proposed method. • We compare the results with Monte Carlo Simulation (MCS).

  5. Reliability analysis for new technology-based transmitters

    Energy Technology Data Exchange (ETDEWEB)

    Brissaud, Florent, E-mail: florent.brissaud.2007@utt.f [Institut National de l' Environnement Industriel et des Risques (INERIS), Parc Technologique Alata, BP 2, 60550 Verneuil-en-Halatte (France); Universite de Technologie de Troyes (UTT), Institut Charles Delaunay (ICD) and STMR UMR CNRS 6279, 12 rue Marie Curie, BP 2060, 10010 Troyes cedex (France); Barros, Anne; Berenguer, Christophe [Universite de Technologie de Troyes (UTT), Institut Charles Delaunay (ICD) and STMR UMR CNRS 6279, 12 rue Marie Curie, BP 2060, 10010 Troyes cedex (France); Charpentier, Dominique [Institut National de l' Environnement Industriel et des Risques (INERIS), Parc Technologique Alata, BP 2, 60550 Verneuil-en-Halatte (France)

    2011-02-15

    The reliability analysis of new technology-based transmitters has to deal with specific issues: various interactions between both material elements and functions, undefined behaviours under faulty conditions, several transmitted data, and little reliability feedback. To handle these particularities, a '3-step' model is proposed, based on goal tree-success tree (GTST) approaches to represent both the functional and material aspects, and includes the faults and failures as a third part for supporting reliability analyses. The behavioural aspects are provided by relationship matrices, also denoted master logic diagrams (MLD), with stochastic values which represent direct relationships between system elements. Relationship analyses are then proposed to assess the effect of any fault or failure on any material element or function. Taking these relationships into account, the probabilities of malfunction and failure modes are evaluated according to time. Furthermore, uncertainty analyses tend to show that even if the input data and system behaviour are not well known, these previous results can be obtained in a relatively precise way. An illustration is provided by a case study on an infrared gas transmitter. These properties make the proposed model and corresponding reliability analyses especially suitable for intelligent transmitters (or 'smart sensors').

  6. Reliability engineering analysis of ATLAS data reprocessing campaigns

    International Nuclear Information System (INIS)

    Vaniachine, A; Golubkov, D; Karpenko, D

    2014-01-01

    During three years of LHC data taking, the ATLAS collaboration completed three petascale data reprocessing campaigns on the Grid, with up to 2 PB of data being reprocessed every year. In reprocessing on the Grid, failures can occur for a variety of reasons, while Grid heterogeneity makes failures hard to diagnose and repair quickly. As a result, Big Data processing on the Grid must tolerate a continuous stream of failures, errors and faults. While ATLAS fault-tolerance mechanisms improve the reliability of Big Data processing in the Grid, their benefits come at costs and result in delays making the performance prediction difficult. Reliability Engineering provides a framework for fundamental understanding of the Big Data processing on the Grid, which is not a desirable enhancement but a necessary requirement. In ATLAS, cost monitoring and performance prediction became critical for the success of the reprocessing campaigns conducted in preparation for the major physics conferences. In addition, our Reliability Engineering approach supported continuous improvements in data reprocessing throughput during LHC data taking. The throughput doubled in 2011 vs. 2010 reprocessing, then quadrupled in 2012 vs. 2011 reprocessing. We present the Reliability Engineering analysis of ATLAS data reprocessing campaigns providing the foundation needed to scale up the Big Data processing technologies beyond the petascale.

  7. Identifying functions for ex-core neutron noise analysis

    International Nuclear Information System (INIS)

    Avila, J.M.; Oliveira, J.C.

    1987-01-01

    A method of performing the phase analysis of signals arising from neutron detectors placed in the periphery of a pressurized water reactor is proposed. It consists in the definition of several identifying functions, based on the phases of cross power spectral densities corresponding to four ex-core neutron detectors. Each of these functions enhances the appearance of different sources of noise. The method, applied to the ex-core neutron fluctuation analysis of a French PWR, proved to be very useful as it allows quick recognition of various patterns in the power spectral densities. (orig.) [de

  8. Multi-Level Simulated Fault Injection for Data Dependent Reliability Analysis of RTL Circuit Descriptions

    Directory of Open Access Journals (Sweden)

    NIMARA, S.

    2016-02-01

    Full Text Available This paper proposes data-dependent reliability evaluation methodology for digital systems described at Register Transfer Level (RTL. It uses a hybrid hierarchical approach, combining the accuracy provided by Gate Level (GL Simulated Fault Injection (SFI and the low simulation overhead required by RTL fault injection. The methodology comprises the following steps: the correct simulation of the RTL system, according to a set of input vectors, hierarchical decomposition of the system into basic RTL blocks, logic synthesis of basic RTL blocks, data-dependent SFI for the GL netlists, and RTL SFI. The proposed methodology has been validated in terms of accuracy on a medium sized circuit – the parallel comparator used in Check Node Unit (CNU of the Low-Density Parity-Check (LDPC decoders. The methodology has been applied for the reliability analysis of a 128-bit Advanced Encryption Standard (AES crypto-core, for which the GL simulation was prohibitive in terms of required computational resources.

  9. Application of noise analysis to investigate core degradation process during PHEBUS-FPT1 test

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1997-01-01

    Noise analysis has been performed for measurement data obtained during PHEBUS-FPT1 test. The purpose of the study is to evaluate the applicability of the noise analysis to the following problems: To get more knowledge about the physical processes going on during severe core conditions; To better understand the core melting process; To establish appropriate on-line shut-down data. Results of the study indicate that the noise analysis is quite promising as a tool for investigating physical processes during the experiment. Compared with conventional approach of evaluating the signal's mean value behaviour, the noise analysis can provide additional, more detailed information: It was found that the neutron flux signal is subjected to additional reactivity perturbations in conjunction with fuel melting and relocation. This can easily be detected by applying noise analysis for the neutron flux signal. It has been demonstrated that the method developed in the present study can provide more accurate estimates of the onset of fuel relocation than using temperature signals from thermocouples in the thermal shroud. Moreover, the result suggests a potential of the present method for tracking the whole process of relocation. The result of the data analysis suggests a possibility of sensor diagnostics which may be important for confirming the quality and reliability of the recorded data. Based on the results achieved it is believed that the combined use of noise analysis and thermocouple signals will provide reliable shut-down criteria for the experiment. 8 refs

  10. Analysis of Irradiation Holes of In-Core Region

    Energy Technology Data Exchange (ETDEWEB)

    In, Won-ho; Lee, Yong-sub; Kim, Tae-hwan; Lim, Kyoung-hwan; Ahn, Hyung-jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Test fuels and materials are irradiated in the in-core region in side of the chimney. The inner chimney is composed of In-Core and Out-Core regions. The In-Core region has 23 hexagonal vertical irradiation holes named from R01 to R20, CT, IR1 and IR2 and 8 cylindrical irradiation holes named from CAR1 to CAR4 and SOR1 to SOR4. The Out-Core region is composed of 8 cylindrical irradiation holes named from OR1 to OR8 which are installed near the inner shell of the reflector tank. HANARO is the multi-purpose research reactor which utilizes in-core irradiation holes, which is being used in various field. Over the past 7 years we have used CT 8 times, IR once, IR2 and OR3 twice, OR4 three times and OR5 ten times. These irradiation holes are used to perform an evaluation of the neutron irradiation properties and the tests were all completed and done successfully. HANARO has been used successfully, and it still will be used continuously in various fields such as nuclear in-pile tests, the production of radioisotopes, neutron transmutation doping, neutron activation analysis, neutron beam research, radiography, environmental science.

  11. Some developments in human reliability analysis approaches and tools

    Energy Technology Data Exchange (ETDEWEB)

    Hannaman, G W; Worledge, D H

    1988-01-01

    Since human actions have been recognized as an important contributor to safety of operating plants in most industries, research has been performed to better understand and account for the way operators interact during accidents through the control room and equipment interface. This paper describes the integration of a series of research projects sponsored by the Electric Power Research Institute to strengthen the methods for performing the human reliability analysis portion of the probabilistic safety studies. It focuses on the analytical framework used to guide the analysis, the development of the models for quantifying time-dependent actions, and simulator experiments used to validate the models.

  12. ANALYSIS OF AVAILABILITY AND RELIABILITY IN RHIC OPERATIONS

    International Nuclear Information System (INIS)

    PILAT, F.; INGRASSIA, P.; MICHNOFF, R.

    2006-01-01

    RHIC has been successfully operated for 5 years as a collider for different species, ranging from heavy ions including gold and copper, to polarized protons. We present a critical analysis of reliability data for RHIC that not only identifies the principal factors limiting availability but also evaluates critical choices at design times and assess their impact on present machine performance. RHIC availability data are typical when compared to similar high-energy colliders. The critical analysis of operations data is the basis for studies and plans to improve RHIC machine availability beyond the 50-60% typical of high-energy colliders

  13. A comparative reliability analysis of free-piston Stirling machines

    Science.gov (United States)

    Schreiber, Jeffrey G.

    2001-02-01

    A free-piston Stirling power convertor is being developed for use in an advanced radioisotope power system to provide electric power for NASA deep space missions. These missions are typically long lived, lasting for up to 14 years. The Department of Energy (DOE) is responsible for providing the radioisotope power system for the NASA missions, and has managed the development of the free-piston power convertor for this application. The NASA Glenn Research Center has been involved in the development of Stirling power conversion technology for over 25 years and is currently providing support to DOE. Due to the nature of the potential missions, long life and high reliability are important features for the power system. Substantial resources have been spent on the development of long life Stirling cryocoolers for space applications. As a very general statement, free-piston Stirling power convertors have many features in common with free-piston Stirling cryocoolers, however there are also significant differences. For example, designs exist for both power convertors and cryocoolers that use the flexure bearing support system to provide noncontacting operation of the close-clearance moving parts. This technology and the operating experience derived from one application may be readily applied to the other application. This similarity does not pertain in the case of outgassing and contamination. In the cryocooler, the contaminants normally condense in the critical heat exchangers and foul the performance. In the Stirling power convertor just the opposite is true as contaminants condense on non-critical surfaces. A methodology was recently published that provides a relative comparison of reliability, and is applicable to systems. The methodology has been applied to compare the reliability of a Stirling cryocooler relative to that of a free-piston Stirling power convertor. The reliability analysis indicates that the power convertor should be able to have superior reliability

  14. Modeling of seismic hazards for dynamic reliability analysis

    International Nuclear Information System (INIS)

    Mizutani, M.; Fukushima, S.; Akao, Y.; Katukura, H.

    1993-01-01

    This paper investigates the appropriate indices of seismic hazard curves (SHCs) for seismic reliability analysis. In the most seismic reliability analyses of structures, the seismic hazards are defined in the form of the SHCs of peak ground accelerations (PGAs). Usually PGAs play a significant role in characterizing ground motions. However, PGA is not always a suitable index of seismic motions. When random vibration theory developed in the frequency domain is employed to obtain statistics of responses, it is more convenient for the implementation of dynamic reliability analysis (DRA) to utilize an index which can be determined in the frequency domain. In this paper, we summarize relationships among the indices which characterize ground motions. The relationships between the indices and the magnitude M are arranged as well. In this consideration, duration time plays an important role in relating two distinct class, i.e. energy class and power class. Fourier and energy spectra are involved in the energy class, and power and response spectra and PGAs are involved in the power class. These relationships are also investigated by using ground motion records. Through these investigations, we have shown the efficiency of employing the total energy as an index of SHCs, which can be determined in the time and frequency domains and has less variance than the other indices. In addition, we have proposed the procedure of DRA based on total energy. (author)

  15. Interrater reliability of videotaped observational gait-analysis assessments.

    Science.gov (United States)

    Eastlack, M E; Arvidson, J; Snyder-Mackler, L; Danoff, J V; McGarvey, C L

    1991-06-01

    The purpose of this study was to determine the interrater reliability of videotaped observational gait-analysis (VOGA) assessments. Fifty-four licensed physical therapists with varying amounts of clinical experience served as raters. Three patients with rheumatoid arthritis who demonstrated an abnormal gait pattern served as subjects for the videotape. The raters analyzed each patient's most severely involved knee during the four subphases of stance for the kinematic variables of knee flexion and genu valgum. Raters were asked to determine whether these variables were inadequate, normal, or excessive. The temporospatial variables analyzed throughout the entire gait cycle were cadence, step length, stride length, stance time, and step width. Generalized kappa coefficients ranged from .11 to .52. Intraclass correlation coefficients (2,1) and (3,1) were slightly higher. Our results indicate that physical therapists' VOGA assessments are only slightly to moderately reliable and that improved interrater reliability of the assessments of physical therapists utilizing this technique is needed. Our data suggest that there is a need for greater standardization of gait-analysis training.

  16. Modeling and Analysis of Component Faults and Reliability

    DEFF Research Database (Denmark)

    Le Guilly, Thibaut; Olsen, Petur; Ravn, Anders Peter

    2016-01-01

    This chapter presents a process to design and validate models of reactive systems in the form of communicating timed automata. The models are extended with faults associated with probabilities of occurrence. This enables a fault tree analysis of the system using minimal cut sets that are automati......This chapter presents a process to design and validate models of reactive systems in the form of communicating timed automata. The models are extended with faults associated with probabilities of occurrence. This enables a fault tree analysis of the system using minimal cut sets...... that are automatically generated. The stochastic information on the faults is used to estimate the reliability of the fault affected system. The reliability is given with respect to properties of the system state space. We illustrate the process on a concrete example using the Uppaal model checker for validating...... the ideal system model and the fault modeling. Then the statistical version of the tool, UppaalSMC, is used to find reliability estimates....

  17. Damage tolerance reliability analysis of automotive spot-welded joints

    International Nuclear Information System (INIS)

    Mahadevan, Sankaran; Ni Kan

    2003-01-01

    This paper develops a damage tolerance reliability analysis methodology for automotive spot-welded joints under multi-axial and variable amplitude loading history. The total fatigue life of a spot weld is divided into two parts, crack initiation and crack propagation. The multi-axial loading history is obtained from transient response finite element analysis of a vehicle model. A three-dimensional finite element model of a simplified joint with four spot welds is developed for static stress/strain analysis. A probabilistic Miner's rule is combined with a randomized strain-life curve family and the stress/strain analysis result to develop a strain-based probabilistic fatigue crack initiation life prediction for spot welds. Afterwards, the fatigue crack inside the base material sheet is modeled as a surface crack. Then a probabilistic crack growth model is combined with the stress analysis result to develop a probabilistic fatigue crack growth life prediction for spot welds. Both methods are implemented with MSC/NASTRAN and MSC/FATIGUE software, and are useful for reliability assessment of automotive spot-welded joints against fatigue and fracture

  18. Reliability analysis of offshore structures using OMA based fatigue stresses

    DEFF Research Database (Denmark)

    Silva Nabuco, Bruna; Aissani, Amina; Glindtvad Tarpø, Marius

    2017-01-01

    focus is on the uncertainty observed on the different stresses used to predict the damage. This uncertainty can be reduced by Modal Based Fatigue Monitoring which is a technique based on continuously measuring of the accelerations in few points of the structure with the use of accelerometers known...... points of the structure, the stress history can be calculated in any arbitrary point of the structure. The accuracy of the estimated actual stress is analyzed by experimental tests on a scale model where the obtained stresses are compared to strain gauges measurements. After evaluating the fatigue...... stresses directly from the operational response of the structure, a reliability analysis is performed in order to estimate the reliability of using Modal Based Fatigue Monitoring for long term fatigue studies....

  19. Issues in benchmarking human reliability analysis methods: A literature review

    International Nuclear Information System (INIS)

    Boring, Ronald L.; Hendrickson, Stacey M.L.; Forester, John A.; Tran, Tuan Q.; Lois, Erasmia

    2010-01-01

    There is a diversity of human reliability analysis (HRA) methods available for use in assessing human performance within probabilistic risk assessments (PRA). Due to the significant differences in the methods, including the scope, approach, and underlying models, there is a need for an empirical comparison investigating the validity and reliability of the methods. To accomplish this empirical comparison, a benchmarking study comparing and evaluating HRA methods in assessing operator performance in simulator experiments is currently underway. In order to account for as many effects as possible in the construction of this benchmarking study, a literature review was conducted, reviewing past benchmarking studies in the areas of psychology and risk assessment. A number of lessons learned through these studies is presented in order to aid in the design of future HRA benchmarking endeavors.

  20. Photovoltaic module reliability improvement through application testing and failure analysis

    Science.gov (United States)

    Dumas, L. N.; Shumka, A.

    1982-01-01

    During the first four years of the U.S. Department of Energy (DOE) National Photovoltatic Program, the Jet Propulsion Laboratory Low-Cost Solar Array (LSA) Project purchased about 400 kW of photovoltaic modules for test and experiments. In order to identify, report, and analyze test and operational problems with the Block Procurement modules, a problem/failure reporting and analysis system was implemented by the LSA Project with the main purpose of providing manufacturers with feedback from test and field experience needed for the improvement of product performance and reliability. A description of the more significant types of failures is presented, taking into account interconnects, cracked cells, dielectric breakdown, delamination, and corrosion. Current design practices and reliability evaluations are also discussed. The conducted evaluation indicates that current module designs incorporate damage-resistant and fault-tolerant features which address field failure mechanisms observed to date.

  1. Mechanical Properties for Reliability Analysis of Structures in Glassy Carbon

    CERN Document Server

    Garion, Cédric

    2014-01-01

    Despite its good physical properties, the glassy carbon material is not widely used, especially for structural applications. Nevertheless, its transparency to particles and temperature resistance are interesting properties for the applications to vacuum chambers and components in high energy physics. For example, it has been proposed for fast shutter valve in particle accelerator [1] [2]. The mechanical properties have to be carefully determined to assess the reliability of structures in such a material. In this paper, mechanical tests have been carried out to determine the elastic parameters, the strength and toughness on commercial grades. A statistical approach, based on the Weibull’s distribution, is used to characterize the material both in tension and compression. The results are compared to the literature and the difference of properties for these two loading cases is shown. Based on a Finite Element analysis, a statistical approach is applied to define the reliability of a structural component in gl...

  2. User's manual of a support system for human reliability analysis

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Tamura, Kazuo.

    1995-10-01

    Many kinds of human reliability analysis (HRA) methods have been developed. However, users are required to be skillful so as to use them, and also required complicated works such as drawing event tree (ET) and calculation of uncertainty bounds. Moreover, each method is not so complete that only one method of them is not enough to evaluate human reliability. Therefore, a personal computer (PC) based support system for HRA has been developed to execute HRA practically and efficiently. The system consists of two methods, namely, simple method and detailed one. The former uses ASEP that is a simplified THERP-technique, and combined method of OAT and HRA-ET/DeBDA is used for the latter. Users can select a suitable method for their purpose. Human error probability (HEP) data were collected and a database of them was built to use for the support system. This paper describes outline of the HRA methods, support functions and user's guide of the system. (author)

  3. Issues in benchmarking human reliability analysis methods : a literature review.

    Energy Technology Data Exchange (ETDEWEB)

    Lois, Erasmia (US Nuclear Regulatory Commission); Forester, John Alan; Tran, Tuan Q. (Idaho National Laboratory, Idaho Falls, ID); Hendrickson, Stacey M. Langfitt; Boring, Ronald L. (Idaho National Laboratory, Idaho Falls, ID)

    2008-04-01

    There is a diversity of human reliability analysis (HRA) methods available for use in assessing human performance within probabilistic risk assessment (PRA). Due to the significant differences in the methods, including the scope, approach, and underlying models, there is a need for an empirical comparison investigating the validity and reliability of the methods. To accomplish this empirical comparison, a benchmarking study is currently underway that compares HRA methods with each other and against operator performance in simulator studies. In order to account for as many effects as possible in the construction of this benchmarking study, a literature review was conducted, reviewing past benchmarking studies in the areas of psychology and risk assessment. A number of lessons learned through these studies are presented in order to aid in the design of future HRA benchmarking endeavors.

  4. Review of the treat upgrade reactor scram system reliability analysis

    International Nuclear Information System (INIS)

    Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.

    1984-10-01

    In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals

  5. Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

    International Nuclear Information System (INIS)

    Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

    2001-01-01

    The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

  6. A retrospective analysis of ultrasound-guided large core needle ...

    African Journals Online (AJOL)

    A retrospective analysis of ultrasound-guided large core needle biopsies of breast lesions at a regional public hospital in Durban, KwaZulu-Natal, South Africa. ... Objective: To assess the influence of technical variables on the diagnostic yield of breast specimens obtained by using US-LCNB, and the sensitivity of detecting ...

  7. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  8. Reliability analysis of self-supply system of V-1 nuclear power plant

    International Nuclear Information System (INIS)

    Kuklik, B.

    The results are summarized of the fault tree analysis of the V-1 power plant self-consumption system. The 6 kV busbars providing power for the main circulating pumps, the steam generator feed pumps and other important components including the 0.4 kV busbars are of the highest importance for nuclear safety. A fault tree analysis was also made of the emergency core cooling system of the reactor. Dangerous faults are defined and fault trees are developed. A brief description is given of the calculation algorithm for a digital computer. Some results are discussed. The calculated reliability of the emergency core cooling system is 10 5 years, of the 6 kV busbars it is 6.6x10 4 years. In case of a permanent or a long-term outage of the 220 kV stand-bye power supply, the system reliability is reduced to 7x10 2 years. (Z.M.)

  9. Reliability Analysis of the CERN Radiation Monitoring Electronic System CROME

    CERN Document Server

    AUTHOR|(CDS)2126870

    For the new in-house developed CERN Radiation Monitoring Electronic System (CROME) a reliability analysis is necessary to ensure compliance with the statu-tory requirements regarding the Safety Integrity Level. The required Safety Integrity Level by IEC 60532 standard is SIL 2 (for the Safety Integrated Functions Measurement, Alarm Triggering and Interlock Triggering). The first step of the reliability analysis was a system and functional analysis which served as basis for the implementation of the CROME system in the software “Iso-graph”. In the “Prediction” module of Isograph the failure rates of all components were calculated. Failure rates for passive components were calculated by the Military Standard 217 and failure rates for active components were obtained from lifetime tests by the manufacturers. The FMEA was carried out together with the board designers and implemented in the “FMECA” module of Isograph. The FMEA served as basis for the Fault Tree Analysis and the detection of weak points...

  10. Decision theory, the context for risk and reliability analysis

    International Nuclear Information System (INIS)

    Kaplan, S.

    1985-01-01

    According to this model of the decision process then, the optimum decision is that option having the largest expected utility. This is the fundamental model of a decision situation. It is necessary to remark that in order for the model to represent a real-life decision situation, it must include all the options present in that situation, including, for example, the option of not deciding--which is itself a decision, although usually not the optimum one. Similarly, it should include the option of delaying the decision while the authors gather further information. Both of these options have probabilities, outcomes, impacts, and utilities like any option and should be included explicitly in the decision diagram. The reason for doing a quantitative risk or reliability analysis is always that, somewhere underlying there is a decision to be made. The decision analysis therefore always forms the context for the risk or reliability analysis, and this context shapes the form and language of that analysis. Therefore, they give in this section a brief review of the well-known decision theory diagram

  11. Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Soares, Wellington Antonio; Marques, Raíssa Oliveira; Silva Júnior, Silvério Ferreira da; Raso, Amanda Laureano

    2017-01-01

    Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. NDI is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI methods are reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components. Among these can be highlighted Failure Modes and Effects Analysis (FMEA) and THERP (Technique for Human Error Rate Prediction). The application of these techniques is illustrated in an example of qualitative and quantitative studies to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues. (author)

  12. Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

    Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. NDI is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI methods are reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components. Among these can be highlighted Failure Modes and Effects Analysis (FMEA) and THERP (Technique for Human Error Rate Prediction). The application of these techniques is illustrated in an example of qualitative and quantitative studies to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues. (author)

  13. Core-Shell Columns in High-Performance Liquid Chromatography: Food Analysis Applications

    Science.gov (United States)

    Preti, Raffaella

    2016-01-01

    The increased separation efficiency provided by the new technology of column packed with core-shell particles in high-performance liquid chromatography (HPLC) has resulted in their widespread diffusion in several analytical fields: from pharmaceutical, biological, environmental, and toxicological. The present paper presents their most recent applications in food analysis. Their use has proved to be particularly advantageous for the determination of compounds at trace levels or when a large amount of samples must be analyzed fast using reliable and solvent-saving apparatus. The literature hereby described shows how the outstanding performances provided by core-shell particles column on a traditional HPLC instruments are comparable to those obtained with a costly UHPLC instrumentation, making this novel column a promising key tool in food analysis. PMID:27143972

  14. In-core LOCA (PTR) analysis with poisoned moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, T. M.; Choi, J. H.; Kim, Yun Ho; Choi, Hoon

    2005-01-01

    CANDU reactors have been analyzed and evaluated for the postulated in-core LOCA while the reactor is operating normally with low moderator poison concentration. However, when the reactor is operating with relatively large amounts of boron and/or gadolinium poisons in the moderator, the assessment for fuel integrity was required for pressure tube rupture (PTR) accident. The methodology of in-core LOCA analysis with poisoned moderator is developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for CANDU reactor recently. The developed methodology and results are presented

  15. A Research Roadmap for Computation-Based Human Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

  16. A Research Roadmap for Computation-Based Human Reliability Analysis

    International Nuclear Information System (INIS)

    Boring, Ronald; Mandelli, Diego; Joe, Jeffrey; Smith, Curtis; Groth, Katrina

    2015-01-01

    The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

  17. Limitations in simulator time-based human reliability analysis methods

    International Nuclear Information System (INIS)

    Wreathall, J.

    1989-01-01

    Developments in human reliability analysis (HRA) methods have evolved slowly. Current methods are little changed from those of almost a decade ago, particularly in the use of time-reliability relationships. While these methods were suitable as an interim step, the time (and the need) has come to specify the next evolution of HRA methods. As with any performance-oriented data source, power plant simulator data have no direct connection to HRA models. Errors reported in data are normal deficiencies observed in human performance; failures are events modeled in probabilistic risk assessments (PRAs). Not all errors cause failures; not all failures are caused by errors. Second, the times at which actions are taken provide no measure of the likelihood of failures to act correctly within an accident scenario. Inferences can be made about human reliability, but they must be made with great care. Specific limitations are discussed. Simulator performance data are useful in providing qualitative evidence of the variety of error types and their potential influences on operating systems. More work is required to combine recent developments in the psychology of error with the qualitative data collected at stimulators. Until data become openly available, however, such an advance will not be practical

  18. The DYLAM approach for the dynamic reliability analysis of systems

    International Nuclear Information System (INIS)

    Cojazzi, Giacomo

    1996-01-01

    In many real systems, failures occurring to the components, control failures and human interventions often interact with the physical system evolution in such a way that a simple reliability analysis, de-coupled from process dynamics, is very difficult or even impossible. In the last ten years many dynamic reliability approaches have been proposed to properly assess the reliability of these systems characterized by dynamic interactions. The DYLAM methodology, now implemented in its latest version, DYLAM-3, offers a powerful tool for integrating deterministic and failure events. This paper describes the main features of the DYLAM-3 code with reference to the classic fault-tree and event-tree techniques. Some aspects connected to the practical problems underlying dynamic event-trees are also discussed. A simple system, already analyzed with other dynamic methods is used as a reference for the numerical applications. The same system is also studied with a time-dependent fault-tree approach in order to show some features of dynamic methods vs classical techniques. Examples including stochastic failures, without and with repair, failures on demand and time dependent failure rates give an extensive overview of DYLAM-3 capabilities

  19. Reliability analysis of self-actuated shutdown system

    International Nuclear Information System (INIS)

    Itooka, S.; Kumasaka, K.; Okabe, A.; Satoh, K.; Tsukui, Y.

    1991-01-01

    An analytical study was performed for the reliability of a self-actuated shutdown system (SASS) under the unprotected loss of flow (ULOF) event in a typical loop-type liquid metal fast breeder reactor (LMFBR) by the use of the response surface Monte Carlo analysis method. Dominant parameters for the SASS, such as Curie point characteristics, subassembly outlet coolant temperature, electromagnetic surface condition, etc., were selected and their probability density functions (PDFs) were determined by the design study information and experimental data. To get the response surface function (RSF) for the maximum coolant temperature, transient analyses of ULOF were performed by utilizing the experimental design method in the determination of analytical cases. Then, the RSF was derived by the multi-variable regression analysis. The unreliability of the SASS was evaluated as a probability that the maximum coolant temperature exceeded an acceptable level, employing the Monte Carlo calculation using the above PDFs and RSF. In this study, sensitivities to the dominant parameter were compared. The dispersion of subassembly outlet coolant temperature near the SASS-was found to be one of the most sensitive parameters. Fault tree analysis was performed using this value for the SASS in order to evaluate the shutdown system reliability. As a result of this study, the effectiveness of the SASS on the reliability improvement in the LMFBR shutdown system was analytically confirmed. This study has been performed as a part of joint research and development projects for DFBR under the sponsorship of the nine Japanese electric power companies, Electric Power Development Company and the Japan Atomic Power Company. (author)

  20. Failure and Reliability Analysis for the Master Pump Shutdown System

    International Nuclear Information System (INIS)

    BEVINS, R.R.

    2000-01-01

    The Master Pump Shutdown System (MPSS) will be installed in the 200 Areas of the Hanford Site to monitor and control the transfer of liquid waste between tank farms and between the 200 West and 200 East areas through the Cross-Site Transfer Line. The Safety Function provided by the MPSS is to shutdown any waste transfer process within or between tank farms if a waste leak should occur along the selected transfer route. The MPSS, which provides this Safety Class Function, is composed of Programmable Logic Controllers (PLCs), interconnecting wires, relays, Human to Machine Interfaces (HMI), and software. These components are defined as providing a Safety Class Function and will be designated in this report as MPSS/PLC. Input signals to the MPSS/PLC are provided by leak detection systems from each of the tank farm leak detector locations along the waste transfer route. The combination of the MPSS/PLC, leak detection system, and transfer pump controller system will be referred to as MPSS/SYS. The components addressed in this analysis are associated with the MPSS/SYS. The purpose of this failure and reliability analysis is to address the following design issues of the Project Development Specification (PDS) for the MPSS/SYS (HNF 2000a): (1) Single Component Failure Criterion, (2) System Status Upon Loss of Electrical Power, (3) Physical Separation of Safety Class cables, (4) Physical Isolation of Safety Class Wiring from General Service Wiring, and (5) Meeting the MPSS/PLC Option 1b (RPP 1999) Reliability estimate. The failure and reliability analysis examined the system on a component level basis and identified any hardware or software elements that could fail and/or prevent the system from performing its intended safety function

  1. Fifty Years of THERP and Human Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring

    2012-06-01

    In 1962 at a Human Factors Society symposium, Alan Swain presented a paper introducing a Technique for Human Error Rate Prediction (THERP). This was followed in 1963 by a Sandia Laboratories monograph outlining basic human error quantification using THERP and, in 1964, by a special journal edition of Human Factors on quantification of human performance. Throughout the 1960s, Swain and his colleagues focused on collecting human performance data for the Sandia Human Error Rate Bank (SHERB), primarily in connection with supporting the reliability of nuclear weapons assembly in the US. In 1969, Swain met with Jens Rasmussen of Risø National Laboratory and discussed the applicability of THERP to nuclear power applications. By 1975, in WASH-1400, Swain had articulated the use of THERP for nuclear power applications, and the approach was finalized in the watershed publication of the NUREG/CR-1278 in 1983. THERP is now 50 years old, and remains the most well known and most widely used HRA method. In this paper, the author discusses the history of THERP, based on published reports and personal communication and interviews with Swain. The author also outlines the significance of THERP. The foundations of human reliability analysis are found in THERP: human failure events, task analysis, performance shaping factors, human error probabilities, dependence, event trees, recovery, and pre- and post-initiating events were all introduced in THERP. While THERP is not without its detractors, and it is showing signs of its age in the face of newer technological applications, the longevity of THERP is a testament of its tremendous significance. THERP started the field of human reliability analysis. This paper concludes with a discussion of THERP in the context of newer methods, which can be seen as extensions of or departures from Swain’s pioneering work.

  2. Reliability and Robustness Analysis of the Masinga Dam under Uncertainty

    Directory of Open Access Journals (Sweden)

    Hayden Postle-Floyd

    2017-02-01

    Full Text Available Kenya’s water abstraction must meet the projected growth in municipal and irrigation demand by the end of 2030 in order to achieve the country’s industrial and economic development plan. The Masinga dam, on the Tana River, is the key to meeting this goal to satisfy the growing demands whilst also continuing to provide hydroelectric power generation. This study quantitatively assesses the reliability and robustness of the Masinga dam system under uncertain future supply and demand using probabilistic climate and population projections, and examines how long-term planning may improve the longevity of the dam. River flow and demand projections are used alongside each other as inputs to the dam system simulation model linked to an optimisation engine to maximise water availability. Water availability after demand satisfaction is assessed for future years, and the projected reliability of the system is calculated for selected years. The analysis shows that maximising power generation on a short-term year-by-year basis achieves 80%, 50% and 1% reliability by 2020, 2025 and 2030 onwards, respectively. Longer term optimal planning, however, has increased system reliability to up to 95% in 2020, 80% in 2025, and more than 40% in 2030 onwards. In addition, increasing the capacity of the reservoir by around 25% can significantly improve the robustness of the system for all future time periods. This study provides a platform for analysing the implication of different planning and management of Masinga dam and suggests that careful consideration should be given to account for growing municipal needs and irrigation schemes in both the immediate and the associated Tana River basin.

  3. Interim reliability evaluation program: analysis of the Arkansas Nuclear One. Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kolb, G.J.; Kunsman, D.M.; Bell, B.J.

    1982-06-01

    This report represents the results of the analysis of Arkansas Nuclear One (ANO) Unit 1 nuclear power plant which was performed as part of the Interim Reliability Evaluation Program (IREP). The IREP has several objectives, two of which are achieved by the analysis presented in this report. These objectives are: (1) the identification, in a preliminary way, of those accident sequences which are expected to dominate the public health and safety risks; and (2) the development of state-of-the-art plant system models which can be used as a foundation for subsequent, more intensive applications of probabilistic risk assessment. The primary methodological tools used in the analysis were event trees and fault trees. These tools were used to study core melt accidents initiated by loss of coolant accidents (LOCAs) of six different break size ranges and eight different types of transients

  4. Code Coupling for Multi-Dimensional Core Transient Analysis

    International Nuclear Information System (INIS)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

    2015-01-01

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

  5. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  6. Refurbishment, core conversion and safety analysis of Apsara reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K.; Sasidharan, K.; Sengupta, S. [Bhabha Atomic Research Centre, Mumbai (India)]. E-mail: nram@@apsara.barc.ernet.in

    1998-07-01

    Apsara, a 1 MWt pool type reactor using HEU fuel has been in operation at the Bhabha Atomic Research Centre, Trombay since 1956. In view of the long service period seen by the reactor it is now planned to carry out extensive refurbishment of the reactor with a view to extend its useful life. It is also proposed to modify the design of the reactor wherein the core will be surrounded by a heavy water reflector tank to obtain a good thermal neutron flux over a large radial distance from the core. Beam holes and the majority of the irradiation facilities will be located inside the reflector tank. The coolant flow direction through the core will be changed from the existing upward flow to downward flow. A delay tank, located inside the pool, is provided to facilitate decay of short lived radioactivity in the coolant outlet from the core in order to bring down radiation field in the operating areas. Analysis of various anticipated operational occurrences and accident conditions like loss of normal power, core coolant flow bypass, fuel channel blockage and degradation of primary coolant pressure boundary have been performed for the proposed design. Details of the proposed design modifications and the safety analyses are given in the paper. (author)

  7. Preliminary analysis of the proposed BN-600 benchmark core

    International Nuclear Information System (INIS)

    John, T.M.

    2000-01-01

    The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

  8. Human reliability analysis during PSA at Trillo NPP: main characteristics and analysis of diagnostic errors

    International Nuclear Information System (INIS)

    Barquin, M.A.; Gomez, F.

    1998-01-01

    The design difference between Trillo NPP and other Spanish nuclear power plants (basic Westinghouse and General Electric designs) were made clear in the Human Reliability Analysis of the Probabilistic Safety Analysis (PSA) for Trillo NPP. The object of this paper is to describe the most significant characteristics of the Human Reliability Analysis carried out in the PSA, with special emphasis on the possible diagnostic errors and their consequences, based on the characteristics in the Emergency Operations Manual for Trillo NPP. - In the case of human errors before the initiating event (type 1), the existence of four redundancies in most of the plant safety systems, means that the impact of this type or error on the final results of the PSA is insignificant. However, in the case common cause errors, especially in certain calibration errors, some actions are significant in the final equation for core damage - The number of human actions that the operator has to carry out during the accidents (type 3) modelled, is relatively small in comparison with this value in other PSAs. This is basically due to the high level of automation at Rillo NPP - The Plant Operations Manual cannot be strictly considered to be a symptoms-based procedure. The operation Group must select the chapter from the Operations Manual to be followed, after having diagnosed the perturbing event, using for this purpose and Emergency and Anomaly Decision Tree (M.O.3.0.1) based on the different indications, alarms and symptoms present in the plant after the perturbing event. For this reason, it was decided to analyse the possible diagnosis errors. In the bibliography on diagnosis and commission errors available at the present time, there is no precise methodology for the analysis of this type of error and its incorporation into PSAs. The method used in the PSA for Trillo y NPP to evaluate this type of interaction, is to develop a Diagnosis Error Table, the object of which is to identify the situations in

  9. Reliability analysis of structures under periodic proof tests in service

    Science.gov (United States)

    Yang, J.-N.

    1976-01-01

    A reliability analysis of structures subjected to random service loads and periodic proof tests treats gust loads and maneuver loads as random processes. Crack initiation, crack propagation, and strength degradation are treated as the fatigue process. The time to fatigue crack initiation and ultimate strength are random variables. Residual strength decreases during crack propagation, so that failure rate increases with time. When a structure fails under periodic proof testing, a new structure is built and proof-tested. The probability of structural failure in service is derived from treatment of all the random variables, strength degradations, service loads, proof tests, and the renewal of failed structures. Some numerical examples are worked out.

  10. Creation and Reliability Analysis of Vehicle Dynamic Weighing Model

    Directory of Open Access Journals (Sweden)

    Zhi-Ling XU

    2014-08-01

    Full Text Available In this paper, it is modeled by using ADAMS to portable axle load meter of dynamic weighing system, controlling a single variable simulation weighing process, getting the simulation weighing data under the different speed and weight; simultaneously using portable weighing system with the same parameters to achieve the actual measurement, comparative analysis the simulation results under the same conditions, at 30 km/h or less, the simulation value and the measured value do not differ by more than 5 %, it is not only to verify the reliability of dynamic weighing model, but also to create possible for improving algorithm study efficiency by using dynamic weighing model simulation.

  11. Human Performance Modeling for Dynamic Human Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Boring, Ronald Laurids [Idaho National Laboratory; Joe, Jeffrey Clark [Idaho National Laboratory; Mandelli, Diego [Idaho National Laboratory

    2015-08-01

    Part of the U.S. Department of Energy’s (DOE’s) Light Water Reac- tor Sustainability (LWRS) Program, the Risk-Informed Safety Margin Charac- terization (RISMC) Pathway develops approaches to estimating and managing safety margins. RISMC simulations pair deterministic plant physics models with probabilistic risk models. As human interactions are an essential element of plant risk, it is necessary to integrate human actions into the RISMC risk framework. In this paper, we review simulation based and non simulation based human reliability analysis (HRA) methods. This paper summarizes the founda- tional information needed to develop a feasible approach to modeling human in- teractions in RISMC simulations.

  12. WWER-440 reactor thermal power increase. Up-to-date approaches to substantiation of the core heat-engineering reliability

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Lushin, V.; Zubtsov, D.

    2006-01-01

    Increasing the Units power is an urgent problem for nuclear power plants with WWER-440 reactors. Improving the fuel assembly designs and calculated codes creates all prerequisites to fulfil this purpose. The decrease in the core power peaking is reached by using the profiled fuel assemblies, burnable absorber integrated into the fuel, the FA with the modernized interface attachment, modern calculated codes that allows to reduce conservatism of the RP safety substantiation. A wide spectrum of experimental study of behaviour of the fuel having reached burn-up (50-60) MW days / kg U under the transients and accident conditions was carried out, the post-irradiated examination of the fuel assemblies, fuel rods and fuel pellets with four and five annual operating fuel cycle were performed as well and confirmed the high reliability of the fuel, the presence of large margins of the fuel stack state that contributes to reactor operation at the increased power. The results of the carried out experiments on implementing the five and six annual fuel cycles show that the limiting fuel state as to its serviceability in the WWER-440 reactors is far from being reached. Presently there is an experience of the increased power operation of Kola NPP, Units 1, 2, 4 and Rovno NPP, Unit 2. The Loviisa NPP Units are operated at 109 % power. The Russian experts had gained an experience in substantiating the core operation at 108 % power for Paks NPP, Unit 4. In this paper the additional conditions for increasing the power of the Kola NPP, Units 1 and 2 and the main results of substantiation of increase in power of the Paks NPP, Unit 4 up to 1485 MW are presented in details

  13. IDHEAS – A NEW APPROACH FOR HUMAN RELIABILITY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    G. W. Parry; J.A Forester; V.N. Dang; S. M. L. Hendrickson; M. Presley; E. Lois; J. Xing

    2013-09-01

    This paper describes a method, IDHEAS (Integrated Decision-Tree Human Event Analysis System) that has been developed jointly by the US NRC and EPRI as an improved approach to Human Reliability Analysis (HRA) that is based on an understanding of the cognitive mechanisms and performance influencing factors (PIFs) that affect operator responses. The paper describes the various elements of the method, namely the performance of a detailed cognitive task analysis that is documented in a crew response tree (CRT), and the development of the associated time-line to identify the critical tasks, i.e. those whose failure results in a human failure event (HFE), and an approach to quantification that is based on explanations of why the HFE might occur.

  14. Scoping Analysis on Core Disruptive Accident in PGSFR (2015 Results)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Won; Chang, Won-Pyo; Ha, Kwi-Seok; Ahn, Sang June; Kang, Seok Hun; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Kim, Jin Su; Jeong, Taekyeong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In general, the severe accident is classified by three phases. The first phase is the initiation (pre-disassembly) phase that occurs the gradual core meltdown from accident initiation to the point of neutronic shutdown with an intact geometry. The second phase is the transition phase that happens the fuel transition from a solid to a liquid phase. Fuel and cladding can melt to form a molten pool and core can boil, then criticality conditions can recur. The third phase is the disassembly phase. In other words, this phase is Core Disruptive Accident (CDA). Power excursion is followed until the core is disassembled in this phase. In the early considerations of Liquid Metal Fast Breeder Reactor (LMFBR) energetics, the term Hypothetical Core Disruptive Accidents (HCDAs) was in common use. This was not only to connote the extremely low probability of initiation of such accidents, but also the tentative nature of our understanding of their behavior and resulting consequences. A numerical analysis is conducted to estimate the energy release, pressure behavior and core expansion behavior induced by CDA of PGSFR using CDA-ER and CDA-CEME codes. Conservatively, the calculated results of energy release and pressure behavior induced by CDA without Doppler effect in PGSFR when whole cores were melted (100 $/s) were 7.844 GJ and 4.845 GPa, respectively. With Doppler effect, the analyzed maximum energy release and pressure were 6.696 GJ and 3.449 GPa, respectively. The calculated results of the core expansion behavior during 0.015 seconds after the explosion without Doppler effect in PGSFR when whole cores were melted (100 $/s) were as follows: The total energy is calculated to be 1.87 GJ. At 0.01 s, the kinetic energy of the sodium is 1.85 GJ, while the expansion work and internal energy of the bubble are 19.7 MJ and 0.98 J, respectively. With Doppler effect, the total energy is calculated to be 1.33 GJ. At 0.01 s, the kinetic energy of the sodium is 1.31 GJ, while the expansion

  15. Analysis of core damage frequency: Surry, Unit 1 internal events

    International Nuclear Information System (INIS)

    Bertucio, R.C.; Julius, J.A.; Cramond, W.R.

    1990-04-01

    This document contains the accident sequence analysis of internally initiated events for the Surry Nuclear Station, Unit 1. This is one of the five plant analyses conducted as part of the NUREG-1150 effort by the Nuclear Regulatory Commission. NUREG-1150 documents the risk of a selected group of nuclear power plants. The work performed and described here is an extensive of that published in November 1986 as NUREG/CR-4450, Volume 3. It addresses comments form numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved. The context and detail of this report are directed toward PRA practitioners who need to know how the work was performed and the details for use in further studies. The mean core damage frequency at Surry was calculated to be 4.05-E-5 per year, with a 95% upper bound of 1.34E-4 and 5% lower bound of 6.8E-6 per year. Station blackout type accidents (loss of all AC power) were the largest contributors to the core damage frequency, accounting for approximately 68% of the total. The next type of dominant contributors were Loss of Coolant Accidents (LOCAs). These sequences account for 15% of core damage frequency. No other type of sequence accounts for more than 10% of core damage frequency. 49 refs., 52 figs., 70 tabs

  16. Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents

    International Nuclear Information System (INIS)

    Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.

    1987-01-01

    This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering

  17. Tailoring a Human Reliability Analysis to Your Industry Needs

    Science.gov (United States)

    DeMott, D. L.

    2016-01-01

    Companies at risk of accidents caused by human error that result in catastrophic consequences include: airline industry mishaps, medical malpractice, medication mistakes, aerospace failures, major oil spills, transportation mishaps, power production failures and manufacturing facility incidents. Human Reliability Assessment (HRA) is used to analyze the inherent risk of human behavior or actions introducing errors into the operation of a system or process. These assessments can be used to identify where errors are most likely to arise and the potential risks involved if they do occur. Using the basic concepts of HRA, an evolving group of methodologies are used to meet various industry needs. Determining which methodology or combination of techniques will provide a quality human reliability assessment is a key element to developing effective strategies for understanding and dealing with risks caused by human errors. There are a number of concerns and difficulties in "tailoring" a Human Reliability Assessment (HRA) for different industries. Although a variety of HRA methodologies are available to analyze human error events, determining the most appropriate tools to provide the most useful results can depend on industry specific cultures and requirements. Methodology selection may be based on a variety of factors that include: 1) how people act and react in different industries, 2) expectations based on industry standards, 3) factors that influence how the human errors could occur such as tasks, tools, environment, workplace, support, training and procedure, 4) type and availability of data, 5) how the industry views risk & reliability, and 6) types of emergencies, contingencies and routine tasks. Other considerations for methodology selection should be based on what information is needed from the assessment. If the principal concern is determination of the primary risk factors contributing to the potential human error, a more detailed analysis method may be employed

  18. A task analysis-linked approach for integrating the human factor in reliability assessments of nuclear power plants

    International Nuclear Information System (INIS)

    Ryan, T.G.

    1988-01-01

    This paper describes an emerging Task Analysis-Linked Evaluation Technique (TALENT) for assessing the contributions of human error to nuclear power plant systems unreliability and risk. Techniques such as TALENT are emerging as a recognition that human error is a primary contributor to plant safety, however, it has been a peripheral consideration to data in plant reliability evaluations. TALENT also recognizes that involvement of persons with behavioral science expertise is required to support plant reliability and risk analyses. A number of state-of-knowledge human reliability analysis tools are also discussed which support the TALENT process. The core of TALENT is comprised of task, timeline and interface analysis data which provide the technology base for event and fault tree development, serve as criteria for selecting and evaluating performance shaping factors, and which provide a basis for auditing TALENT results. Finally, programs and case studies used to refine the TALENT process are described along with future research needs in the area. (author)

  19. Inclusion of fatigue effects in human reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Griffith, Candice D. [Vanderbilt University, Nashville, TN (United States); Mahadevan, Sankaran, E-mail: sankaran.mahadevan@vanderbilt.edu [Vanderbilt University, Nashville, TN (United States)

    2011-11-15

    The effect of fatigue on human performance has been observed to be an important factor in many industrial accidents. However, defining and measuring fatigue is not easily accomplished. This creates difficulties in including fatigue effects in probabilistic risk assessments (PRA) of complex engineering systems that seek to include human reliability analysis (HRA). Thus the objectives of this paper are to discuss (1) the importance of the effects of fatigue on performance, (2) the difficulties associated with defining and measuring fatigue, (3) the current status of inclusion of fatigue in HRA methods, and (4) the future directions and challenges for the inclusion of fatigue, specifically sleep deprivation, in HRA. - Highlights: >We highlight the need for fatigue and sleep deprivation effects on performance to be included in human reliability analysis (HRA) methods. Current methods do not explicitly include sleep deprivation effects. > We discuss the difficulties in defining and measuring fatigue. > We review sleep deprivation research, and discuss the limitations and future needs of the current HRA methods.

  20. A methodology for strain-based fatigue reliability analysis

    International Nuclear Information System (INIS)

    Zhao, Y.X.

    2000-01-01

    A significant scatter of the cyclic stress-strain (CSS) responses should be noted for a nuclear reactor material, 1Cr18Ni9Ti pipe-weld metal. Existence of the scatter implies that a random cyclic strain applied history will be introduced under any of the loading modes even a deterministic loading history. A non-conservative evaluation might be given in the practice without considering the scatter. A methodology for strain-based fatigue reliability analysis, which has taken into account the scatter, is developed. The responses are approximately modeled by probability-based CSS curves of Ramberg-Osgood relation. The strain-life data are modeled, similarly, by probability-based strain-life curves of Coffin-Manson law. The reliability assessment is constructed by considering interference of the random fatigue strain applied and capacity histories. Probability density functions of the applied and capacity histories are analytically given. The methodology could be conveniently extrapolated to the case of deterministic CSS relation as the existent methods did. Non-conservative evaluation of the deterministic CSS relation and availability of present methodology have been indicated by an analysis of the material test results

  1. Inclusion of fatigue effects in human reliability analysis

    International Nuclear Information System (INIS)

    Griffith, Candice D.; Mahadevan, Sankaran

    2011-01-01

    The effect of fatigue on human performance has been observed to be an important factor in many industrial accidents. However, defining and measuring fatigue is not easily accomplished. This creates difficulties in including fatigue effects in probabilistic risk assessments (PRA) of complex engineering systems that seek to include human reliability analysis (HRA). Thus the objectives of this paper are to discuss (1) the importance of the effects of fatigue on performance, (2) the difficulties associated with defining and measuring fatigue, (3) the current status of inclusion of fatigue in HRA methods, and (4) the future directions and challenges for the inclusion of fatigue, specifically sleep deprivation, in HRA. - Highlights: →We highlight the need for fatigue and sleep deprivation effects on performance to be included in human reliability analysis (HRA) methods. Current methods do not explicitly include sleep deprivation effects. → We discuss the difficulties in defining and measuring fatigue. → We review sleep deprivation research, and discuss the limitations and future needs of the current HRA methods.

  2. Analysis of dependent failures in risk assessment and reliability evaluation

    International Nuclear Information System (INIS)

    Fleming, K.N.; Mosleh, A.; Kelley, A.P. Jr.; Gas-Cooled Reactors Associates, La Jolla, CA)

    1983-01-01

    The ability to estimate the risk of potential reactor accidents is largely determined by the ability to analyze statistically dependent multiple failures. The importance of dependent failures has been indicated in recent probabilistic risk assessment (PRA) studies as well as in reports of reactor operating experiences. This article highlights the importance of several different types of dependent failures from the perspective of the risk and reliability analyst and provides references to the methods and data available for their analysis. In addition to describing the current state of the art, some recent advances, pitfalls, misconceptions, and limitations of some approaches to dependent failure analysis are addressed. A summary is included of the discourse on this subject, which is presented in the Institute of Electrical and Electronics Engineers/American Nuclear Society PRA Procedures Guide

  3. Current Human Reliability Analysis Methods Applied to Computerized Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring

    2012-06-01

    Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no US nuclear power plant has implemented CPs in its main control room (Fink et al., 2009). Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of enhanced ease of use and easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

  4. Standardization of domestic human reliability analysis and experience of human reliability analysis in probabilistic safety assessment for NPPs under design

    International Nuclear Information System (INIS)

    Kang, D. I.; Jung, W. D.

    2002-01-01

    This paper introduces the background and development activities of domestic standardization of procedure and method for Human Reliability Analysis (HRA) to avoid the intervention of subjectivity by HRA analyst in Probabilistic Safety Assessment (PSA) as possible, and the review of the HRA results for domestic nuclear power plants under design studied by Korea Atomic Energy Research Institute. We identify the HRA methods used for PSA for domestic NPPs and discuss the subjectivity of HRA analyst shown in performing a HRA. Also, we introduce the PSA guidelines published in USA and review the HRA results based on them. We propose the system of a standard procedure and method for HRA to be developed

  5. Bypass Flow and Hot Spot Analysis for PMR200 Block-Core Design with Core Restraint Mechanism

    International Nuclear Information System (INIS)

    Lim, Hong Sik; Kim, Min Hwan

    2009-01-01

    The accurate prediction of local hot spot during normal operation is important to ensure core thermal margin in a very high temperature gas-cooled reactor because of production of its high temperature output. The active cooling of the reactor core determining local hot spot is strongly affected by core bypass flows through the inter-column gaps between graphite blocks and the cross gaps between two stacked fuel blocks. The bypass gap sizes vary during core life cycle by the thermal expansion at the elevated temperature and the shrinkage/swelling by fast neutron irradiation. This study is to investigate the impacts of the variation of bypass gaps during core life cycle as well as core restraint mechanism on the amount of bypass flow and thus maximum fuel temperature. The core thermo fluid analysis is performed using the GAMMA+ code for the PMR200 block-core design. For the analysis not only are some modeling features, developed for solid conduction and bypass flow, are implemented into the GAMMA+ code but also non-uniform bypass gap distribution taken from a tool calculating the thermal expansion and the shrinkage/swell of graphite during core life cycle under the design options with and without core restraint mechanism is used

  6. Systems reliability analysis: applications of the SPARCS System-Reliability Assessment Computer Program

    International Nuclear Information System (INIS)

    Locks, M.O.

    1978-01-01

    SPARCS-2 (Simulation Program for Assessing the Reliabilities of Complex Systems, Version 2) is a PL/1 computer program for assessing (establishing interval estimates for) the reliability and the MTBF of a large and complex s-coherent system of any modular configuration. The system can consist of a complex logical assembly of independently failing attribute (binomial-Bernoulli) and time-to-failure (Poisson-exponential) components, without regard to their placement. Alternatively, it can be a configuration of independently failing modules, where each module has either or both attribute and time-to-failure components. SPARCS-2 also has an improved super modularity feature. Modules with minimal-cut unreliabiliy calculations can be mixed with those having minimal-path reliability calculations. All output has been standardized to system reliability or probability of success, regardless of the form in which the input data is presented, and whatever the configuration of modules or elements within modules

  7. Monte carlo depletion analysis of SMART core by MCNAP code

    International Nuclear Information System (INIS)

    Jung, Jong Sung; Sim, Hyung Jin; Kim, Chang Hyo; Lee, Jung Chan; Ji, Sung Kyun

    2001-01-01

    Depletion an analysis of SMART, a small-sized advanced integral PWR under development by KAERI, is conducted using the Monte Carlo (MC) depletion analysis program, MCNAP. The results are compared with those of the CASMO-3/ MASTER nuclear analysis. The difference between MASTER and MCNAP on k eff prediction is observed about 600pcm at BOC, and becomes smaller as the core burnup increases. The maximum difference bet ween two predict ions on fuel assembly (FA) normalized power distribution is about 6.6% radially , and 14.5% axially but the differences are observed to lie within standard deviation of MC estimations

  8. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2009-08-15

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

  9. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  10. Nonlinear seismic analysis of a graphite reactor core

    International Nuclear Information System (INIS)

    Laframboise, W.L.; Desmond, T.P.

    1988-01-01

    Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

  11. Analysis of the Ford Nuclear Reactor LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1983-09-01

    This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

  12. Monte Carlo analysis of Musashi TRIGA mark II reactor core

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo

    1999-01-01

    The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k eff ) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k eff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

  13. Analysis of Homogeneous BFS-73-1 MA Benchmark Core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

    2007-06-15

    Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty.

  14. Analysis of Homogeneous BFS-73-1 MA Benchmark Core

    International Nuclear Information System (INIS)

    Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

    2007-06-01

    Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty

  15. Reliability and availability requirements analysis for DEMO: fuel cycle system

    International Nuclear Information System (INIS)

    Pinna, T.; Borgognoni, F.

    2015-01-01

    The Demonstration Power Plant (DEMO) will be a fusion reactor prototype designed to demonstrate the capability to produce electrical power in a commercially acceptable way. Two of the key elements of the engineering development of the DEMO reactor are the definitions of reliability and availability requirements (or targets). The availability target for a hypothesized Fuel Cycle has been analysed as a test case. The analysis has been done on the basis of the experience gained in operating existing tokamak fusion reactors and developing the ITER design. Plant Breakdown Structure (PBS) and Functional Breakdown Structure (FBS) related to the DEMO Fuel Cycle and correlations between PBS and FBS have been identified. At first, a set of availability targets has been allocated to the various systems on the basis of their operating, protection and safety functions. 75% and 85% of availability has been allocated to the operating functions of fuelling system and tritium plant respectively. 99% of availability has been allocated to the overall systems in executing their safety functions. The chances of the systems to achieve the allocated targets have then been investigated through a Failure Mode and Effect Analysis and Reliability Block Diagram analysis. The following results have been obtained: 1) the target of 75% for the operations of the fuelling system looks reasonable, while the target of 85% for the operations of the whole tritium plant should be reduced to 80%, even though all the tritium plant systems can individually reach quite high availability targets, over 90% - 95%; 2) all the DEMO Fuel Cycle systems can reach the target of 99% in accomplishing their safety functions. (authors)

  16. reliability reliability

    African Journals Online (AJOL)

    eobe

    Corresponding author, Tel: +234-703. RELIABILITY .... V , , given by the code of practice. However, checks must .... an optimization procedure over the failure domain F corresponding .... of Concrete Members based on Utility Theory,. Technical ...

  17. Reliability analysis of large scaled structures by optimization technique

    International Nuclear Information System (INIS)

    Ishikawa, N.; Mihara, T.; Iizuka, M.

    1987-01-01

    This paper presents a reliability analysis based on the optimization technique using PNET (Probabilistic Network Evaluation Technique) method for the highly redundant structures having a large number of collapse modes. This approach makes the best use of the merit of the optimization technique in which the idea of PNET method is used. The analytical process involves the minimization of safety index of the representative mode, subjected to satisfaction of the mechanism condition and of the positive external work. The procedure entails the sequential performance of a series of the NLP (Nonlinear Programming) problems, where the correlation condition as the idea of PNET method pertaining to the representative mode is taken as an additional constraint to the next analysis. Upon succeeding iterations, the final analysis is achieved when a collapse probability at the subsequent mode is extremely less than the value at the 1st mode. The approximate collapse probability of the structure is defined as the sum of the collapse probabilities of the representative modes classified by the extent of correlation. Then, in order to confirm the validity of the proposed method, the conventional Monte Carlo simulation is also revised by using the collapse load analysis. Finally, two fairly large structures were analyzed to illustrate the scope and application of the approach. (orig./HP)

  18. Application of human reliability analysis methodology of second generation

    International Nuclear Information System (INIS)

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

    2009-10-01

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

  19. Analogical reasoning for reliability analysis based on generic data

    Energy Technology Data Exchange (ETDEWEB)

    Kozin, Igor O

    1996-10-01

    The paper suggests using the systemic concept 'analogy' for the foundation of an approach to analyze system reliability on the basis of generic data, describing the method of structuring the set that defines analogical models, an approach of transition from the analogical model to a reliability model and a way of obtaining reliability intervals of analogous objects.

  20. Analogical reasoning for reliability analysis based on generic data

    International Nuclear Information System (INIS)

    Kozin, Igor O.

    1996-01-01

    The paper suggests using the systemic concept 'analogy' for the foundation of an approach to analyze system reliability on the basis of generic data, describing the method of structuring the set that defines analogical models, an approach of transition from the analogical model to a reliability model and a way of obtaining reliability intervals of analogous objects

  1. Applicability of simplified human reliability analysis methods for severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Boring, R.; St Germain, S. [Idaho National Lab., Idaho Falls, Idaho (United States); Banaseanu, G.; Chatri, H.; Akl, Y. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2016-03-15

    Most contemporary human reliability analysis (HRA) methods were created to analyse design-basis accidents at nuclear power plants. As part of a comprehensive expansion of risk assessments at many plants internationally, HRAs will begin considering severe accident scenarios. Severe accidents, while extremely rare, constitute high consequence events that significantly challenge successful operations and recovery. Challenges during severe accidents include degraded and hazardous operating conditions at the plant, the shift in control from the main control room to the technical support center, the unavailability of plant instrumentation, and the need to use different types of operating procedures. Such shifts in operations may also test key assumptions in existing HRA methods. This paper discusses key differences between design basis and severe accidents, reviews efforts to date to create customized HRA methods suitable for severe accidents, and recommends practices for adapting existing HRA methods that are already being used for HRAs at the plants. (author)

  2. Fatigue Reliability Analysis of Wind Turbine Cast Components

    DEFF Research Database (Denmark)

    Rafsanjani, Hesam Mirzaei; Sørensen, John Dalsgaard; Fæster, Søren

    2017-01-01

    .) and to quantify the relevant uncertainties using available fatigue tests. Illustrative results are presented as obtained by statistical analysis of a large set of fatigue data for casted test components typically used for wind turbines. Furthermore, the SN curves (fatigue life curves based on applied stress......The fatigue life of wind turbine cast components, such as the main shaft in a drivetrain, is generally determined by defects from the casting process. These defects may reduce the fatigue life and they are generally distributed randomly in components. The foundries, cutting facilities and test...... facilities can affect the verification of properties by testing. Hence, it is important to have a tool to identify which foundry, cutting and/or test facility produces components which, based on the relevant uncertainties, have the largest expected fatigue life or, alternatively, have the largest reliability...

  3. Basic aspects of stochastic reliability analysis for redundancy systems

    International Nuclear Information System (INIS)

    Doerre, P.

    1989-01-01

    Much confusion has been created by trying to establish common cause failure (CCF) as an extra phenomenon which has to be treated with extra methods in reliability and data analysis. This paper takes another approach which can be roughly described by the statement that dependent failure is the basic phenomenon, while 'independent failure' refers to a special limiting case, namely the perfectly homogeneous population. This approach is motivated by examples demonstrating that common causes do not lead to dependent failure, so far as physical dependencies like shared components are excluded, and that stochastic dependencies are not related to common causes. The possibility to select more than one failure behaviour from an inhomogeneous population is identified as an additional random process which creates stochastic dependence. However, this source of randomness is usually treated in the deterministic limit, which destroys dependence and hence yields incorrect multiple failure frequencies for redundancy structures, thus creating the need for applying corrective CCF models. (author)

  4. An Application of Graph Theory in Markov Chains Reliability Analysis

    Directory of Open Access Journals (Sweden)

    Pavel Skalny

    2014-01-01

    Full Text Available The paper presents reliability analysis which was realized for an industrial company. The aim of the paper is to present the usage of discrete time Markov chains and the flow in network approach. Discrete Markov chains a well-known method of stochastic modelling describes the issue. The method is suitable for many systems occurring in practice where we can easily distinguish various amount of states. Markov chains are used to describe transitions between the states of the process. The industrial process is described as a graph network. The maximal flow in the network corresponds to the production. The Ford-Fulkerson algorithm is used to quantify the production for each state. The combination of both methods are utilized to quantify the expected value of the amount of manufactured products for the given time period.

  5. Time-dependent reliability analysis and condition assessment of structures

    International Nuclear Information System (INIS)

    Ellingwood, B.R.

    1997-01-01

    Structures generally play a passive role in assurance of safety in nuclear plant operation, but are important if the plant is to withstand the effect of extreme environmental or abnormal events. Relative to mechanical and electrical components, structural systems and components would be difficult and costly to replace. While the performance of steel or reinforced concrete structures in service generally has been very good, their strengths may deteriorate during an extended service life as a result of changes brought on by an aggressive environment, excessive loading, or accidental loading. Quantitative tools for condition assessment of aging structures can be developed using time-dependent structural reliability analysis methods. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process

  6. Reliability analysis for the quench detection in the LHC machine

    CERN Document Server

    Denz, R; Vergara-Fernández, A

    2002-01-01

    The Large Hadron Collider (LHC) will incorporate a large amount of superconducting elements that require protection in case of a quench. Key elements in the quench protection system are the electronic quench detectors. Their reliability will have an important impact on the down time as well as on the operational cost of the collider. The expected rates of both false and missed quenches have been computed for several redundant detection schemes. The developed model takes account of the maintainability of the system to optimise the frequency of foreseen checks, and evaluate their influence on the performance of different detection topologies. Seen the uncertainty of the failure rate of the components combined with the LHC tunnel environment, the study has been completed with a sensitivity analysis of the results. The chosen detection scheme and the maintainability strategy for each detector family are given.

  7. Severe core damage experiments and analysis for CANDU applications

    International Nuclear Information System (INIS)

    Mathew, P.M.; White, A.J.; Snell, V.G.; Bonechi, M.

    2003-01-01

    AECL uses the MAAP CANDU code to calculate the progression of a severe core damage accident in a CANDU reactor to support Level 2 Probabilistic Safety Assessment and Severe Accident Management activities. Experimental data are required to ensure that the core damage models used in MAAP CANDU code are adequate. In SMiRT 16, details of single channel experiments were presented to elucidate the mechanisms of core debris formation. This paper presents the progress made in severe core damage experiments since then using single channels in an inert atmosphere and results of the model development work to support the experiments. The core disassembly experiments are conducted with one-fifth scale channels made of Zr-2.5wt%Nb containing twelve simulated fuel bundles in an inert atmosphere. The reference fuel channel geometry consists of a pressure tube/calandria tube composite, with the pressure tube ballooned into circumferential contact with the calandria tube. Experimental results from single channel tests showed the development of time-dependent sag when the reference channel temperature exceeded 850 degC. The test results also showed significant strain localization in the gap at the bundle junctions along the bottom side of the channel, thus suggesting creep to be the main deformation mechanism for debris formation. An ABAQUS finite element model using two-dimensional beam elements with circular cross-section was developed to explain the experimental findings. A comparison of the calculated central sag (at mid-span), the axial displacement at the free end of the channel and the post-test sag profile showed good agreement with the experiments, when strain localization was included in the model, suggesting such a simple modelling approach would be adequate to explain the test findings. The results of the tests are important not only in the context of the validation of the analytical tools and models adopted by AECL for the severe accident analysis of CANDU reactors but

  8. High enrichment to low enrichment core's conversion. Accidents analysis

    International Nuclear Information System (INIS)

    Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

    1990-01-01

    This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

  9. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  10. Analysis of forces on core structures during a loss-of-coolant accident. Final report

    International Nuclear Information System (INIS)

    Griggs, D.P.; Vilim, R.B.; Wang, C.H.; Meyer, J.E.

    1980-08-01

    There are several design requirements related to the emergency core cooling which would follow a hypothetical loss-of-coolant accident (LOCA). One of these requirements is that the core must retain a coolable geometry throughout the accident. A possible cause of core damage leading to an uncoolable geometry is the action of forces on the core and associated support structures during the very early (blowdown) stage of the LOCA. An equally unsatisfactory design result would occur if calculated deformations and failures were so extensive that the geometry used for calculating the next stages of the LOCA (refill and reflood) could not be known reasonably well. Subsidiary questions involve damage preventing the operation of control assemblies and loss of integrity of other needed safety systems. A reliable method of calculating these forces is therefore an important part of LOCA analysis. These concerns provided the motivation for the study. The general objective of the study was to review the state-of-the-art in LOCA force determination. Specific objectives were: (1) determine state-of-the-art by reviewing current (and projected near future) techniques for LOCA force determination, and (2) consider each of the major assumptions involved in force determination and make a qualitative assessment of their validity

  11. ERP Reliability Analysis (ERA) Toolbox: An open-source toolbox for analyzing the reliability of event-related brain potentials.

    Science.gov (United States)

    Clayson, Peter E; Miller, Gregory A

    2017-01-01

    Generalizability theory (G theory) provides a flexible, multifaceted approach to estimating score reliability. G theory's approach to estimating score reliability has important advantages over classical test theory that are relevant for research using event-related brain potentials (ERPs). For example, G theory does not require parallel forms (i.e., equal means, variances, and covariances), can handle unbalanced designs, and provides a single reliability estimate for designs with multiple sources of error. This monograph provides a detailed description of the conceptual framework of G theory using examples relevant to ERP researchers, presents the algorithms needed to estimate ERP score reliability, and provides a detailed walkthrough of newly-developed software, the ERP Reliability Analysis (ERA) Toolbox, that calculates score reliability using G theory. The ERA Toolbox is open-source, Matlab software that uses G theory to estimate the contribution of the number of trials retained for averaging, group, and/or event types on ERP score reliability. The toolbox facilitates the rigorous evaluation of psychometric properties of ERP scores recommended elsewhere in this special issue. Copyright © 2016 Elsevier B.V. All rights reserved.

  12. DEPEND-HRA-A method for consideration of dependency in human reliability analysis

    International Nuclear Information System (INIS)

    Cepin, Marko

    2008-01-01

    A consideration of dependencies between human actions is an important issue within the human reliability analysis. A method was developed, which integrates the features of existing methods and the experience from a full scope plant simulator. The method is used on real plant-specific human reliability analysis as a part of the probabilistic safety assessment of a nuclear power plant. The method distinguishes dependency for pre-initiator events from dependency for initiator and post-initiator events. The method identifies dependencies based on scenarios, where consecutive human actions are modeled, and based on a list of minimal cut sets, which is obtained by running the minimal cut set analysis considering high values of human error probabilities in the evaluation. A large example study, which consisted of a large number of human failure events, demonstrated the applicability of the method. Comparative analyses that were performed show that both selection of dependency method and selection of dependency levels within the method largely impact the results of probabilistic safety assessment. If the core damage frequency is not impacted much, the listings of important basic events in terms of risk increase and risk decrease factors may change considerably. More efforts are needed on the subject, which will prepare the background for more detailed guidelines, which will remove the subjectivity from the evaluations as much as it is possible

  13. A reliability analysis of the revised competitiveness index.

    Science.gov (United States)

    Harris, Paul B; Houston, John M

    2010-06-01

    This study examined the reliability of the Revised Competitiveness Index by investigating the test-retest reliability, interitem reliability, and factor structure of the measure based on a sample of 280 undergraduates (200 women, 80 men) ranging in age from 18 to 28 years (M = 20.1, SD = 2.1). The findings indicate that the Revised Competitiveness Index has high test-retest reliability, high inter-item reliability, and a stable factor structure. The results support the assertion that the Revised Competitiveness Index assesses competitiveness as a stable trait rather than a dynamic state.

  14. Core disruptive accident and recriticality analysis with FX2-POOL

    International Nuclear Information System (INIS)

    Abramson, P.B.

    1976-01-01

    The current state of development of FX2-POOL, a two-dimensional hydrodynamic, thermodynamic and neutronic scoping model for Hypothetical Core Disruptive Accident analysis is described. Checkout comparisons to VENUS for prompt burst conditions were good. Use of FX2-POOL to examine the importance of fuel to steel heat transfer during a prompt burst indicates that heat transfer plays no important role on that time scale. Scoping studies of material thermohydrodynamics for about 20 to 30 milliseconds following the prompt burst indicate that heat transfer is important on the time scale necessary for the CDA bubble to grow to the size of the original core. Preliminary results are presented for energetics of boiling fuel steel pools which are forced recritical by local surface pressurization

  15. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

    Han, S.H.; Yoo, K.J.; Kim, T.W.

    1987-01-01

    The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

  16. Human reliability analysis for advanced control room of KNGR

    International Nuclear Information System (INIS)

    Kim, Myung-Ro; Park, Seong-Kyu

    2000-01-01

    There are two purposes in Human Reliability Analysis (HRA) which was performed during Korean Next Generation Reactor (KNGR) Phase 2 research project. One is to present the human error probability quantification results for Probabilistic Safety Assessment (PSA) and the other is to provide a list of the critical operator actions for Human Factor Engineering (HFE). Critical operator actions were identified from the KNGR HRA/RSA based on selection criteria and incorporated in the MMI Task Analysis, where they receive additional treatment. The use of HRA/PSA results in design, procedure development, and training was ensured by their incorporation in the MMI task analysis and MCR design such as fixed position alarms, displays and controls. Any dominant PSA sequence that takes credit for human performance to achieve acceptable results was incorporated in MMIS validation activities through the PSA-based critical operator actions. The integration of KNGR HRA into MMI design was sufficiently addressed all applicable review criteria of NUREG-0800, Chapter 18, Section 2 F and NUREG-0711. (S.Y.)

  17. Models and data requirements for human reliability analysis

    International Nuclear Information System (INIS)

    1989-03-01

    It has been widely recognised for many years that the safety of the nuclear power generation depends heavily on the human factors related to plant operation. This has been confirmed by the accidents at Three Mile Island and Chernobyl. Both these cases revealed how human actions can defeat engineered safeguards and the need for special operator training to cover the possibility of unexpected plant conditions. The importance of the human factor also stands out in the analysis of abnormal events and insights from probabilistic safety assessments (PSA's), which reveal a large proportion of cases having their origin in faulty operator performance. A consultants' meeting, organized jointly by the International Atomic Energy Agency (IAEA) and the International Institute for Applied Systems Analysis (IIASA) was held at IIASA in Laxenburg, Austria, December 7-11, 1987, with the aim of reviewing existing models used in Probabilistic Safety Assessment (PSA) for Human Reliability Analysis (HRA) and of identifying the data required. The report collects both the contributions offered by the members of the Expert Task Force and the findings of the extensive discussions that took place during the meeting. Refs, figs and tabs

  18. Reliability analysis of component of affination centrifugal 1 machine by using reliability engineering

    Science.gov (United States)

    Sembiring, N.; Ginting, E.; Darnello, T.

    2017-12-01

    Problems that appear in a company that produces refined sugar, the production floor has not reached the level of critical machine availability because it often suffered damage (breakdown). This results in a sudden loss of production time and production opportunities. This problem can be solved by Reliability Engineering method where the statistical approach to historical damage data is performed to see the pattern of the distribution. The method can provide a value of reliability, rate of damage, and availability level, of an machine during the maintenance time interval schedule. The result of distribution test to time inter-damage data (MTTF) flexible hose component is lognormal distribution while component of teflon cone lifthing is weibull distribution. While from distribution test to mean time of improvement (MTTR) flexible hose component is exponential distribution while component of teflon cone lifthing is weibull distribution. The actual results of the flexible hose component on the replacement schedule per 720 hours obtained reliability of 0.2451 and availability 0.9960. While on the critical components of teflon cone lifthing actual on the replacement schedule per 1944 hours obtained reliability of 0.4083 and availability 0.9927.

  19. Bearing Procurement Analysis Method by Total Cost of Ownership Analysis and Reliability Prediction

    Science.gov (United States)

    Trusaji, Wildan; Akbar, Muhammad; Sukoyo; Irianto, Dradjad

    2018-03-01

    In making bearing procurement analysis, price and its reliability must be considered as decision criteria, since price determines the direct cost as acquisition cost and reliability of bearing determine the indirect cost such as maintenance cost. Despite the indirect cost is hard to identify and measured, it has high contribution to overall cost that will be incurred. So, the indirect cost of reliability must be considered when making bearing procurement analysis. This paper tries to explain bearing evaluation method with the total cost of ownership analysis to consider price and maintenance cost as decision criteria. Furthermore, since there is a lack of failure data when bearing evaluation phase is conducted, reliability prediction method is used to predict bearing reliability from its dynamic load rating parameter. With this method, bearing with a higher price but has higher reliability is preferable for long-term planning. But for short-term planning the cheaper one but has lower reliability is preferable. This contextuality can give rise to conflict between stakeholders. Thus, the planning horizon needs to be agreed by all stakeholder before making a procurement decision.

  20. A reliability analysis framework with Monte Carlo simulation for weld structure of crane's beam

    Science.gov (United States)

    Wang, Kefei; Xu, Hongwei; Qu, Fuzheng; Wang, Xin; Shi, Yanjun

    2018-04-01

    The reliability of the crane product in engineering is the core competitiveness of the product. This paper used Monte Carlo method analyzed the reliability of the weld metal structure of the bridge crane whose limit state function is mathematical expression. Then we obtained the minimum reliable welding feet height value for the welds between cover plate and web plate on main beam in different coefficients of variation. This paper provides a new idea and reference for the growth of the inherent reliability of crane.

  1. Reliability analysis of land pipelines for hydrocarbons transportation in Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Leon, D.; Cortes, C. [Inst. Mexicano del Petroleo (Mexico)

    2004-07-01

    The reliability of a land pipeline operated by PEMEX in Mexico was estimated under a range of failure modes. Reliability and safety were evaluated in terms of the pipeline's internal pressure, bending, fracture toughness and its tension failure mode conditions. Loading conditions were applied individually, while bending and tension loads were applied in a combined fashion. The mechanical properties of the steel were also considered along with the degradation effect of internal corrosion resulting from the type of product being transported. A set of internal pressures and mechanical properties were generated randomly using Monte Carlo simulation. Commercial software was used to obtain the pipeline response under each modeled condition. The response analysis was based on the nonlinear finite element method involving a calculation of maximum stresses and stress concentration factors under conditions of corrosion and no corrosion. The margin between maximum stresses due to internal pressure, tension and bending was evaluated along with the margin between stress concentration factor and fracture initiation toughness. The study showed that internal pressure, stress concentration and tension-bending were the critical failure modes. It was suggested that more research should be conducted to improve the modeling of the deteriorating effects of corrosion and to adjust the probability distribution for fracture toughness and the length/depth defect ratio. The consideration of welding geometries along with features of marine pipelines and pipeline products would help to generalize the study to facilitate the creation of codes for the construction, design, inspection and maintenance of pipelines in Mexico. 7 refs., 1 tab., 14 figs.

  2. Reliability Engineering

    International Nuclear Information System (INIS)

    Lee, Sang Yong

    1992-07-01

    This book is about reliability engineering, which describes definition and importance of reliability, development of reliability engineering, failure rate and failure probability density function about types of it, CFR and index distribution, IFR and normal distribution and Weibull distribution, maintainability and movability, reliability test and reliability assumption in index distribution type, normal distribution type and Weibull distribution type, reliability sampling test, reliability of system, design of reliability and functionality failure analysis by FTA.

  3. RELOSS, Reliability of Safety System by Fault Tree Analysis

    International Nuclear Information System (INIS)

    Allan, R.N.; Rondiris, I.L.; Adraktas, A.

    1981-01-01

    1 - Description of problem or function: Program RELOSS is used in the reliability/safety assessment of any complex system with predetermined operational logic in qualitative and (if required) quantitative terms. The program calculates the possible system outcomes following an abnormal operating condition and the probability of occurrence, if required. Furthermore, the program deduces the minimal cut or tie sets of the system outcomes and identifies the potential common mode failures. 4. Method of solution: The reliability analysis performed by the program is based on the event tree methodology. Using this methodology, the program develops the event tree of a system or a module of that system and relates each path of this tree to its qualitative and/or quantitative impact on specified system or module outcomes. If the system being analysed is subdivided into modules the program assesses each module in turn as described previously and then combines the module information to obtain results for the overall system. Having developed the event tree of a module or a system, the program identifies which paths lead or do not lead to various outcomes depending on whether the cut or the tie sets of the outcomes are required and deduces the corresponding sets. Furthermore the program identifies for a specific system outcome, the potential common mode failures and the cut or tie sets containing potential dependent failures of some components. 5. Restrictions on the complexity of the problem: The present dimensions of the program are as follows. They can however be easily modified: Maximum number of modules (equivalent components): 25; Maximum number of components in a module: 15; Maximum number of levels of parentheses in a logical statement: 10 Maximum number of system outcomes: 3; Maximum number of module outcomes: 2; Maximum number of points in time for which quantitative analysis is required: 5; Maximum order of any cut or tie set: 10; Maximum order of a cut or tie of any

  4. Neutronic analysis of LMFBRs during severe core disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1979-01-01

    A number of numerical experiments were performed to assess the validity of diffusion theory and various perturbation methods for calculating the reactivity state of a severely disrupted liquid metal cooled fast breeder reactor (LMFBR). The disrupted configurations correspond, in general, to phases through which an LMFBR core could pass during a core disruptive accident (CDA). Two-reactor models were chosen for this study, the two zone, homogeneous Clinch River Breeder Reactor and the Large Heterogeneous Reactor Design Study Core. The various phases were chosen to approximate the CDA results predicted by the safety analysis code SAS3D. The calculational methods investigated in this study include the eigenvalue difference technique based on both discrete ordinate transport theory and diffusion theory, first-order perturbation theory, exact perturbation theory, and a new hybrid perturbation theory. Selected cases were analyzed using Monte Carlo methods. It was found that in all cases, diffusion theory and perturbation theory yielded results for the change in reactivity that significantly disagreed with both the discrete ordinate and Monte Carlo results. These differences were, in most cases, in a nonconservative direction

  5. The reliability of mercury analysis in environmental materials

    International Nuclear Information System (INIS)

    Heinonen, J.; Suschny, O.

    1973-01-01

    Mercury occurs in nature in its native elemental as well as in different mineral forms. It has been mined for centuries and is used in many branches of industry, agriculture and medicine. Mercury is very toxic to man and reports of poisoning due to the presence of the element in fish and shellfish caught at Minamata and Niigata, Japan have led not only to local investigations but to multi-national research into the sources and the levels of mercury in the environment. The concentrations at which the element has to be determined in these studies are extremely small, usually of the order of a few parts in 10 9 parts of environmental material. Few analytical techniques provide the required sensitivity for analysis at such low concentrations, and only two are normally used for mercury: neutron activation analysis and atomic absorption photometry. They are also the most convenient end points of various separation schemes for different organic mercury compounds. Mercury analysis at the ppb-level is beset with many problems: volatility of the metal and its compounds, impurity of reagents, interference by other elements and many other analytical difficulties may influence the results. To be able to draw valid conclusions from the analyses it is necessary to know the reliability attached to the values obtained. To assist laboratories in the evaluation of their analytical performance, the International Atomic Energy Agency through its own laboratory at Seibersdorf already organised in 1967 an intercomparison of mercury analysis in flour. Based on the results obtained at that time, a whole series of intercomparisons of mercury determinations in nine different environmental materials was undertaken in 1971. The materials investigated included corn and wheat flour, spray-dried animal blood serum, fish solubles, milk powder, saw dust, cellulose, lacquer paint and coloric material

  6. The reliability of mercury analysis in environmental materials

    Energy Technology Data Exchange (ETDEWEB)

    Heinonen, J.; Suschny, O

    1973-01-01

    Mercury occurs in nature in its native elemental as well as in different mineral forms. It has been mined for centuries and is used in many branches of industry, agriculture and medicine. Mercury is very toxic to man and reports of poisoning due to the presence of the element in fish and shellfish caught at Minamata and Niigata, Japan have led not only to local investigations but to multi-national research into the sources and the levels of mercury in the environment. The concentrations at which the element has to be determined in these studies are extremely small, usually of the order of a few parts in 10{sup 9} parts of environmental material. Few analytical techniques provide the required sensitivity for analysis at such low concentrations, and only two are normally used for mercury: neutron activation analysis and atomic absorption photometry. They are also the most convenient end points of various separation schemes for different organic mercury compounds. Mercury analysis at the ppb-level is beset with many problems: volatility of the metal and its compounds, impurity of reagents, interference by other elements and many other analytical difficulties may influence the results. To be able to draw valid conclusions from the analyses it is necessary to know the reliability attached to the values obtained. To assist laboratories in the evaluation of their analytical performance, the International Atomic Energy Agency through its own laboratory at Seibersdorf already organised in 1967 an intercomparison of mercury analysis in flour. Based on the results obtained at that time, a whole series of intercomparisons of mercury determinations in nine different environmental materials was undertaken in 1971. The materials investigated included corn and wheat flour, spray-dried animal blood serum, fish solubles, milk powder, saw dust, cellulose, lacquer paint and coloric material.

  7. Multi-Unit Considerations for Human Reliability Analysis

    Energy Technology Data Exchange (ETDEWEB)

    St. Germain, S.; Boring, R.; Banaseanu, G.; Akl, Y.; Chatri, H.

    2017-03-01

    This paper uses the insights from the Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) methodology to help identify human actions currently modeled in the single unit PSA that may need to be modified to account for additional challenges imposed by a multi-unit accident as well as identify possible new human actions that might be modeled to more accurately characterize multi-unit risk. In identifying these potential human action impacts, the use of the SPAR-H strategy to include both errors in diagnosis and errors in action is considered as well as identifying characteristics of a multi-unit accident scenario that may impact the selection of the performance shaping factors (PSFs) used in SPAR-H. The lessons learned from the Fukushima Daiichi reactor accident will be addressed to further help identify areas where improved modeling may be required. While these multi-unit impacts may require modifications to a Level 1 PSA model, it is expected to have much more importance for Level 2 modeling. There is little currently written specifically about multi-unit HRA issues. A review of related published research will be presented. While this paper cannot answer all issues related to multi-unit HRA, it will hopefully serve as a starting point to generate discussion and spark additional ideas towards the proper treatment of HRA in a multi-unit PSA.

  8. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  9. 78 FR 45447 - Revisions to Modeling, Data, and Analysis Reliability Standard

    Science.gov (United States)

    2013-07-29

    ...; Order No. 782] Revisions to Modeling, Data, and Analysis Reliability Standard AGENCY: Federal Energy... Analysis (MOD) Reliability Standard MOD- 028-2, submitted to the Commission for approval by the North... Organization. The Commission finds that the proposed Reliability Standard represents an improvement over the...

  10. Field Programmable Gate Array Reliability Analysis Guidelines for Launch Vehicle Reliability Block Diagrams

    Science.gov (United States)

    Al Hassan, Mohammad; Britton, Paul; Hatfield, Glen Spencer; Novack, Steven D.

    2017-01-01

    Field Programmable Gate Arrays (FPGAs) integrated circuits (IC) are one of the key electronic components in today's sophisticated launch and space vehicle complex avionic systems, largely due to their superb reprogrammable and reconfigurable capabilities combined with relatively low non-recurring engineering costs (NRE) and short design cycle. Consequently, FPGAs are prevalent ICs in communication protocols and control signal commands. This paper will identify reliability concerns and high level guidelines to estimate FPGA total failure rates in a launch vehicle application. The paper will discuss hardware, hardware description language, and radiation induced failures. The hardware contribution of the approach accounts for physical failures of the IC. The hardware description language portion will discuss the high level FPGA programming languages and software/code reliability growth. The radiation portion will discuss FPGA susceptibility to space environment radiation.

  11. The reliability analysis of cutting tools in the HSM processes

    OpenAIRE

    W.S. Lin

    2008-01-01

    Purpose: This article mainly describe the reliability of the cutting tools in the high speed turning by normaldistribution model.Design/methodology/approach: A series of experimental tests have been done to evaluate the reliabilityvariation of the cutting tools. From experimental results, the tool wear distribution and the tool life are determined,and the tool life distribution and the reliability function of cutting tools are derived. Further, the reliability ofcutting tools at anytime for h...

  12. Adjoint sensitivity analysis of dynamic reliability models based on Markov chains - II: Application to IFMIF reliability assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cacuci, D. G. [Commiss Energy Atom, Direct Energy Nucl, Saclay, (France); Cacuci, D. G.; Balan, I. [Univ Karlsruhe, Inst Nucl Technol and Reactor Safetly, Karlsruhe, (Germany); Ionescu-Bujor, M. [Forschungszentrum Karlsruhe, Fus Program, D-76021 Karlsruhe, (Germany)

    2008-07-01

    In Part II of this work, the adjoint sensitivity analysis procedure developed in Part I is applied to perform sensitivity analysis of several dynamic reliability models of systems of increasing complexity, culminating with the consideration of the International Fusion Materials Irradiation Facility (IFMIF) accelerator system. Section II presents the main steps of a procedure for the automated generation of Markov chains for reliability analysis, including the abstraction of the physical system, construction of the Markov chain, and the generation and solution of the ensuing set of differential equations; all of these steps have been implemented in a stand-alone computer code system called QUEFT/MARKOMAG-S/MCADJSEN. This code system has been applied to sensitivity analysis of dynamic reliability measures for a paradigm '2-out-of-3' system comprising five components and also to a comprehensive dynamic reliability analysis of the IFMIF accelerator system facilities for the average availability and, respectively, the system's availability at the final mission time. The QUEFT/MARKOMAG-S/MCADJSEN has been used to efficiently compute sensitivities to 186 failure and repair rates characterizing components and subsystems of the first-level fault tree of the IFMIF accelerator system. (authors)

  13. Adjoint sensitivity analysis of dynamic reliability models based on Markov chains - II: Application to IFMIF reliability assessment

    International Nuclear Information System (INIS)

    Cacuci, D. G.; Cacuci, D. G.; Balan, I.; Ionescu-Bujor, M.

    2008-01-01

    In Part II of this work, the adjoint sensitivity analysis procedure developed in Part I is applied to perform sensitivity analysis of several dynamic reliability models of systems of increasing complexity, culminating with the consideration of the International Fusion Materials Irradiation Facility (IFMIF) accelerator system. Section II presents the main steps of a procedure for the automated generation of Markov chains for reliability analysis, including the abstraction of the physical system, construction of the Markov chain, and the generation and solution of the ensuing set of differential equations; all of these steps have been implemented in a stand-alone computer code system called QUEFT/MARKOMAG-S/MCADJSEN. This code system has been applied to sensitivity analysis of dynamic reliability measures for a paradigm '2-out-of-3' system comprising five components and also to a comprehensive dynamic reliability analysis of the IFMIF accelerator system facilities for the average availability and, respectively, the system's availability at the final mission time. The QUEFT/MARKOMAG-S/MCADJSEN has been used to efficiently compute sensitivities to 186 failure and repair rates characterizing components and subsystems of the first-level fault tree of the IFMIF accelerator system. (authors)

  14. Analysis of the plasma magnetohydrodynamic equilibrium in iron core transformer Tokamak HL-1M

    International Nuclear Information System (INIS)

    Chen Xiaoguang; Yuan Baoshan

    1992-01-01

    The physical and mathematical model are presented on the problem of MHD equilibrium with the self consistent in iron core transformer HL-1M. Calculation and analysis for the plasma equilibrium of the stable boundary and free boundary are shown respectively, in an axisymmetric equilibrium model of two dimensions. First, a variation formulation of the problem is written and the equations of the poloided flux ψ are solved by a finite element method; the Picard and Newton algorithms are tested for the non-linearities. The plasma boundary and the magnetic surfaces are being simulated, with the currents in the coils, the total plasma current, its current density function and the magnetic permeability of the iron being the data for the problem; a certain number of the characteristic parameter of the equilibrium configuration is calculated. Secondly, a simple method of calculation is adopted in the determination of equilibrium fields and currents in iron core HL-1M tokamak device. In the plasma equilibrium, the magnetic effect of the air gaps in the iron core and the iron magnetic shielded plate are considered in HL-1M device. Reliable data are provided for designing and constructing the poloidal field system of HL-1M device. A good computer code is constructed, which may be useful in operating on analysis for the future device

  15. Science Based Human Reliability Analysis: Using Digital Nuclear Power Plant Simulators for Human Reliability Research

    Science.gov (United States)

    Shirley, Rachel Elizabeth

    Nuclear power plant (NPP) simulators are proliferating in academic research institutions and national laboratories in response to the availability of affordable, digital simulator platforms. Accompanying the new research facilities is a renewed interest in using data collected in NPP simulators for Human Reliability Analysis (HRA) research. An experiment conducted in The Ohio State University (OSU) NPP Simulator Facility develops data collection methods and analytical tools to improve use of simulator data in HRA. In the pilot experiment, student operators respond to design basis accidents in the OSU NPP Simulator Facility. Thirty-three undergraduate and graduate engineering students participated in the research. Following each accident scenario, student operators completed a survey about perceived simulator biases and watched a video of the scenario. During the video, they periodically recorded their perceived strength of significant Performance Shaping Factors (PSFs) such as Stress. This dissertation reviews three aspects of simulator-based research using the data collected in the OSU NPP Simulator Facility: First, a qualitative comparison of student operator performance to computer simulations of expected operator performance generated by the Information Decision Action Crew (IDAC) HRA method. Areas of comparison include procedure steps, timing of operator actions, and PSFs. Second, development of a quantitative model of the simulator bias introduced by the simulator environment. Two types of bias are defined: Environmental Bias and Motivational Bias. This research examines Motivational Bias--that is, the effect of the simulator environment on an operator's motivations, goals, and priorities. A bias causal map is introduced to model motivational bias interactions in the OSU experiment. Data collected in the OSU NPP Simulator Facility are analyzed using Structural Equation Modeling (SEM). Data include crew characteristics, operator surveys, and time to recognize

  16. Reliability and risk analysis data base development: an historical perspective

    International Nuclear Information System (INIS)

    Fragola, Joseph R.

    1996-01-01

    Collection of empirical data and data base development for use in the prediction of the probability of future events has a long history. Dating back at least to the 17th century, safe passage events and mortality events were collected and analyzed to uncover prospective underlying classes and associated class attributes. Tabulations of these developed classes and associated attributes formed the underwriting basis for the fledgling insurance industry. Much earlier, master masons and architects used design rules of thumb to capture the experience of the ages and thereby produce structures of incredible longevity and reliability (Antona, E., Fragola, J. and Galvagni, R. Risk based decision analysis in design. Fourth SRA Europe Conference Proceedings, Rome, Italy, 18-20 October 1993). These rules served so well in producing robust designs that it was not until almost the 19th century that the analysis (Charlton, T.M., A History Of Theory Of Structures In The 19th Century, Cambridge University Press, Cambridge, UK, 1982) of masonry voussoir arches, begun by Galileo some two centuries earlier (Galilei, G. Discorsi e dimostrazioni mathematiche intorno a due nuove science, (Discourses and mathematical demonstrations concerning two new sciences, Leiden, The Netherlands, 1638), was placed on a sound scientific basis. Still, with the introduction of new materials (such as wrought iron and steel) and the lack of theoretical knowledge and computational facilities, approximate methods of structural design abounded well into the second half of the 20th century. To this day structural designers account for material variations and gaps in theoretical knowledge by employing factors of safety (Benvenuto, E., An Introduction to the History of Structural Mechanics, Part II: Vaulted Structures and Elastic Systems, Springer-Verlag, NY, 1991) or codes of practice (ASME Boiler and Pressure Vessel Code, ASME, New York) originally developed in the 19th century (Antona, E., Fragola, J. and

  17. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  18. Development of local TDC model in core thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Kwon, H.S.; Park, J.R.; Hwang, D.H.; Lee, S.K.

    2004-01-01

    The local TDC model consisting of natural mixing and forced mixing part was developed to obtain more realistic local fluid properties in the core subchannel analysis. To evaluate the performance of local TDC model, the CHF prediction capability was tested with the various CHF correlations and local fluid properties at CHF location which are based on the local TDC model. The results show that the standard deviation of measured to predicted CHF ratio (M/P) based on local TDC model can be reduced by about 7% compared to those based on global TDC model when the CHF correlation has no term to account for distance from the spacer grid. (author)

  19. Case for integral core-disruptive accident analysis

    International Nuclear Information System (INIS)

    Luck, L.B.; Bell, C.R.

    1985-01-01

    Integral analysis is an approach used at the Los Alamos National Laboratory to cope with the broad multiplicity of accident paths and complex phenomena that characterize the transition phase of core-disruptive accident progression in a liquid-metal-cooled fast breeder reactor. The approach is based on the combination of a reference calculation, which is intended to represent a band of similar accident paths, and associated system- and separate-effect studies, which are designed to determine the effect of uncertainties. Results are interpreted in the context of a probabilistic framework. The approach was applied successfully in two studies; illustrations from the Clinch River Breeder Reactor licensing assessment are included

  20. Design and analysis of PCRV core cavity closure

    International Nuclear Information System (INIS)

    Lee, T.T.; Schwartz, A.A.; Koopman, D.C.A.

    1980-05-01

    Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction

  1. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  2. Reliability analysis of the service water system of Angra 1 reactor

    International Nuclear Information System (INIS)

    Tayt-Sohn, L.C.; Oliveira, L.F.S. de.

    1984-01-01

    A reliability analysis of the service water system is done aiming to use in the evaluation of the non reliability of the Component Cooling System (SRC) for great loss of cooling accidents in nuclear power plants. (E.G.) [pt

  3. Reliability analysis of the service water system of Angra 1 reactor

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Fleming, P.V.; Frutuoso e Melo, P.F.F.; Tayt-Sohn, L.C.

    1983-01-01

    A reliability analysis of the service water system is done aiming to use in the evaluation of the non reliability of the component cooling system (SRC) for great loss of cooling accidents in nuclear power plants. (E.G.) [pt

  4. Human Reliability Analysis in Support of Risk Assessment for Positive Train Control

    Science.gov (United States)

    2003-06-01

    This report describes an approach to evaluating the reliability of human actions that are modeled in a probabilistic risk assessment : (PRA) of train control operations. This approach to human reliability analysis (HRA) has been applied in the case o...

  5. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    Nicolas, Anne

    1989-01-01

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated [fr

  6. Comparison of Methods for Dependency Determination between Human Failure Events within Human Reliability Analysis

    International Nuclear Information System (INIS)

    Cepin, M.

    2008-01-01

    The human reliability analysis (HRA) is a highly subjective evaluation of human performance, which is an input for probabilistic safety assessment, which deals with many parameters of high uncertainty. The objective of this paper is to show that subjectivism can have a large impact on human reliability results and consequently on probabilistic safety assessment results and applications. The objective is to identify the key features, which may decrease subjectivity of human reliability analysis. Human reliability methods are compared with focus on dependency comparison between Institute Jozef Stefan human reliability analysis (IJS-HRA) and standardized plant analysis risk human reliability analysis (SPAR-H). Results show large differences in the calculated human error probabilities for the same events within the same probabilistic safety assessment, which are the consequence of subjectivity. The subjectivity can be reduced by development of more detailed guidelines for human reliability analysis with many practical examples for all steps of the process of evaluation of human performance

  7. Comparison of methods for dependency determination between human failure events within human reliability analysis

    International Nuclear Information System (INIS)

    Cepis, M.

    2007-01-01

    The Human Reliability Analysis (HRA) is a highly subjective evaluation of human performance, which is an input for probabilistic safety assessment, which deals with many parameters of high uncertainty. The objective of this paper is to show that subjectivism can have a large impact on human reliability results and consequently on probabilistic safety assessment results and applications. The objective is to identify the key features, which may decrease of subjectivity of human reliability analysis. Human reliability methods are compared with focus on dependency comparison between Institute Jozef Stefan - Human Reliability Analysis (IJS-HRA) and Standardized Plant Analysis Risk Human Reliability Analysis (SPAR-H). Results show large differences in the calculated human error probabilities for the same events within the same probabilistic safety assessment, which are the consequence of subjectivity. The subjectivity can be reduced by development of more detailed guidelines for human reliability analysis with many practical examples for all steps of the process of evaluation of human performance. (author)

  8. Operator reliability analysis during NPP small break LOCA

    International Nuclear Information System (INIS)

    Zhang Jiong; Chen Shenglin

    1990-01-01

    To assess the human factor characteristic of a NPP main control room (MCR) design, the MCR operator reliability during a small break LOCA is analyzed, and some approaches for improving the MCR operator reliability are proposed based on the analyzing results

  9. Wind turbine reliability : a database and analysis approach.

    Energy Technology Data Exchange (ETDEWEB)

    Linsday, James (ARES Corporation); Briand, Daniel; Hill, Roger Ray; Stinebaugh, Jennifer A.; Benjamin, Allan S. (ARES Corporation)

    2008-02-01

    The US wind Industry has experienced remarkable growth since the turn of the century. At the same time, the physical size and electrical generation capabilities of wind turbines has also experienced remarkable growth. As the market continues to expand, and as wind generation continues to gain a significant share of the generation portfolio, the reliability of wind turbine technology becomes increasingly important. This report addresses how operations and maintenance costs are related to unreliability - that is the failures experienced by systems and components. Reliability tools are demonstrated, data needed to understand and catalog failure events is described, and practical wind turbine reliability models are illustrated, including preliminary results. This report also presents a continuing process of how to proceed with controlling industry requirements, needs, and expectations related to Reliability, Availability, Maintainability, and Safety. A simply stated goal of this process is to better understand and to improve the operable reliability of wind turbine installations.

  10. The effect of loss functions on empirical Bayes reliability analysis

    Directory of Open Access Journals (Sweden)

    Camara Vincent A. R.

    1998-01-01

    Full Text Available The aim of the present study is to investigate the sensitivity of empirical Bayes estimates of the reliability function with respect to changing of the loss function. In addition to applying some of the basic analytical results on empirical Bayes reliability obtained with the use of the “popular” squared error loss function, we shall derive some expressions corresponding to empirical Bayes reliability estimates obtained with the Higgins–Tsokos, the Harris and our proposed logarithmic loss functions. The concept of efficiency, along with the notion of integrated mean square error, will be used as a criterion to numerically compare our results. It is shown that empirical Bayes reliability functions are in general sensitive to the choice of the loss function, and that the squared error loss does not always yield the best empirical Bayes reliability estimate.

  11. Extending Failure Modes and Effects Analysis Approach for Reliability Analysis at the Software Architecture Design Level

    NARCIS (Netherlands)

    Sözer, Hasan; Tekinerdogan, B.; Aksit, Mehmet; de Lemos, Rogerio; Gacek, Cristina

    2007-01-01

    Several reliability engineering approaches have been proposed to identify and recover from failures. A well-known and mature approach is the Failure Mode and Effect Analysis (FMEA) method that is usually utilized together with Fault Tree Analysis (FTA) to analyze and diagnose the causes of failures.

  12. Human reliability analysis for steam generator feed-and-bleed accident in Bushehr NPP-1

    Energy Technology Data Exchange (ETDEWEB)

    Jafarian, Reza [Valiasr University of Rafsanjan, Rafsanjan, 28 (Iran, Islamic Republic of); Sepanloo, Kamran [Atomic Energy Organization of Iran (AEOI), external link End of North Karegar Av., Tehran 14155-1339 (Iran, Islamic Republic of)

    2006-07-01

    According to the incident/accident reports, unsuccessful implementation of steam generator feed-and-bleed procedure is one of the most important events in nuclear power plants operation which greatly contributes to the level of risk of the plants. Generally, the loss of all feed water pumps flow (as one of the precursors) results in failure to maintain adequate cooling of the reactor core unless the operating crew initiate and follow the feed-and-bleed procedure correctly and timely. In this paper, firstly, a Human Reliability Analysis (HRA) event tree is presented delineating the major human activities and errors in the implementation of the steam generator (SG) feed-and-bleed procedure following the loss of (both normal and emergency) water feed to four SGs of Bushehr Nuclear Power Plant Unit 1 (BNPP-1). Secondly, the graphical method of task analysis as a part of HRA is used as a means of delineating correct and incorrect human actions. To be used in the probabilistic risk assessment (PRA), the outputs of the HRA event trees are fed into the system event trees, functional event trees or system fault trees. As a part of a probabilistic risk assessment of BNPP-1 and to assess the reliability of control room operators, a human reliability analysis model is applied based on the THERP (Technique for Human Error Rate Prediction) technique. The THERP method is used in the form of event trees named as the probability tree diagrams. In this research the Human Reliability Analysis event tree is constructed based on the background information and assumptions made and on a similar NPP task analysis. It is done so because the BNPP-1 is not an operational nuclear power plant. Thirdly, based on NUREG/CR-1278 Handbook, a computer program has been developed in Visual Basic language and used to illustrate the major human activities and determination of error rates of operators in the course of the implementation of the steam generator feed-and-bleed procedure. Finally, total

  13. Human reliability analysis for steam generator feed-and-bleed accident in Bushehr NPP-1

    International Nuclear Information System (INIS)

    Jafarian, Reza; Sepanloo, Kamran

    2006-01-01

    According to the incident/accident reports, unsuccessful implementation of steam generator feed-and-bleed procedure is one of the most important events in nuclear power plants operation which greatly contributes to the level of risk of the plants. Generally, the loss of all feed water pumps flow (as one of the precursors) results in failure to maintain adequate cooling of the reactor core unless the operating crew initiate and follow the feed-and-bleed procedure correctly and timely. In this paper, firstly, a Human Reliability Analysis (HRA) event tree is presented delineating the major human activities and errors in the implementation of the steam generator (SG) feed-and-bleed procedure following the loss of (both normal and emergency) water feed to four SGs of Bushehr Nuclear Power Plant Unit 1 (BNPP-1). Secondly, the graphical method of task analysis as a part of HRA is used as a means of delineating correct and incorrect human actions. To be used in the probabilistic risk assessment (PRA), the outputs of the HRA event trees are fed into the system event trees, functional event trees or system fault trees. As a part of a probabilistic risk assessment of BNPP-1 and to assess the reliability of control room operators, a human reliability analysis model is applied based on the THERP (Technique for Human Error Rate Prediction) technique. The THERP method is used in the form of event trees named as the probability tree diagrams. In this research the Human Reliability Analysis event tree is constructed based on the background information and assumptions made and on a similar NPP task analysis. It is done so because the BNPP-1 is not an operational nuclear power plant. Thirdly, based on NUREG/CR-1278 Handbook, a computer program has been developed in Visual Basic language and used to illustrate the major human activities and determination of error rates of operators in the course of the implementation of the steam generator feed-and-bleed procedure. Finally, total

  14. Human Reliability Analysis for steam generator feed-and-bleed accident in Bushehr NPP-1

    International Nuclear Information System (INIS)

    Jafarian, R.; Sepanloo, K.

    2005-01-01

    According to the incident/accident reports, unsuccessful implementation of steam generator feed-and-bleed procedure is one of the most important events in nuclear power plants operation which greatly contributes to the level of risk of the plants. Generally, the loss of all feed water pumps flow (as one of the precursors) results in failure to maintain adequate cooling of the reactor core unless the operating crew initiate and follow the feed-and-bleed procedure correctly and timely. In this paper, firstly, a Human Reliability Analysis (HRA) event tree is presented delineating the major human activities and errors in the implementation of the steam generator (SG) feed-and-bleed procedure following the loss of (both normal and emergency) water feed to four SGs of Bushehr Nuclear Power Plant unit1 (BNPP-1). Secondly, the graphical method of task analysis as a part of HRA is used as a means of delineating correct and incorrect human actions. To be used in the probabilistic risk assessment (PRA), the outputs of the HRA event trees are fed into the system event trees, functional event trees or system fault trees. As a part of a probabilistic risk assessment of BNPP-1 and to assess the reliability of control room operators, a human reliability analysis model is applied based on the THERP (Technique for Human Error Rate Prediction) technique. The THERP method is used in the form of event trees named as the probability tree diagrams. In this research the Human Reliability Analysis event tree is constructed based on the background information and assumptions made and on a similar NPP task analysis. It is done so because the BNPP-1 is not an operational nuclear power plant. Thirdly, based on NUREG/CR-1278 Handbook, a computer program has been developed in Visual Basic language and used to illustrate the major human activities and determination of error rates of operators in the course of the implementation of the steam generator feed-and-bleed procedure. Finally, total

  15. Reliability analysis of water distribution systems under uncertainty

    International Nuclear Information System (INIS)

    Kansal, M.L.; Kumar, Arun; Sharma, P.B.

    1995-01-01

    In most of the developing countries, the Water Distribution Networks (WDN) are of intermittent type because of the shortage of safe drinking water. Failure of a pipeline(s) in such cases will cause not only the fall in one or more nodal heads but also the poor connectivity of source with various demand nodes of the system. Most of the previous works have used the two-step algorithm based on pathset or cutset approach for connectivity analysis. The computations become more cumbersome when connectivity of all demand nodes taken together with that of supply is carried out. In the present paper, network connectivity based on the concept of Appended Spanning Tree (AST) is suggested to compute global network connectivity which is defined as the probability of the source node being connected with all the demand nodes simultaneously. The concept of AST has distinct advantages as it attacks the problem directly rather than in an indirect way as most of the studies so far have done. Since the water distribution system is a repairable one, a general expression for pipeline avialability using the failure/repair rate is considered. Furthermore, the sensitivity of global reliability estimates due to the likely error in the estimation of failure/repair rates of various pipelines is also studied

  16. CRITICAL ANALYSIS OF THE RELIABILITY OF INTUITIVE MORAL DECISIONS

    Directory of Open Access Journals (Sweden)

    V. V. Nadurak

    2017-06-01

    Full Text Available Purpose of the research is a critical analysis of the reliability of intuitive moral decisions. Methodology. The work is based on the methodological attitude of empirical ethics, involving the use of findings from empirical research in ethical reflection and decision making. Originality. The main kinds of intuitive moral decisions are identified: 1 intuitively emotional decisions (i.e. decisions made under the influence of emotions that accompanies the process of moral decision making; 2 decisions made under the influence of moral risky psychological aptitudes (unconscious human tendencies that makes us think in a certain way and make decisions, unacceptable from the logical and ethical point of view; 3 intuitively normative decisions (decisions made under the influence of socially learned norms, that cause evaluative feeling «good-bad», without conscious reasoning. It was found that all of these kinds of intuitive moral decisions can lead to mistakes in the moral life. Conclusions. Considering the fact that intuition systematically leads to erroneous moral decisions, intuitive reaction cannot be the only source for making such decisions. The conscious rational reasoning can compensate for weaknesses of intuition. In this case, there is a necessity in theoretical model that would structure the knowledge about the interactions between intuitive and rational factors in moral decisions making and became the basis for making suggestions that would help us to make the right moral decision.

  17. Monte Carlo methods for the reliability analysis of Markov systems

    International Nuclear Information System (INIS)

    Buslik, A.J.

    1985-01-01

    This paper presents Monte Carlo methods for the reliability analysis of Markov systems. Markov models are useful in treating dependencies between components. The present paper shows how the adjoint Monte Carlo method for the continuous time Markov process can be derived from the method for the discrete-time Markov process by a limiting process. The straightforward extensions to the treatment of mean unavailability (over a time interval) are given. System unavailabilities can also be estimated; this is done by making the system failed states absorbing, and not permitting repair from them. A forward Monte Carlo method is presented in which the weighting functions are related to the adjoint function. In particular, if the exact adjoint function is known then weighting factors can be constructed such that the exact answer can be obtained with a single Monte Carlo trial. Of course, if the exact adjoint function is known, there is no need to perform the Monte Carlo calculation. However, the formulation is useful since it gives insight into choices of the weight factors which will reduce the variance of the estimator

  18. Reliability analysis of load-sharing systems with memory.

    Science.gov (United States)

    Wang, Dewei; Jiang, Chendi; Park, Chanseok

    2018-02-22

    The load-sharing model has been studied since the early 1940s to account for the stochastic dependence of components in a parallel system. It assumes that, as components fail one by one, the total workload applied to the system is shared by the remaining components and thus affects their performance. Such dependent systems have been studied in many engineering applications which include but are not limited to fiber composites, manufacturing, power plants, workload analysis of computing, software and hardware reliability, etc. Many statistical models have been proposed to analyze the impact of each redistribution of the workload; i.e., the changes on the hazard rate of each remaining component. However, they do not consider how long a surviving component has worked for prior to the redistribution. We name such load-sharing models as memoryless. To remedy this potential limitation, we propose a general framework for load-sharing models that account for the work history. Through simulation studies, we show that an inappropriate use of the memoryless assumption could lead to inaccurate inference on the impact of redistribution. Further, a real-data example of plasma display devices is analyzed to illustrate our methods.

  19. The Test-Retest Reliability OfTthe Onset Of Core And Vasti Eectromyographic Activity While Ascending And Descending Stairs In Healthy Controls Aand patellofemoral Pain Patients

    Directory of Open Access Journals (Sweden)

    Mohammad-Ali Sanjari

    2011-02-01

    Full Text Available Backgroundentity.It is hypothesized to result from abnormal patellar tracking caused by altered motorcontrol. Deficit in neuromotor control of the core may be a remote contributing factor to thedevelopment of PFP. Application of reliable EMG measures would be helpful to handle thistheory. Therefore, the purpose of this study was to determine the test-retest reliability of thecore and vasti EMG onsets, while ascending/descending stairs.: Patellofemoral pain (PFP is a common affliction and complex clinicalMethodsand Core EMG onsets during stair stepping were assessed two times a day. Intraclass correlationcoefficients (ICCs and standard errors of measurement (SEMs were calculated.: Ten males with PFP and ten healthy controls participated in this study. VastiResultsonsets of control cases (ICC 3,1 ≥ 0.70 except Quadratus Lumborum (QL which showeda moderate reliability (ICC for ascending=0.59 and for descending = 0.61. In controls,Vasti in both tasks showed the highest absolute reliability. During ascending, highreliability (ICC ≥ 0.70 in PFP group was demonstrated for all EMG onsets except Gluteusmaximus (GMAX and QL which showed a moderate reliability (ICC = 0.69 and 0.63 respectively.In this group while descending stairs, all EMG onsets showed high relativereliability (ICC ≥ 0.70. Moderate to high absolute reliability was obtained for onset timeswhile ascending/descending stairs in PFP group.: During both ascending/descending, high reliability was found for all EMGConclusionreliability.: Most EMG onsets during stair scending/descending had moderate to high

  20. Procedure for conducting a human-reliability analysis for nuclear power plants. Final report

    International Nuclear Information System (INIS)

    Bell, B.J.; Swain, A.D.

    1983-05-01

    This document describes in detail a procedure to be followed in conducting a human reliability analysis as part of a probabilistic risk assessment when such an analysis is performed according to the methods described in NUREG/CR-1278, Handbook for Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications. An overview of the procedure describing the major elements of a human reliability analysis is presented along with a detailed description of each element and an example of an actual analysis. An appendix consists of some sample human reliability analysis problems for further study

  1. Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

    International Nuclear Information System (INIS)

    Kalchev, B.

    1995-01-01

    The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E -3 , 1/RY to 1.07E -3 , 1/RY. 5 refs., 2 tabs., 5 figs

  2. Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalchev, B [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E{sup -3}, 1/RY to 1.07E{sup -3}, 1/RY. 5 refs., 2 tabs., 5 figs.

  3. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

    Armstrong, B.C.; Buechel, R.J.

    1990-01-01

    Control rod repositioning is a method for minimizing rod cluster control assembly (RCCA) wear in the upper internals area where the guide cards interface with the rodlets of the RCCAs. A number of utilities have implemented strategies for rod repositioning, which often has no impact on the nuclear analysis for cases where the control rods are never repositioned into the active fuel. Other strategies involve repositioning the control rods several steps into the active fuel. The impact of this type of repositioning on the axial power shape and consequently the total peaking factor F Q T varies, depending on the method in which the repositioning is implemented at the plant. Operating for long periods with all the control and shutdown rods inserted several steps in the active fuel followed by withdrawing them fully from the core results in a shifting of the power distribution toward the top of the core and must be accounted for in the design analysis. On the other hand, an optional plan for control rod repositioning that considers margins available in related design parameters can be devised that minimizes the effects of the repositioning for the reload. This paper summarizes a rod repositioning strategy implemented for a recent reload and some calculated power shape results for this strategy and other scenarios

  4. Analysis the Response Function of the HTR Ex-core Neutron Detectors in Different Core Status

    International Nuclear Information System (INIS)

    Fan Kai; Li Fu; Zhou Xuhua

    2014-01-01

    Modular high temperature gas cooled reactor HTR-PM demonstration plant, designed by INET, Tsinghua University, is being built in Shidao Bay, Shandong province, China. HTR-PM adopts pebble bed concept. The harmonic synthesis method has been developed to reconstruct the power distributions on HTR-PM. The method based on the assumption that the neutron detector readings are mainly determined by the status of the core through the power distribution, and the response functions changed little when the status of the core changed. To verify the assumption, the influence factors to the ex-core neutron detectors are calculated in this paper, including the control rod position and the temperature of the core. The results shows that when the status of the core changed, the power distribution changed more remarkable than the response function, but the detector readings could change about 5% because of the response function changing. (author)

  5. Establishing guidance for the review of human reliability analysis in PSA

    International Nuclear Information System (INIS)

    Reer, B.; Dang, V.N.; Hirschberg, S.; Meyer, P.

    2000-01-01

    PSI was commissioned to develop Guidelines for the Regulatory Review of the Human Reliability Analysis (HRA) within Probabilistic Safety Assessments (PSAs) for nuclear power plants. In the Guidelines, HRA quality is addressed in terms of 97 indicators. Each indicator is formulated as a question, described as a specific feature of the analysis, and then explained in detail. Two analysis stages are distinguished: the selection of the human errors to be modelled, and their quantification to determine their impact on the core damage frequency. Review findings are grouped under two headings: transparency and adequacy. An analysis is 'transparent' if an externally qualified person is able to reproduce the analysis results, and 'adequate' if such results reflect the plant-specific conditions related to safety. To allocate resources efficiently, the review is structured in two phases: (1) The Quick Review, which clarifies whether the HRA has a fundamental deficiency and, furthermore, if it points to information needs and areas of emphasis for the detailed review, and (2) The Detailed Review, which results in well-grounded findings, based on extended examinations and close-plant contacts. (authors)

  6. Spinal appearance questionnaire: factor analysis, scoring, reliability, and validity testing.

    Science.gov (United States)

    Carreon, Leah Y; Sanders, James O; Polly, David W; Sucato, Daniel J; Parent, Stefan; Roy-Beaudry, Marjolaine; Hopkins, Jeffrey; McClung, Anna; Bratcher, Kelly R; Diamond, Beverly E

    2011-08-15

    Cross sectional. This study presents the factor analysis of the Spinal Appearance Questionnaire (SAQ) and its psychometric properties. Although the SAQ has been administered to a large sample of patients with adolescent idiopathic scoliosis (AIS) treated surgically, its psychometric properties have not been fully evaluated. This study presents the factor analysis and scoring of the SAQ and evaluates its psychometric properties. The SAQ and the Scoliosis Research Society-22 (SRS-22) were administered to AIS patients who were being observed, braced or scheduled for surgery. Standard demographic data and radiographic measures including Lenke type and curve magnitude were also collected. Of the 1802 patients, 83% were female; with a mean age of 14.8 years and mean initial Cobb angle of 55.8° (range, 0°-123°). From the 32 items of the SAQ, 15 loaded on two factors with consistent and significant correlations across all Lenke types. There is an Appearance (items 1-10) and an Expectations factor (items 12-15). Responses are summed giving a range of 5 to 50 for the Appearance domain and 5 to 20 for the Expectations domain. The Cronbach's α was 0.88 for both domains and Total score with a test-retest reliability of 0.81 for Appearance and 0.91 for Expectations. Correlations with major curve magnitude were higher for the SAQ Appearance and SAQ Total scores compared to correlations between the SRS Appearance and SRS Total scores. The SAQ and SRS-22 Scores were statistically significantly different in patients who were scheduled for surgery compared to those who were observed or braced. The SAQ is a valid measure of self-image in patients with AIS with greater correlation to curve magnitude than SRS Appearance and Total score. It also discriminates between patients who require surgery from those who do not.

  7. An efficient phased mission reliability analysis for autonomous vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Remenyte-Prescott, R., E-mail: R.Remenyte-Prescott@nottingham.ac.u [Nottingham Transportation Engineering Centre, Faculty of Engineering, University of Nottingham, Nottingham NG7 2RD (United Kingdom); Andrews, J.D. [Nottingham Transportation Engineering Centre, Faculty of Engineering, University of Nottingham, Nottingham NG7 2RD (United Kingdom); Chung, P.W.H. [Department of Computer Science, Loughborough University, Loughborough LE11 3TU (United Kingdom)

    2010-03-15

    Autonomous systems are becoming more commonly used, especially in hazardous situations. Such systems are expected to make their own decisions about future actions when some capabilities degrade due to failures of their subsystems. Such decisions are made without human input, therefore they need to be well-informed in a short time when the situation is analysed and future consequences of the failure are estimated. The future planning of the mission should take account of the likelihood of mission failure. The reliability analysis for autonomous systems can be performed using the methodologies developed for phased mission analysis, where the causes of failure for each phase in the mission can be expressed by fault trees. Unmanned autonomous vehicles (UAVs) are of a particular interest in the aeronautical industry, where it is a long term ambition to operate them routinely in civil airspace. Safety is the main requirement for the UAV operation and the calculation of failure probability of each phase and the overall mission is the topic of this paper. When components or subsystems fail or environmental conditions throughout the mission change, these changes can affect the future mission. The new proposed methodology takes into account the available diagnostics data and is used to predict future capabilities of the UAV in real time. Since this methodology is based on the efficient BDD method, the quickly provided advice can be used in making decisions. When failures occur appropriate actions are required in order to preserve safety of the autonomous vehicle. The overall decision making strategy for autonomous vehicles is explained in this paper. Some limitations of the methodology are discussed and further improvements are presented based on experimental results.

  8. An efficient phased mission reliability analysis for autonomous vehicles

    International Nuclear Information System (INIS)

    Remenyte-Prescott, R.; Andrews, J.D.; Chung, P.W.H.

    2010-01-01

    Autonomous systems are becoming more commonly used, especially in hazardous situations. Such systems are expected to make their own decisions about future actions when some capabilities degrade due to failures of their subsystems. Such decisions are made without human input, therefore they need to be well-informed in a short time when the situation is analysed and future consequences of the failure are estimated. The future planning of the mission should take account of the likelihood of mission failure. The reliability analysis for autonomous systems can be performed using the methodologies developed for phased mission analysis, where the causes of failure for each phase in the mission can be expressed by fault trees. Unmanned autonomous vehicles (UAVs) are of a particular interest in the aeronautical industry, where it is a long term ambition to operate them routinely in civil airspace. Safety is the main requirement for the UAV operation and the calculation of failure probability of each phase and the overall mission is the topic of this paper. When components or subsystems fail or environmental conditions throughout the mission change, these changes can affect the future mission. The new proposed methodology takes into account the available diagnostics data and is used to predict future capabilities of the UAV in real time. Since this methodology is based on the efficient BDD method, the quickly provided advice can be used in making decisions. When failures occur appropriate actions are required in order to preserve safety of the autonomous vehicle. The overall decision making strategy for autonomous vehicles is explained in this paper. Some limitations of the methodology are discussed and further improvements are presented based on experimental results.

  9. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  10. Distribution-level electricity reliability: Temporal trends using statistical analysis

    International Nuclear Information System (INIS)

    Eto, Joseph H.; LaCommare, Kristina H.; Larsen, Peter; Todd, Annika; Fisher, Emily

    2012-01-01

    This paper helps to address the lack of comprehensive, national-scale information on the reliability of the U.S. electric power system by assessing trends in U.S. electricity reliability based on the information reported by the electric utilities on power interruptions experienced by their customers. The research analyzes up to 10 years of electricity reliability information collected from 155 U.S. electric utilities, which together account for roughly 50% of total U.S. electricity sales. We find that reported annual average duration and annual average frequency of power interruptions have been increasing over time at a rate of approximately 2% annually. We find that, independent of this trend, installation or upgrade of an automated outage management system is correlated with an increase in the reported annual average duration of power interruptions. We also find that reliance on IEEE Standard 1366-2003 is correlated with higher reported reliability compared to reported reliability not using the IEEE standard. However, we caution that we cannot attribute reliance on the IEEE standard as having caused or led to higher reported reliability because we could not separate the effect of reliance on the IEEE standard from other utility-specific factors that may be correlated with reliance on the IEEE standard. - Highlights: ► We assess trends in electricity reliability based on the information reported by the electric utilities. ► We use rigorous statistical techniques to account for utility-specific differences. ► We find modest declines in reliability analyzing interruption duration and frequency experienced by utility customers. ► Installation or upgrade of an OMS is correlated to an increase in reported duration of power interruptions. ► We find reliance in IEEE Standard 1366 is correlated with higher reported reliability.

  11. Developing engineering design core competences through analysis of industrial products

    DEFF Research Database (Denmark)

    Hansen, Claus Thorp; Lenau, Torben Anker

    2011-01-01

    Most product development work carried out in industrial practice is characterised by being incremental, i.e. the industrial company has had a product in production and on the market for some time, and now time has come to design a new and upgraded variant. This type of redesign project requires...... that the engineering designers have core design competences to carry through an analysis of the existing product encompassing both a user-oriented side and a technical side, as well as to synthesise solution proposals for the new and upgraded product. The authors of this paper see an educational challenge in staging...... a course module, in which students develop knowledge, understanding and skills, which will prepare them for being able to participate in and contribute to redesign projects in industrial practice. In the course module Product Analysis and Redesign that has run for 8 years we have developed and refined...

  12. Modeling cognition dynamics and its application to human reliability analysis

    International Nuclear Information System (INIS)

    Mosleh, A.; Smidts, C.; Shen, S.H.

    1996-01-01

    For the past two decades, a number of approaches have been proposed for the identification and estimation of the likelihood of human errors, particularly for use in the risk and reliability studies of nuclear power plants. Despite the wide-spread use of the most popular among these methods, their fundamental weaknesses are widely recognized, and the treatment of human reliability has been considered as one of the soft spots of risk studies of large technological systems. To alleviate the situation, new efforts have focused on the development of human reliability models based on a more fundamental understanding of operator response and its cognitive aspects

  13. The effect of loss functions on empirical Bayes reliability analysis

    Directory of Open Access Journals (Sweden)

    Vincent A. R. Camara

    1999-01-01

    Full Text Available The aim of the present study is to investigate the sensitivity of empirical Bayes estimates of the reliability function with respect to changing of the loss function. In addition to applying some of the basic analytical results on empirical Bayes reliability obtained with the use of the “popular” squared error loss function, we shall derive some expressions corresponding to empirical Bayes reliability estimates obtained with the Higgins–Tsokos, the Harris and our proposed logarithmic loss functions. The concept of efficiency, along with the notion of integrated mean square error, will be used as a criterion to numerically compare our results.

  14. Qualification testing program plan for SIMMER. A computer code for LMFBR disrupted core analysis

    International Nuclear Information System (INIS)

    Basdekas, D.L.; Silberberg, M.; Curtis, R.T.; Kelber, C.N.

    1978-07-01

    The objective of SIMMER qualification testing program is to assure that the mathematical models and input parameters are derived from experimental data, which, on the basis of criteria still to be established, are representative of the phenomena and processes governing the progression of a CDA in an LMFBR. At the present time, the work to meet this objective can be classified into three general task areas as they relate to the use of SIMMER in CDA analysis: (1) The whole-core energetic disassembly accident, or the ''vessel problem'': The objective here is to predict the partition of the total energy release, by a postulated severe power excursion, between the primary containment and the energy absorbed through nondestructive dissipative processes. (2) Single subassembly accident: The objective here is to determine the pertinent phenomena and to develop the capability to assess the significance and likelihood that such accidents might propagate to involvement of larger fraction of the core. (3) The whole-core transition phase accident: The objective here is to advance the understanding of the phenomena and processes involved, so that reliable predictions can be made of the possible consequences of a CDA and the potential for further nuclear excursions through recriticality

  15. Cooperative Strategies for Maximum-Flow Problem in Uncertain Decentralized Systems Using Reliability Analysis

    Directory of Open Access Journals (Sweden)

    Hadi Heidari Gharehbolagh

    2016-01-01

    Full Text Available This study investigates a multiowner maximum-flow network problem, which suffers from risky events. Uncertain conditions effect on proper estimation and ignoring them may mislead decision makers by overestimation. A key question is how self-governing owners in the network can cooperate with each other to maintain a reliable flow. Hence, the question is answered by providing a mathematical programming model based on applying the triangular reliability function in the decentralized networks. The proposed method concentrates on multiowner networks which suffer from risky time, cost, and capacity parameters for each network’s arcs. Some cooperative game methods such as τ-value, Shapley, and core center are presented to fairly distribute extra profit of cooperation. A numerical example including sensitivity analysis and the results of comparisons are presented. Indeed, the proposed method provides more reality in decision-making for risky systems, hence leading to significant profits in terms of real cost estimation when compared with unforeseen effects.

  16. Reliability analysis of reinforced concrete grids with nonlinear material behavior

    Energy Technology Data Exchange (ETDEWEB)

    Neves, Rodrigo A [EESC-USP, Av. Trabalhador Sao Carlense, 400, 13566-590 Sao Carlos (Brazil); Chateauneuf, Alaa [LaMI-UBP and IFMA, Campus de Clermont-Fd, Les Cezeaux, BP 265, 63175 Aubiere cedex (France)]. E-mail: alaa.chateauneuf@ifma.fr; Venturini, Wilson S [EESC-USP, Av. Trabalhador Sao Carlense, 400, 13566-590 Sao Carlos (Brazil)]. E-mail: venturin@sc.usp.br; Lemaire, Maurice [LaMI-UBP and IFMA, Campus de Clermont-Fd, Les Cezeaux, BP 265, 63175 Aubiere cedex (France)

    2006-06-15

    Reinforced concrete grids are usually used to support large floor slabs. These grids are characterized by a great number of critical cross-sections, where the overall failure is usually sudden. However, nonlinear behavior of concrete leads to the redistribution of internal forces and accurate reliability assessment becomes mandatory. This paper presents a reliability study on reinforced concrete (RC) grids based on coupling Monte Carlo simulations with the response surface techniques. This approach allows us to analyze real RC grids with large number of failure components. The response surface is used to evaluate the structural safety by using first order reliability methods. The application to simple grids shows the interest of the proposed method and the role of moment redistribution in the reliability assessment.

  17. Analysis of travel time reliability on Indiana interstates.

    Science.gov (United States)

    2009-09-15

    Travel-time reliability is a key performance measure in any transportation system. It is a : measure of quality of travel time experienced by transportation system users and reflects the efficiency : of the transportation system to serve citizens, bu...

  18. Development of a Conservative Model Validation Approach for Reliable Analysis

    Science.gov (United States)

    2015-01-01

    CIE 2015 August 2-5, 2015, Boston, Massachusetts, USA [DRAFT] DETC2015-46982 DEVELOPMENT OF A CONSERVATIVE MODEL VALIDATION APPROACH FOR RELIABLE...obtain a conservative simulation model for reliable design even with limited experimental data. Very little research has taken into account the...3, the proposed conservative model validation is briefly compared to the conventional model validation approach. Section 4 describes how to account

  19. Statistical analysis of dynamic parameters of the core

    International Nuclear Information System (INIS)

    Ionov, V.S.

    2007-01-01

    The transients of various types were investigated for the cores of zero power critical facilities in RRC KI and NPP. Dynamic parameters of neutron transients were explored by tool statistical analysis. Its have sufficient duration, few channels for currents of chambers and reactivity and also some channels for technological parameters. On these values the inverse period. reactivity, lifetime of neutrons, reactivity coefficients and some effects of a reactivity are determinate, and on the values were restored values of measured dynamic parameters as result of the analysis. The mathematical means of statistical analysis were used: approximation(A), filtration (F), rejection (R), estimation of parameters of descriptive statistic (DSP), correlation performances (kk), regression analysis(KP), the prognosis (P), statistician criteria (SC). The calculation procedures were realized by computer language MATLAB. The reasons of methodical and statistical errors are submitted: inadequacy of model operation, precision neutron-physical parameters, features of registered processes, used mathematical model in reactivity meters, technique of processing for registered data etc. Examples of results of statistical analysis. Problems of validity of the methods used for definition and certification of values of statistical parameters and dynamic characteristics are considered (Authors)

  20. CARES/PC - CERAMICS ANALYSIS AND RELIABILITY EVALUATION OF STRUCTURES

    Science.gov (United States)

    Szatmary, S. A.

    1994-01-01

    The beneficial properties of structural ceramics include their high-temperature strength, light weight, hardness, and corrosion and oxidation resistance. For advanced heat engines, ceramics have demonstrated functional abilities at temperatures well beyond the operational limits of metals. This is offset by the fact that ceramic materials tend to be brittle. When a load is applied, their lack of significant plastic deformation causes the material to crack at microscopic flaws, destroying the component. CARES/PC performs statistical analysis of data obtained from the fracture of simple, uniaxial tensile or flexural specimens and estimates the Weibull and Batdorf material parameters from this data. CARES/PC is a subset of the program CARES (COSMIC program number LEW-15168) which calculates the fast-fracture reliability or failure probability of ceramic components utilizing the Batdorf and Weibull models to describe the effects of multi-axial stress states on material strength. CARES additionally requires that the ceramic structure be modeled by a finite element program such as MSC/NASTRAN or ANSYS. The more limited CARES/PC does not perform fast-fracture reliability estimation of components. CARES/PC estimates ceramic material properties from uniaxial tensile or from three- and four-point bend bar data. In general, the parameters are obtained from the fracture stresses of many specimens (30 or more are recommended) whose geometry and loading configurations are held constant. Parameter estimation can be performed for single or multiple failure modes by using the least-squares analysis or the maximum likelihood method. Kolmogorov-Smirnov and Anderson-Darling goodness-of-fit tests measure the accuracy of the hypothesis that the fracture data comes from a population with a distribution specified by the estimated Weibull parameters. Ninety-percent confidence intervals on the Weibull parameters and the unbiased value of the shape parameter for complete samples are provided

  1. Measuring health-related quality of life in children with cancer living in mainland China: feasibility, reliability and validity of the Chinese mandarin version of PedsQL 4.0 Generic Core Scales and 3.0 Cancer Module

    Directory of Open Access Journals (Sweden)

    Ji Yi

    2011-11-01

    Full Text Available Abstract Background The Pediatric Quality of Life Inventory (PedsQL is widely used instrument to measure pediatric health-related quality of life (HRQOL for children aged 2 to 18 years. The purpose of the current study was to investigate the feasibility, reliability and validity of the Chinese mandarin version of the PedsQL 4.0 Generic Core Scales and 3.0 Cancer Module in a group of Chinese children with cancer. Methods The PedsQL 4.0 Genetic Core Scales and the PedsQL 3.0 Cancer Module were administered to children with cancer (aged 5-18 years and parents of such children (aged 2-18 years. For comparison, a survey on a demographically group-matched sample of the general population with children (aged 5-18 and parents of children (aged 2-18 years was conducted with the PedsQL 4.0 Genetic Core Scales. Result The minimal mean percentage of missing item responses (except the School Functioning scale supported the feasibility of the PedsQL 4.0 Generic Core Scales and 3.0 Cancer Module for Chinese children with cancer. Most of the scales showed satisfactory reliability with Cronbach's α of exceeding 0.70, and all scales demonstrated sufficient test-retest reliability. Assessing the clinical validity of the questionnaires, statistically significant difference was found between healthy children and children with cancer, and between children on-treatment versus off-treatment ≥12 months. Positive significant correlations were observed between the scores of the PedsQL 4.0 Generic Core Scale and the PedsQL 3.0 Cancer Module. Exploratory factor analysis demonstrated sufficient factorial validity. Moderate to good agreement was found between child self- and parent proxy-reports. Conclusion The findings support the feasibility, reliability and validity of the Chinese Mandarin version of PedsQL 4.0 Generic Core Scales and 3.0 Cancer Module in children with cancer living in mainland China.

  2. Reliability model analysis and primary experimental evaluation of laser triggered pulse trigger

    International Nuclear Information System (INIS)

    Chen Debiao; Yang Xinglin; Li Yuan; Li Jin

    2012-01-01

    High performance pulse trigger can enhance performance and stability of the PPS. It is necessary to evaluate the reliability of the LTGS pulse trigger, so we establish the reliability analysis model of this pulse trigger based on CARMES software, the reliability evaluation is accord with the statistical results. (authors)

  3. Reliability analysis for the creep rupture mode of failure

    International Nuclear Information System (INIS)

    Vaidyanathan, S.

    1975-01-01

    An analytical study has been carried out to relate the factors of safety employed in the design of a component to the probability of failure in the thermal creep rupture mode. The analysis considers the statistical variations in the operating temperature, stress and rupture time, and applies the life fraction damage criterion as the indicator of failure. Typical results for solution annealed type 304-stainless steel material for the temperature and stress variations expected in an LMFBR environment have been obtained. The analytical problem was solved by considering the joint distribution of the independent variables and deriving the distribution for the function associated with the probability of failure by integrating over proper regions as dictated by the deterministic design rule. This leads to a triple integral for the final probability of failure where the coefficients of variation associated with the temperature, stress and rupture time distributions can be specified by the user. The derivation is general, and can be used for time varying stress histories and the case of irradiated material where the rupture time varies with accumulated fluence. Example calculations applied to solution annealed type 304 stainless steel material have been carried out for an assumed coefficient of variation of 2% for temperature and 6% for stress. The results show that the probability of failure associated with dependent stress intensity limits specified in the ASME Boiler and Pressure Vessel Section III Code Case 1592 is less than 5x10 -8 . Rupture under thermal creep conditions is a highly complicated phenomenon. It is believed that the present study will help in quantizing the reliability to be expected with deterministic design factors of safety

  4. Experimental programme and analysis, ZENITH II, Core 4

    Energy Technology Data Exchange (ETDEWEB)

    Ingram, G.; Sanders, J. E.; Sherwin, J.

    1974-10-15

    The Phase 3 program of reactor physics experiments on the HTR (or Mk 3 GCR) lattices continued during the first half of 1974 with a study of a series of critical builds in Zenith II aimed at testing predictions of shut-down margins in the local criticality-situations arising during power reactor refueling. The paper describes the experimental program and the subsequent theoretical analysis using methods developed in the United Kingdom for calculating low-enriched uranium HTR fuel systems. The importance of improving the accuracy of predictions of shut-down margins arises from the basic requirement that the core in its most reactive condition and with a specified number of absorbers removed from the array must remain sub-critical with a margin adequate to cover the total uncertainty of +/- 1 Nile (that is, 1 % delta-k). The major uncertainty is that in modelling the complex fuel/absorber configuration, and this is the aspect essentially covered in the Zenith II Core 4 studies.

  5. Improvement of numerical analysis method for FBR core characteristics. 3

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yamamoto, Toshihisa; Kitada, Takanori; Katagi, Yousuke

    1998-03-01

    As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method; A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. Part 2: Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory; Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. Part 3: Improvement of Nodal Transport Calculation for Hexagonal Geometry; A method to evaluate the intra-subassembly power distribution from the nodal averaged neutron flux and surface fluxes at the node boundaries, was developed based on the transport theory. (J.P.N.)

  6. FBR core mock-up RAPSODIE I - experimental analysis

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.; Gantenbein, F.

    1990-01-01

    The main phenomena which influence the LMFBR core response to a seismic excitation are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour, seismic tests have been performed on the core mock-up RAPSODIE with and without fluid and restraint ring and for different levels of excitation. This paper summarizes the results of these tests. (author)

  7. Reliability analysis of a sensitive and independent stabilometry parameter set.

    Science.gov (United States)

    Nagymáté, Gergely; Orlovits, Zsanett; Kiss, Rita M

    2018-01-01

    Recent studies have suggested reduced independent and sensitive parameter sets for stabilometry measurements based on correlation and variance analyses. However, the reliability of these recommended parameter sets has not been studied in the literature or not in every stance type used in stabilometry assessments, for example, single leg stances. The goal of this study is to evaluate the test-retest reliability of different time-based and frequency-based parameters that are calculated from the center of pressure (CoP) during bipedal and single leg stance for 30- and 60-second measurement intervals. Thirty healthy subjects performed repeated standing trials in a bipedal stance with eyes open and eyes closed conditions and in a single leg stance with eyes open for 60 seconds. A force distribution measuring plate was used to record the CoP. The reliability of the CoP parameters was characterized by using the intraclass correlation coefficient (ICC), standard error of measurement (SEM), minimal detectable change (MDC), coefficient of variation (CV) and CV compliance rate (CVCR). Based on the ICC, SEM and MDC results, many parameters yielded fair to good reliability values, while the CoP path length yielded the highest reliability (smallest ICC > 0.67 (0.54-0.79), largest SEM% = 19.2%). Usually, frequency type parameters and extreme value parameters yielded poor reliability values. There were differences in the reliability of the maximum CoP velocity (better with 30 seconds) and mean power frequency (better with 60 seconds) parameters between the different sampling intervals.

  8. Reliability analysis of a sensitive and independent stabilometry parameter set

    Science.gov (United States)

    Nagymáté, Gergely; Orlovits, Zsanett

    2018-01-01

    Recent studies have suggested reduced independent and sensitive parameter sets for stabilometry measurements based on correlation and variance analyses. However, the reliability of these recommended parameter sets has not been studied in the literature or not in every stance type used in stabilometry assessments, for example, single leg stances. The goal of this study is to evaluate the test-retest reliability of different time-based and frequency-based parameters that are calculated from the center of pressure (CoP) during bipedal and single leg stance for 30- and 60-second measurement intervals. Thirty healthy subjects performed repeated standing trials in a bipedal stance with eyes open and eyes closed conditions and in a single leg stance with eyes open for 60 seconds. A force distribution measuring plate was used to record the CoP. The reliability of the CoP parameters was characterized by using the intraclass correlation coefficient (ICC), standard error of measurement (SEM), minimal detectable change (MDC), coefficient of variation (CV) and CV compliance rate (CVCR). Based on the ICC, SEM and MDC results, many parameters yielded fair to good reliability values, while the CoP path length yielded the highest reliability (smallest ICC > 0.67 (0.54–0.79), largest SEM% = 19.2%). Usually, frequency type parameters and extreme value parameters yielded poor reliability values. There were differences in the reliability of the maximum CoP velocity (better with 30 seconds) and mean power frequency (better with 60 seconds) parameters between the different sampling intervals. PMID:29664938

  9. Analysis of NPP protection structure reliability under impact of a falling aircraft

    International Nuclear Information System (INIS)

    Shul'man, G.S.

    1996-01-01

    Methodology for evaluation of NPP protection structure reliability by impact of aircraft fall down is considered. The methodology is base on the probabilistic analysis of all potential events. The problem is solved in three stages: determination of loads on structural units, calculation of local reliability of protection structures by assigned loads and estimation of the structure reliability. The methodology proposed may be applied at the NPP design stage and by determination of reliability of already available structures

  10. Method of reliability allocation based on fault tree analysis and fuzzy math in nuclear power plants

    International Nuclear Information System (INIS)

    Chen Zhaobing; Deng Jian; Cao Xuewu

    2005-01-01

    Reliability allocation is a kind of a difficult multi-objective optimization problem. It can not only be applied to determine the reliability characteristic of reactor systems, subsystem and main components but also be performed to improve the design, operation and maintenance of nuclear plants. The fuzzy math known as one of the powerful tools for fuzzy optimization and the fault analysis deemed to be one of the effective methods of reliability analysis can be applied to the reliability allocation model so as to work out the problems of fuzzy characteristic of some factors and subsystem's choice respectively in this paper. Thus we develop a failure rate allocation model on the basis of the fault tree analysis and fuzzy math. For the choice of the reliability constraint factors, we choose the six important ones according to practical need for conducting the reliability allocation. The subsystem selected by the top-level fault tree analysis is to avoid allocating reliability for all the equipment and components including the unnecessary parts. During the reliability process, some factors can be calculated or measured quantitatively while others only can be assessed qualitatively by the expert rating method. So we adopt fuzzy decision and dualistic contrast to realize the reliability allocation with the help of fault tree analysis. Finally the example of the emergency diesel generator's reliability allocation is used to illustrate reliability allocation model and improve this model simple and applicable. (authors)

  11. Improvement of human reliability analysis method for PRA

    International Nuclear Information System (INIS)

    Tanji, Junichi; Fujimoto, Haruo

    2013-09-01

    It is required to refine human reliability analysis (HRA) method by, for example, incorporating consideration for the cognitive process of operator into the evaluation of diagnosis errors and decision-making errors, as a part of the development and improvement of methods used in probabilistic risk assessments (PRAs). JNES has been developed a HRA method based on ATHENA which is suitable to handle the structured relationship among diagnosis errors, decision-making errors and operator cognition process. This report summarizes outcomes obtained from the improvement of HRA method, in which enhancement to evaluate how the plant degraded condition affects operator cognitive process and to evaluate human error probabilities (HEPs) which correspond to the contents of operator tasks is made. In addition, this report describes the results of case studies on the representative accident sequences to investigate the applicability of HRA method developed. HEPs of the same accident sequences are also estimated using THERP method, which is most popularly used HRA method, and comparisons of the results obtained using these two methods are made to depict the differences of these methods and issues to be solved. Important conclusions obtained are as follows: (1) Improvement of HRA method using operator cognitive action model. Clarification of factors to be considered in the evaluation of human errors, incorporation of degraded plant safety condition into HRA and investigation of HEPs which are affected by the contents of operator tasks were made to improve the HRA method which can integrate operator cognitive action model into ATHENA method. In addition, the detail of procedures of the improved method was delineated in the form of flowchart. (2) Case studies and comparison with the results evaluated by THERP method. Four operator actions modeled in the PRAs of representative BWR5 and 4-loop PWR plants were selected and evaluated as case studies. These cases were also evaluated using

  12. Analysis of core damage frequency from internal events: Methodology guidelines: Volume 1

    International Nuclear Information System (INIS)

    Drouin, M.T.; Harper, F.T.; Camp, A.L.

    1987-09-01

    NUREG-1150 examines the risk to the public from a selected group of nuclear power plants. This report describes the methodology used to estimate the internal event core damage frequencies of four plants in support of NUREG-1150. In principle, this methodology is similar to methods used in past probabilistic risk assessments; however, based on past studies and using analysts that are experienced in these techniques, the analyses can be focused in certain areas. In this approach, only the most important systems and failure modes are modeled in detail. Further, the data and human reliability analyses are simplified, with emphasis on the most important components and human actions. Using these methods, an analysis can be completed in six to nine months using two to three full-time systems analysts and part-time personnel in other areas, such as data analysis and human reliability analysis. This is significantly faster and less costly than previous analyses and provides most of the insights that are obtained by the more costly studies. 82 refs., 35 figs., 27 tabs

  13. Automated software analysis of nuclear core discharge data

    International Nuclear Information System (INIS)

    Larson, T.W.; Halbig, J.K.; Howell, J.A.; Eccleston, G.W.; Klosterbuer, S.F.

    1993-03-01

    Monitoring the fueling process of an on-load nuclear reactor is a full-time job for nuclear safeguarding agencies. Nuclear core discharge monitors (CDMS) can provide continuous, unattended recording of the reactor's fueling activity for later, qualitative review by a safeguards inspector. A quantitative analysis of this collected data could prove to be a great asset to inspectors because more information can be extracted from the data and the analysis time can be reduced considerably. This paper presents a prototype for an automated software analysis system capable of identifying when fuel bundle pushes occurred and monitoring the power level of the reactor. Neural network models were developed for calculating the region on the reactor face from which the fuel was discharged and predicting the burnup. These models were created and tested using actual data collected from a CDM system at an on-load reactor facility. Collectively, these automated quantitative analysis programs could help safeguarding agencies to gain a better perspective on the complete picture of the fueling activity of an on-load nuclear reactor. This type of system can provide a cost-effective solution for automated monitoring of on-load reactors significantly reducing time and effort

  14. Application of Metric-based Software Reliability Analysis to Example Software

    International Nuclear Information System (INIS)

    Kim, Man Cheol; Smidts, Carol

    2008-07-01

    The software reliability of TELLERFAST ATM software is analyzed by using two metric-based software reliability analysis methods, a state transition diagram-based method and a test coverage-based method. The procedures for the software reliability analysis by using the two methods and the analysis results are provided in this report. It is found that the two methods have a relation of complementary cooperation, and therefore further researches on combining the two methods to reflect the benefit of the complementary cooperative effect to the software reliability analysis are recommended

  15. Development and analysis of U-core switched reluctance machine

    DEFF Research Database (Denmark)

    Jæger, Rasmus; Nielsen, Simon Staal; Rasmussen, Peter Omand

    2016-01-01

    Switched reluctance machines (SRMs) have seen a lot of interest due to their rugged and fault tolerant construction as well as their high efficiency over a wide speed range. The technology however suffers from torque ripple, acoustic noise and low torque density. Many concepts to address these di......Switched reluctance machines (SRMs) have seen a lot of interest due to their rugged and fault tolerant construction as well as their high efficiency over a wide speed range. The technology however suffers from torque ripple, acoustic noise and low torque density. Many concepts to address...... and reduced flux reversal, reducing core losses. Due to an increased number of poles, torque density is increased and torque ripple reduced. A prototype is built and through a number of tests, the machine is mapped and all loss components are analysed. As a result of the analysis, an assessment is presented...

  16. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Tsuchida, Noboru; Shiraishi, Tadao; Takahashi, Yutaka; Inada, Seiji; Saito, Minoru; Futamura, Yoshiaki; Kitano, Kyoshiro.

    1992-10-01

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

  17. A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2005-02-01

    In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

  18. Reliability Analysis Multiple Redundancy Controller for Nuclear Safety Systems

    International Nuclear Information System (INIS)

    Son, Gwangseop; Kim, Donghoon; Son, Choulwoong

    2013-01-01

    This controller is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). The architecture of MRC is briefly described, and the Markov model is developed. Based on the model, the reliability and Mean Time To Failure (MTTF) are analyzed. In this paper, the architecture of MRC for nuclear safety systems is described. The MRC is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). Markov models for MRC architecture was developed, and then the reliability was analyzed by using the model. From the reliability analyses for the MRC, it is obtained that the failure rate of each module in the MRC should be less than 2 Χ 10 -4 /hour and the MTTF average increase rate depending on FCF increment, i. e. ΔMTTF/ΔFCF, is 4 months/0.1

  19. Reliability Engineering Analysis of ATLAS Data Reprocessing Campaigns

    CERN Document Server

    Vaniachine, A; The ATLAS collaboration; Karpenko, D

    2013-01-01

    During three years of LHC data taking, the ATLAS collaboration completed three petascale data reprocessing campaigns on the Grid, with up to 2 PB of data being reprocessed every year. In reprocessing on the Grid, failures can occur for a variety of reasons, while Grid heterogeneity makes failures hard to diagnose and repair quickly. As a result, Big Data processing on the Grid must tolerate a continuous stream of failures, errors and faults. While ATLAS fault-tolerance mechanisms improve the reliability of Big Data processing in the Grid, their benefits come at costs and result in delays making the performance prediction difficult. Reliability Engineering provides a framework for fundamental understanding of the Big Data processing on the Grid, which is not a desirable enhancement but a necessary requirement. In ATLAS, cost monitoring and performance prediction became critical for the success of the reprocessing campaigns conducted in preparation for the major physics conferences. In addition, our Reliability...

  20. Reliability Engineering Analysis of ATLAS Data Reprocessing Campaigns

    CERN Document Server

    Vaniachine, A; The ATLAS collaboration; Karpenko, D

    2014-01-01

    During three years of LHC data taking, the ATLAS collaboration completed three petascale data reprocessing campaigns on the Grid, with up to 2 PB of data being reprocessed every year. In reprocessing on the Grid, failures can occur for a variety of reasons, while Grid heterogeneity makes failures hard to diagnose and repair quickly. As a result, Big Data processing on the Grid must tolerate a continuous stream of failures, errors and faults. While ATLAS fault-tolerance mechanisms improve the reliability of Big Data processing in the Grid, their benefits come at costs and result in delays making the performance prediction difficult. Reliability Engineering provides a framework for fundamental understanding of the Big Data processing on the Grid, which is not a desirable enhancement but a necessary requirement. In ATLAS, cost monitoring and performance prediction became critical for the success of the reprocessing campaigns conducted in preparation for the major physics conferences. In addition, our Reliability...

  1. Long term reliability analysis of standby diesel generators

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1988-01-01

    The long term reliability of 11 diesel generators of 125 to 250 kV A size has been analysed from 26 years of data base information on individual diesel service as standby power supplies for the Chalk River research reactor facilities. Failure to start on demand and failure to run data is presented and failure by diesel subsystem and multiple failures are also analysed. A brief comparison is made with reliability studies of larger diesel generator units used for standby power service in nuclear power plants. (author)

  2. Technical information report: Plasma melter operation, reliability, and maintenance analysis

    International Nuclear Information System (INIS)

    Hendrickson, D.W.

    1995-01-01

    This document provides a technical report of operability, reliability, and maintenance of a plasma melter for low-level waste vitrification, in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. A process description is provided that minimizes maintenance and downtime and includes material and energy balances, equipment sizes and arrangement, startup/operation/maintence/shutdown cycle descriptions, and basis for scale-up to a 200 metric ton/day production facility. Operational requirements are provided including utilities, feeds, labor, and maintenance. Equipment reliability estimates and maintenance requirements are provided which includes a list of failure modes, responses, and consequences

  3. Embedded mechatronic systems 1 analysis of failures, predictive reliability

    CERN Document Server

    El Hami, Abdelkhalak

    2015-01-01

    In operation, mechatronics embedded systems are stressed by loads of different causes: climate (temperature, humidity), vibration, electrical and electromagnetic. These stresses in components which induce failure mechanisms should be identified and modeled for better control. AUDACE is a collaborative project of the cluster Mov'eo that address issues specific to mechatronic reliability embedded systems. AUDACE means analyzing the causes of failure of components of mechatronic systems onboard. The goal of the project is to optimize the design of mechatronic devices by reliability. The projec

  4. Total Longitudinal Moment Calculation and Reliability Analysis of Yacht Structures

    Science.gov (United States)

    Zhi, Wenzheng; Lin, Shaofen

    In order to check the reliability of the yacht in FRP (Fiber Reinforce Plastic) materials, in this paper, the vertical force and the calculation method of the overall longitudinal bending moment on yacht was analyzed. Specially, this paper focuses on the impact of speed on the still water bending moment on yacht. Then considering the mechanical properties of the cap type stiffeners in composite materials, the ultimate bearing capacity of the yacht has been worked out, finally the reliability of the yacht was calculated with using response surface methodology. The result can be used in yacht design and yacht driving.

  5. Reliability modeling and analysis of smart power systems

    CERN Document Server

    Karki, Rajesh; Verma, Ajit Kumar

    2014-01-01

    The volume presents the research work in understanding, modeling and quantifying the risks associated with different ways of implementing smart grid technology in power systems in order to plan and operate a modern power system with an acceptable level of reliability. Power systems throughout the world are undergoing significant changes creating new challenges to system planning and operation in order to provide reliable and efficient use of electrical energy. The appropriate use of smart grid technology is an important drive in mitigating these problems and requires considerable research acti

  6. Exploratory factor analysis and reliability analysis with missing data: A simple method for SPSS users

    Directory of Open Access Journals (Sweden)

    Bruce Weaver

    2014-09-01

    Full Text Available Missing data is a frequent problem for researchers conducting exploratory factor analysis (EFA or reliability analysis. The SPSS FACTOR procedure allows users to select listwise deletion, pairwise deletion or mean substitution as a method for dealing with missing data. The shortcomings of these methods are well-known. Graham (2009 argues that a much better way to deal with missing data in this context is to use a matrix of expectation maximization (EM covariances(or correlations as input for the analysis. SPSS users who have the Missing Values Analysis add-on module can obtain vectors ofEM means and standard deviations plus EM correlation and covariance matrices via the MVA procedure. But unfortunately, MVA has no /MATRIX subcommand, and therefore cannot write the EM correlations directly to a matrix dataset of the type needed as input to the FACTOR and RELIABILITY procedures. We describe two macros that (in conjunction with an intervening MVA command carry out the data management steps needed to create two matrix datasets, one containing EM correlations and the other EM covariances. Either of those matrix datasets can then be used asinput to the FACTOR procedure, and the EM correlations can also be used as input to RELIABILITY. We provide an example that illustrates the use of the two macros to generate the matrix datasets and how to use those datasets as input to the FACTOR and RELIABILITY procedures. We hope that this simple method for handling missing data will prove useful to both students andresearchers who are conducting EFA or reliability analysis.

  7. Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

    International Nuclear Information System (INIS)

    Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

    2011-01-01

    Highlights: → Advanced analysis and design techniques for innovative reactors are addressed. → Detailed investigation of a 3 pass core design with a multi-physics-scales tool. → Coupled 40-group neutron transport/equivalent channels TH core analyses methods. → Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. → High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the

  8. A reliable method for the stability analysis of structures ...

    African Journals Online (AJOL)

    The detection of structural configurations with singular tangent stiffness matrix is essential because they can be unstable. The secondary paths, especially in unstable buckling, can play the most important role in the loss of stability and collapse of the structure. A new method for reliable detection and accurate computation of ...

  9. Fiber Access Networks: Reliability Analysis and Swedish Broadband Market

    Science.gov (United States)

    Wosinska, Lena; Chen, Jiajia; Larsen, Claus Popp

    Fiber access network architectures such as active optical networks (AONs) and passive optical networks (PONs) have been developed to support the growing bandwidth demand. Whereas particularly Swedish operators prefer AON, this may not be the case for operators in other countries. The choice depends on a combination of technical requirements, practical constraints, business models, and cost. Due to the increasing importance of reliable access to the network services, connection availability is becoming one of the most crucial issues for access networks, which should be reflected in the network owner's architecture decision. In many cases protection against failures is realized by adding backup resources. However, there is a trade off between the cost of protection and the level of service reliability since improving reliability performance by duplication of network resources (and capital expenditures CAPEX) may be too expensive. In this paper we present the evolution of fiber access networks and compare reliability performance in relation to investment and management cost for some representative cases. We consider both standard and novel architectures for deployment in both sparsely and densely populated areas. While some recent works focused on PON protection schemes with reduced CAPEX the current and future effort should be put on minimizing the operational expenditures (OPEX) during the access network lifetime.

  10. Windfarm Generation Assessment for ReliabilityAnalysis of Power Systems

    DEFF Research Database (Denmark)

    Negra, Nicola Barberis; Holmstrøm, Ole; Bak-Jensen, Birgitte

    2007-01-01

    Due to the fast development of wind generation in the past ten years, increasing interest has been paid to techniques for assessing different aspects of power systems with a large amount of installed wind generation. One of these aspects concerns power system reliability. Windfarm modelling plays...

  11. Windfarm generation assessment for reliability analysis of power systems

    DEFF Research Database (Denmark)

    Negra, N.B.; Holmstrøm, O.; Bak-Jensen, B.

    2007-01-01

    Due to the fast development of wind generation in the past ten years, increasing interest has been paid to techniques for assessing different aspects of power systems with a large amount of installed wind generation. One of these aspects concerns power system reliability. Windfarm modelling plays...

  12. A comparative reliability analysis of ETCS train radio communications

    NARCIS (Netherlands)

    Hermanns, H.; Becker, B.; Jansen, D.N.; Damm, W.; Usenko, Y.S.; Fränzle, M.; Olderog, E.-R.; Podelski, A.; Wilhelm, R.

    StoCharts have been proposed as a UML statechart extension for performance and dependability evaluation, and were applied in the context of train radio reliability assessment to show the principal tractability of realistic cases with this approach. In this paper, we extend on this bare feasibility

  13. Architecture-Based Reliability Analysis of Web Services

    Science.gov (United States)

    Rahmani, Cobra Mariam

    2012-01-01

    In a Service Oriented Architecture (SOA), the hierarchical complexity of Web Services (WS) and their interactions with the underlying Application Server (AS) create new challenges in providing a realistic estimate of WS performance and reliability. The current approaches often treat the entire WS environment as a black-box. Thus, the sensitivity…

  14. Erratum: Comparative Analysis of Some Reliability Characteristics of ...

    African Journals Online (AJOL)

    ... are analyzed using kolmogorov's forward equation method. Comparisons are performed for specific values of system parameters. Finally, the configurations are ranked based on MTSF and ( AV(∞)) and the results show that configuration 3 is optimal. Keywords: Reliability, Availability, Deterioration, Repair, Replacement.

  15. Reliability analysis of common hazardous waste treatment processes

    International Nuclear Information System (INIS)

    Waters, R.D.

    1993-05-01

    Five hazardous waste treatment processes are analyzed probabilistically using Monte Carlo simulation to elucidate the relationships between process safety factors and reliability levels. The treatment processes evaluated are packed tower aeration, reverse osmosis, activated sludge, upflow anaerobic sludge blanket, and activated carbon adsorption

  16. Reliability analysis of common hazardous waste treatment processes

    Energy Technology Data Exchange (ETDEWEB)

    Waters, Robert D. [Vanderbilt Univ., Nashville, TN (United States)

    1993-05-01

    Five hazardous waste treatment processes are analyzed probabilistically using Monte Carlo simulation to elucidate the relationships between process safety factors and reliability levels. The treatment processes evaluated are packed tower aeration, reverse osmosis, activated sludge, upflow anaerobic sludge blanket, and activated carbon adsorption.

  17. Reliability of Computer Analysis of Electrocardiograms (ECG) of ...

    African Journals Online (AJOL)

    Background: Computer programmes have been introduced to electrocardiography (ECG) with most physicians in Africa depending on computer interpretation of ECG. This study was undertaken to evaluate the reliability of computer interpretation of the 12-Lead ECG in the Black race. Methodology: Using the SCHILLER ...

  18. Reliability analysis of the epidural spinal cord compression scale.

    Science.gov (United States)

    Bilsky, Mark H; Laufer, Ilya; Fourney, Daryl R; Groff, Michael; Schmidt, Meic H; Varga, Peter Paul; Vrionis, Frank D; Yamada, Yoshiya; Gerszten, Peter C; Kuklo, Timothy R

    2010-09-01

    The evolution of imaging techniques, along with highly effective radiation options has changed the way metastatic epidural tumors are treated. While high-grade epidural spinal cord compression (ESCC) frequently serves as an indication for surgical decompression, no consensus exists in the literature about the precise definition of this term. The advancement of the treatment paradigms in patients with metastatic tumors for the spine requires a clear grading scheme of ESCC. The degree of ESCC often serves as a major determinant in the decision to operate or irradiate. The purpose of this study was to determine the reliability and validity of a 6-point, MR imaging-based grading system for ESCC. To determine the reliability of the grading scale, a survey was distributed to 7 spine surgeons who participate in the Spine Oncology Study Group. The MR images of 25 cervical or thoracic spinal tumors were distributed consisting of 1 sagittal image and 3 axial images at the identical level including T1-weighted, T2-weighted, and Gd-enhanced T1-weighted images. The survey was administered 3 times at 2-week intervals. The inter- and intrarater reliability was assessed. The inter- and intrarater reliability ranged from good to excellent when surgeons were asked to rate the degree of spinal cord compression using T2-weighted axial images. The T2-weighted images were superior indicators of ESCC compared with T1-weighted images with and without Gd. The ESCC scale provides a valid and reliable instrument that may be used to describe the degree of ESCC based on T2-weighted MR images. This scale accounts for recent advances in the treatment of spinal metastases and may be used to provide an ESCC classification scheme for multicenter clinical trial and outcome studies.

  19. The reliability of ultrasound-guided core needle biopsy in the evaluation of non-palpable solid breast lesions using 18-gauge needles

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sung Chul; Kim, Young Sook [Chosun University College of Medicine, Gwangju (Korea, Republic of); Sneige, Nour [The University of Texas M.D. Andreson Carcer Canter, Houston (United States)

    2003-04-01

    Ultrasound-guided core needle biopsy (US CNB) is increasingly used in the histologic evaluation of non-palpable solid breast lesions. We retrospectively investigated the diagnostic accuracy of this technique, using an 18-gauge needle in 422 non-palpable breast lesions. 583 female patients with an average age 56 (range, 22-90) years underwent 590 US CNBs. Between January 1994 and December 1999, using 18-gauge needles, an average of four cores per lesion was obtained. Three hundred and eighty-five lesions were subsequently surgically excised; for 14 of these, the pathologic diagnosis was breast carcinoma metastasis, while 23 with benign diagnoses were clinically followed up for {>=}2.5 years and were considered for analysis. Of the 422 lesions, 340 (80.6%) were malignant [308 invasive, 24 ductal carcinoma in situ (DCIS), 7 DCIS with undetermined invasion and 1 DCIS vs. lobular carcinoma in situ], 67 (15.9%) were benign [30 fibroadenoma (FA) and 37 other diagnoses], and five (1.2%) were fibroepithelial lesions. The remaining ten samples (2,4%) included six cases of atypical ductal hyperplasia (ADH), two of atypical hyperplasia (AH), and two of lobular neoplasia. The sensitivity, specificity, positive predictive value, and negative predictive value of CNBs were 99%, 100%, 100%, and 96%, respectively. Two cases of invasive carcinoma were missed at CNB; there was no false-positive diagnosis. Five of six ADHs and one of two AHs were found to be carcinomas (3 DCIS and 3 infiltrating duct carcinomas). Sixteen of 24 (66.7%) cases of DCIS were found at excision to be invasion carcinomas. Of 31 FAs, two (6.5%) were found to be low-grade phyllodes tumor (PT). The five fibroepithelial lesions were shown at excision to be either PT (n=4) or FA (n=1). US CNB using an 18-gauge needle is a safe and reliable means of diagnosing breast carcinoma. Because of the high prevalence of ductal carcinoma is these lesions; findings of ADH/AH at US CNB indicate that surgical excision is needed

  20. Time-dependent reliability analysis of nuclear reactor operators using probabilistic network models

    International Nuclear Information System (INIS)

    Oka, Y.; Miyata, K.; Kodaira, H.; Murakami, S.; Kondo, S.; Togo, Y.

    1987-01-01

    Human factors are very important for the reliability of a nuclear power plant. Human behavior has essentially a time-dependent nature. The details of thinking and decision making processes are important for detailed analysis of human reliability. They have, however, not been well considered by the conventional methods of human reliability analysis. The present paper describes the models for the time-dependent and detailed human reliability analysis. Recovery by an operator is taken into account and two-operators models are also presented

  1. Reliability Analysis Study of Digital Reactor Protection System in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Guo, Xiao Ming; Liu, Tao; Tong, Jie Juan; Zhao, Jun

    2011-01-01

    The Digital I and C systems are believed to improve a plants safety and reliability generally. The reliability analysis of digital I and C system has become one research hotspot. Traditional fault tree method is one of means to quantify the digital I and C system reliability. Review of advanced nuclear power plant AP1000 digital protection system evaluation makes clear both the fault tree application and analysis process to the digital system reliability. One typical digital protection system special for advanced reactor has been developed, which reliability evaluation is necessary for design demonstration. The typical digital protection system construction is introduced in the paper, and the process of FMEA and fault tree application to the digital protection system reliability evaluation are described. Reliability data and bypass logic modeling are two points giving special attention in the paper. Because the factors about time sequence and feedback not exist in reactor protection system obviously, the dynamic feature of digital system is not discussed

  2. Uncertainly propagation analysis for Yonggwang nuclear unit 4 by McCARD/MASTER core analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Dong Hyuk; Shim, Hyung Jin; Kim, Chang Hyo [Seoul National University, Seoul (Korea, Republic of)

    2014-06-15

    This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (k{sub eff}), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

  3. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  4. Reliability importance analysis of Markovian systems at steady state using perturbation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Phuc Do Van [Institut Charles Delaunay - FRE CNRS 2848, Systems Modeling and Dependability Group, Universite de technologie de Troyes, 12, rue Marie Curie, BP 2060-10010 Troyes cedex (France); Barros, Anne [Institut Charles Delaunay - FRE CNRS 2848, Systems Modeling and Dependability Group, Universite de technologie de Troyes, 12, rue Marie Curie, BP 2060-10010 Troyes cedex (France)], E-mail: anne.barros@utt.fr; Berenguer, Christophe [Institut Charles Delaunay - FRE CNRS 2848, Systems Modeling and Dependability Group, Universite de technologie de Troyes, 12, rue Marie Curie, BP 2060-10010 Troyes cedex (France)

    2008-11-15

    Sensitivity analysis has been primarily defined for static systems, i.e. systems described by combinatorial reliability models (fault or event trees). Several structural and probabilistic measures have been proposed to assess the components importance. For dynamic systems including inter-component and functional dependencies (cold spare, shared load, shared resources, etc.), and described by Markov models or, more generally, by discrete events dynamic systems models, the problem of sensitivity analysis remains widely open. In this paper, the perturbation method is used to estimate an importance factor, called multi-directional sensitivity measure, in the framework of Markovian systems. Some numerical examples are introduced to show why this method offers a promising tool for steady-state sensitivity analysis of Markov processes in reliability studies.

  5. Reliability importance analysis of Markovian systems at steady state using perturbation analysis

    International Nuclear Information System (INIS)

    Phuc Do Van; Barros, Anne; Berenguer, Christophe

    2008-01-01

    Sensitivity analysis has been primarily defined for static systems, i.e. systems described by combinatorial reliability models (fault or event trees). Several structural and probabilistic measures have been proposed to assess the components importance. For dynamic systems including inter-component and functional dependencies (cold spare, shared load, shared resources, etc.), and described by Markov models or, more generally, by discrete events dynamic systems models, the problem of sensitivity analysis remains widely open. In this paper, the perturbation method is used to estimate an importance factor, called multi-directional sensitivity measure, in the framework of Markovian systems. Some numerical examples are introduced to show why this method offers a promising tool for steady-state sensitivity analysis of Markov processes in reliability studies

  6. Development of RBDGG Solver and Its Application to System Reliability Analysis

    International Nuclear Information System (INIS)

    Kim, Man Cheol

    2010-01-01

    For the purpose of making system reliability analysis easier and more intuitive, RBDGG (Reliability Block diagram with General Gates) methodology was introduced as an extension of the conventional reliability block diagram. The advantage of the RBDGG methodology is that the structure of a RBDGG model is very similar to the actual structure of the analyzed system, and therefore the modeling of a system for system reliability and unavailability analysis becomes very intuitive and easy. The main idea of the development of the RBDGG methodology is similar with that of the development of the RGGG (Reliability Graph with General Gates) methodology, which is an extension of a conventional reliability graph. The newly proposed methodology is now implemented into a software tool, RBDGG Solver. RBDGG Solver was developed as a WIN32 console application. RBDGG Solver receives information on the failure modes and failure probabilities of each component in the system, along with the connection structure and connection logics among the components in the system. Based on the received information, RBDGG Solver automatically generates a system reliability analysis model for the system, and then provides the analysis results. In this paper, application of RBDGG Solver to the reliability analysis of an example system, and verification of the calculation results are provided for the purpose of demonstrating how RBDGG Solver is used for system reliability analysis

  7. Similarity and uncertainty analysis of the ALLEGRO MOX core

    International Nuclear Information System (INIS)

    Vrban, B.; Hascik, J.; Necas, V.; Slugen, V.

    2015-01-01

    The similarity and uncertainty analysis of the ESNII+ ALLEGRO MOX core has identified specific problems and challenges in the field of neutronic calculations. Similarity assessment identified 9 partly comparable experiments where only one reached ck and E values over 0.9. However the Global Integral Index G remains still low (0.75) and cannot be judge das sufficient. The total uncertainty of calculated k eff induced by XS data is according to our calculation 1.04%. The main contributors to this uncertainty are 239 Pu nubar and 238 U inelastic scattering. The additional margin from uncovered sensitivities was determined to be 0.28%. The identified low number of similar experiments prevents the use of advanced XS adjustment and bias estimation methods. More experimental data are needed and presented results may serve as a basic step in development of necessary critical assemblies. Although exact data are not presented in the paper, faster 44 energy group calculation gives almost the same results in similarity analysis in comparison to more complex 238 group calculation. Finally, it was demonstrated that TSUNAMI-IP utility can play a significant role in the future fast reactor development in Slovakia and in the Visegrad region. Clearly a further Research and Development and strong effort should be carried out in order to receive more complex methodology consisting of more plausible covariance data and related quantities. (authors)

  8. Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

    International Nuclear Information System (INIS)

    Ohtaka, Masahiko; Ohshima, Hiroyuki

    1998-10-01

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

  9. Reliability evaluation of nuclear power plants by fault tree analysis

    International Nuclear Information System (INIS)

    Iwao, H.; Otsuka, T.; Fujita, I.

    1993-01-01

    As a work sponsored by the Ministry of International Trade and Industry, the Safety Information Research Center of NUPEC, using reliability data based on the operational experience of the domestic LWR Plants, has implemented FTA for the standard PWRs and BWRs in Japan with reactor scram due to system failures being at the top event. Up to this point, we have obtained the FT chart and minimal cut set for each type of system failure for qualitative evaluation, and we have estimated system unavailability, Fussell-Vesely importance and risk worth for components for quantitative evaluation. As the second stage of a series in our reliability evaluation work, another program was started to establish a support system. The aim of this system is to assist foreign and domestic plants in creating countermeasures when incidents occur, by providing them with the necessary information using the above analytical method and its results. (author)

  10. Reliability Analysis of Timber Structures through NDT Data Upgrading

    DEFF Research Database (Denmark)

    Sousa, Hélder; Sørensen, John Dalsgaard; Kirkegaard, Poul Henning

    The first part of this document presents, in chapter 2, a description of timber characteristics and common used NDT and MDT for timber elements. Stochastic models for timber properties and damage accumulation models are also referred. According to timber’s properties a framework is proposed...... for a safety reassessment procedure. For that purpose a theoretical background for structural reliability assessment including probabilistic concepts for structural systems and stochastic models are given in chapter 3. System models, both series and parallel systems, are presented as well as methods...... for robustness are dealt in chapter 5. The second part of this document begins in chapter 6, where a practical application of the premise definitions and methodologies is given through the implementation of upgraded models with NDT and MDT data. Structural life-cycle is, therefore, assessed and reliability...

  11. Reliability analysis of numerical simulation in near field behavior

    International Nuclear Information System (INIS)

    Kobayashi, Akira; Yamamoto, Kiyohito; Chijimatsu, Masakazu; Fujita, Tomoo

    2008-01-01

    The uncertainties of the boundary conditions, the elastic modulus and Poisson's ratio on the mechanical behavior at near field of high level radioactive waste repository were examined. The method used to examine the error propagation was the first order second moment method. The reliability of the maximum principal stress, maximum shear stress at crown of the tunnel and the minimum principal stress at spring line was examined for one million years. For elastic model, the reliability of the maximum shear stress gradually decreased while that of the maximum principle stress increased. That of the minimum principal stress was relatively low for one million years. This tendency was similar to that from the damage model. (author)

  12. Setting Component Priorities in Protecting NPPs against Cyber-Attacks Using Reliability Analysis Techniques

    International Nuclear Information System (INIS)

    Choi, Moon Kyoung; Seong, Poong Hyun; Son, Han Seong

    2017-01-01

    The digitalization of infrastructure makes systems vulnerable to cyber threats and hybrid attacks. According to ICS-CERT report, as time goes by, the number of vulnerabilities in ICS industries increases rapidly. Digital I and C systems have been developed and installed in nuclear power plants, and due to installation of the digital I and C systems, cyber security concerns are increasing in nuclear industry. However, there are too many critical digital assets to be inspected in digitalized NPPs. In order to reduce the inefficiency of regulation in nuclear facilities, the critical components that are directly related to an accident are elicited by using the reliability analysis techniques. Target initial events are selected, and their headings are analyzed through event tree analysis about whether the headings can be affected by cyber-attacks or not. Among the headings, the headings that can be proceeded directly to the core damage by the cyber-attack when they are fail are finally selected as the target of deriving the minimum cut-sets. We analyze the fault trees and derive the minimum set-cuts. In terms of original PSA, the value of probability for the cut-sets is important but the probability is not important in terms of cyber security of NPPs. The important factors is the number of basic events consisting of the minimal cut-sets that is proportional to vulnerability.

  13. Reliability Analysis of Public Survey in Satisfaction with Nuclear Safety

    Energy Technology Data Exchange (ETDEWEB)

    Park, Moon Soo; Moon, Joo Hyun; Kang, Chang Sun [Seoul National Univ., Seoul (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) carried out a questionnaire survey on public's understanding nuclear safety and regulation in order to grasp public acceptance for nuclear energy. The survey was planned to help to analyze public opinion on nuclear energy and provide basic data for advertising strategy and policy development. In this study, based on results of the survey, the reliability of the survey was evaluated according to each nuclear site.

  14. Reliability analysis of Airbus A-330 computer flight management system

    OpenAIRE

    Fajmut, Metod

    2010-01-01

    Diploma thesis deals with digitized, computerized flight control system »Fly-by-wire« and security aspects of the computer system of an aircraft Airbus A330. As for space and military aircraft structures is also in commercial airplanes, much of the financial contribution devoted to reliability. Conventional aircraft control systems have, and some are still, to rely on mechanical and hydraulic connections between the controls on aircraft operated by the pilot and control surfaces. But newer a...

  15. Reliability Analysis of Public Survey in Satisfaction with Nuclear Safety

    International Nuclear Information System (INIS)

    Park, Moon Soo; Moon, Joo Hyun; Kang, Chang Sun

    2005-01-01

    Korea Institute of Nuclear Safety (KINS) carried out a questionnaire survey on public's understanding nuclear safety and regulation in order to grasp public acceptance for nuclear energy. The survey was planned to help to analyze public opinion on nuclear energy and provide basic data for advertising strategy and policy development. In this study, based on results of the survey, the reliability of the survey was evaluated according to each nuclear site

  16. Launch and Assembly Reliability Analysis for Human Space Exploration Missions

    Science.gov (United States)

    Cates, Grant; Gelito, Justin; Stromgren, Chel; Cirillo, William; Goodliff, Kandyce

    2012-01-01

    NASA's future human space exploration strategy includes single and multi-launch missions to various destinations including cis-lunar space, near Earth objects such as asteroids, and ultimately Mars. Each campaign is being defined by Design Reference Missions (DRMs). Many of these missions are complex, requiring multiple launches and assembly of vehicles in orbit. Certain missions also have constrained departure windows to the destination. These factors raise concerns regarding the reliability of launching and assembling all required elements in time to support planned departure. This paper describes an integrated methodology for analyzing launch and assembly reliability in any single DRM or set of DRMs starting with flight hardware manufacturing and ending with final departure to the destination. A discrete event simulation is built for each DRM that includes the pertinent risk factors including, but not limited to: manufacturing completion; ground transportation; ground processing; launch countdown; ascent; rendezvous and docking, assembly, and orbital operations leading up to trans-destination-injection. Each reliability factor can be selectively activated or deactivated so that the most critical risk factors can be identified. This enables NASA to prioritize mitigation actions so as to improve mission success.

  17. Reliability analysis of a complex standby redundant systems

    International Nuclear Information System (INIS)

    Subramanian, R.; Anantharaman, V.

    1995-01-01

    In any redundant system, the state of the standby unit is usually taken to be hot, warm or cold. In this paper, we present a new model of a two unit standby system wherein the standby unit is put in cold state for a certain amount of time before it is allowed to become warm. Upon failure of the online unit, the standby unit, if in warm state, instantaneously starts operating online; if it is in cold state, an emergency switching is made which takes it to warm state (and hence online) either instantaneously or non-instantaneously--each with some probability; if it is under repair, the system breaks down. Assuming all the associated distributions to be general except that of the life time of the standby unit in the warm state, various reliability characteristics that are of interest to reliability engineers and system designers are derived. A comprehensive cost function is also constructed and is then optimized with respect to three different control parameters numerically. In addition numerical results are presented to illustrate the behaviour of the various reliability characteristics derived

  18. Microarray analysis in clinical oncology: pre-clinical optimization using needle core biopsies from xenograft tumors

    International Nuclear Information System (INIS)

    Goley, Elizabeth M; Anderson, Soni J; Ménard, Cynthia; Chuang, Eric; Lü, Xing; Tofilon, Philip J; Camphausen, Kevin

    2004-01-01

    labeled, would generate representative array profiles compared to larger excisional biopsy material. In this analysis correlation coefficients were obtained ranging from 0.750–0.834 between U251 biopsy cores and excised tumors, and 0.812–0.846 between DU145 biopsy cores and excised tumors. These data suggest that needle core biopsies can be used as reliable tissue samples for tumor microarray analysis after linear amplification and either indirect or direct labeling of the starting RNA

  19. Assessment of modern methods of human factor reliability analysis in PSA studies

    International Nuclear Information System (INIS)

    Holy, J.

    2001-12-01

    The report is structured as follows: Classical terms and objects (Probabilistic safety assessment as a framework for human reliability assessment; Human failure within the PSA model; Basic types of operator failure modelled in a PSA study and analyzed by HRA methods; Qualitative analysis of human reliability; Quantitative analysis of human reliability used; Process of analysis of nuclear reactor operator reliability in a PSA study); New terms and objects (Analysis of dependences; Errors of omission; Errors of commission; Error forcing context); and Overview and brief assessment of human reliability analysis (Basic characteristics of the methods; Assets and drawbacks of the use of each of HRA method; History and prospects of the use of the methods). (P.A.)

  20. Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

    International Nuclear Information System (INIS)

    Shin, Hyeong-Ki

    1999-01-01

    The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations