WorldWideScience

Sample records for plant creep failure

  1. Residual stress measurements with barkhausen noise in power plant creep failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Karvonen, I. [CoMoTest Oy, Maentsaelae (Finland)] Suominen, L. [Stresstech Oy, Jyvaeskylae (Finland)

    1998-12-31

    Continuously developing power and process industry needs predictive maintenance inspection methods in order to prevent failures with correctly timed and properly specified measures. Materials` monitoring has traditionally been non-destructive inspection to detect growing cracks or other deficiencies. Recently, after the development of portable stress measurement systems, some advances has been reached. Based on stress anomalies due to creep, fatigue or corrosion, new applications have been found in the use of Barkhausen noise inspection. When the Barkhausen noise findings have been simultaneously confirmed with other stress measuring methods, a wider acceptance of the application of the method can be proposed. (orig.) 7 refs.

  2. Residual stress measurements with barkhausen noise in power plant creep failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Karvonen, I. [CoMoTest Oy, Maentsaelae (Finland)] Suominen, L. [Stresstech Oy, Jyvaeskylae (Finland)

    1997-12-31

    Continuously developing power and process industry needs predictive maintenance inspection methods in order to prevent failures with correctly timed and properly specified measures. Materials` monitoring has traditionally been non-destructive inspection to detect growing cracks or other deficiencies. Recently, after the development of portable stress measurement systems, some advances has been reached. Based on stress anomalies due to creep, fatigue or corrosion, new applications have been found in the use of Barkhausen noise inspection. When the Barkhausen noise findings have been simultaneously confirmed with other stress measuring methods, a wider acceptance of the application of the method can be proposed. (orig.) 7 refs.

  3. Creep-Fatigue Failure Diagnosis

    Science.gov (United States)

    Holdsworth, Stuart

    2015-01-01

    Failure diagnosis invariably involves consideration of both associated material condition and the results of a mechanical analysis of prior operating history. This Review focuses on these aspects with particular reference to creep-fatigue failure diagnosis. Creep-fatigue cracking can be due to a spectrum of loading conditions ranging from pure cyclic to mainly steady loading with infrequent off-load transients. These require a range of mechanical analysis approaches, a number of which are reviewed. The microstructural information revealing material condition can vary with alloy class. In practice, the detail of the consequent cracking mechanism(s) can be camouflaged by oxidation at high temperatures, although the presence of oxide on fracture surfaces can be used to date events leading to failure. Routine laboratory specimen post-test examination is strongly recommended to characterise the detail of deformation and damage accumulation under known and well-controlled loading conditions to improve the effectiveness and efficiency of failure diagnosis. PMID:28793676

  4. Multiaxial Creep-Fatigue and Creep-Ratcheting Failures of Grade 91 and Haynes 230 Alloys Toward Addressing Design Issues of Gen IV Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Tasnim [North Carolina State Univ., Raleigh, NC (United States); Lissenden, Cliff [Penn State Univ., University Park, PA (United States); Carroll, Laura [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The proposed research will develop systematic sets of uniaxial and multiaxial experimental data at a very high temperature (850-950°C) for Alloy 617. The loading histories to be prescribed in the experiments will induce creep-fatigue and creep-ratcheting failure mechanisms. These experimental responses will be scrutinized in order to quantify the influences of temperature and creep on fatigue and ratcheting failures. A unified constitutive model (UCM) will be developed and validated against these experimental responses. The improved UCM will be incorporated into the widely used finite element commercial software packages ANSYS. The modified ANSYS will be validated so that it can be used for evaluating the very high temperature ASME-NH design-by-analysis methodology for Alloy 617 and thereby addressing the ASME-NH design code issues.

  5. Multiaxial Creep-Fatigue and Creep-Ratcheting Failures of Grade 91 and Haynes 230 Alloys Toward Addressing Design Issues of Gen IV Nuclear Power Plants

    International Nuclear Information System (INIS)

    Hassan, Tasnim; Lissenden, Cliff; Carroll, Laura

    2015-01-01

    The proposed research will develop systematic sets of uniaxial and multiaxial experimental data at a very high temperature (850-950°C) for Alloy 617. The loading histories to be prescribed in the experiments will induce creep-fatigue and creep-ratcheting failure mechanisms. These experimental responses will be scrutinized in order to quantify the influences of temperature and creep on fatigue and ratcheting failures. A unified constitutive model (UCM) will be developed and validated against these experimental responses. The improved UCM will be incorporated into the widely used finite element commercial software packages ANSYS. The modified ANSYS will be validated so that it can be used for evaluating the very high temperature ASME-NH design-by-analysis methodology for Alloy 617 and thereby addressing the ASME-NH design code issues.

  6. Gradual failure of structures in creep conditions

    International Nuclear Information System (INIS)

    Chrzanowski, M.; Latus, P.

    1993-01-01

    The most characteristic feature of progressive material deterioration in creep conditions Is its time-dependence. In structures this process comprises of three stages: 1. Incubation of a macroscopic defect at the time of First Crack Appearance (FCA); 2. Propagation of a macro-crack throughout the structural member at the Time of Member Failure (TMF); 3. Propagation of failure of consecutive structure members, leading to the Final Structure Collapse (FSC). The importance of a full analysis of a structure which comprises all above stages has been demonstrated previously. Corresponding times are denoted as t 1 , t 2 , t 3 respectively. Depending on many factors, like material properties, loading and supports, the ratio of t 1 /t 2 and t 2 /t 3 may vary significantly, and thus exhibiting a safety margin connected with damage propagation throughout the structure. However, the full analysis becomes very sophisticated since creep and damage evolutions law are often nonlinear ones, and analysis should include changing geometry of a structure. It was was found that the failure propagation in analysed structures appeared to be very sensitive to structures geometry and loading. The time ratio t 1 /t 3 depends on redundancy k (higher the redundancy lower the ratio), but structures collapse by local mechanisms. These mechanisms can be different depending on k. More decisive than redundancy is an overall configuration of loads and structure geometry because of different mechanism of final failure. So far, no general conclusion can be drawn, but the whole analysis resembles that of limit analysis for structures made of ideally plastic or elastic-plastic materials. Nevertheless it Is evident that full analysis of structures in creep conditions can significantly enhance the structures life-time expectation

  7. Creep failure of a spray drier

    CSIR Research Space (South Africa)

    Carter, P

    1998-06-01

    Full Text Available , and creep. The calculations pointed to creep, and no positive metallurgic or physical evidence was discovered to support any of the hypotheses. However, the compression stresses implied that creep deformation could have occurred without inducing discernible...

  8. Brittle Creep Failure, Critical Behavior, and Time-to-Failure Prediction of Concrete under Uniaxial Compression

    Directory of Open Access Journals (Sweden)

    Yingchong Wang

    2015-01-01

    Full Text Available Understanding the time-dependent brittle deformation behavior of concrete as a main building material is fundamental for the lifetime prediction and engineering design. Herein, we present the experimental measures of brittle creep failure, critical behavior, and the dependence of time-to-failure, on the secondary creep rate of concrete under sustained uniaxial compression. A complete evolution process of creep failure is achieved. Three typical creep stages are observed, including the primary (decelerating, secondary (steady state creep regime, and tertiary creep (accelerating creep stages. The time-to-failure shows sample-specificity although all samples exhibit a similar creep process. All specimens exhibit a critical power-law behavior with an exponent of −0.51 ± 0.06, approximately equal to the theoretical value of −1/2. All samples have a long-term secondary stage characterized by a constant strain rate that dominates the lifetime of a sample. The average creep rate expressed by the total creep strain over the lifetime (tf-t0 for each specimen shows a power-law dependence on the secondary creep rate with an exponent of −1. This could provide a clue to the prediction of the time-to-failure of concrete, based on the monitoring of the creep behavior at the steady stage.

  9. Creep failure analysis of butt welded tubes

    International Nuclear Information System (INIS)

    Browne, R.J.; Parker, J.D.; Walters, D.J.

    1981-01-01

    As part of a major research programme to investigate the influence of butt welds on the life expectancy of tubular components, a series of internal-pressure, stress-rupture tests have been carried out. Thick walled 1/2Cr 1/2Mo 1/4V tube specimens were welded with mild steel, 1Cr 1/2Mo steel, 2 1/4Cr 1Mo steel or nominally matching 1/2Cr 1/2Mo 1/4V steel to give a wide range of weld metal creep strengths relative to the parent tube. The weldments were tested at 565 0 C at two values of internal pressure, and gave failure lives of up to 44,000 hrs. Finite element techniques have been used to determine the stationary state stress distribution in the weldment which was represented by a three material model. Significant stress redistribution was indicated and these results enabled the position and orientation of cracking and the rupture life to be predicted. The theoretical and experimental results have been used to highlight the limitations of current design methods which are based on the application of the mean diameter hoop stress to the parent material stress rupture data. (author)

  10. Unexpected damage and/or failures caused by creep below the limit temperature for creep design; Ovaentade krypskador och/eller haverier orsakade av krypmekanismer under graenstemperaturen

    Energy Technology Data Exchange (ETDEWEB)

    Storesund, Jan; Eklund, Anders; Taflin, Anders; Thunvik, Thomas

    2006-07-15

    Recently, several cases of cracking caused by creep have occurred in components operating at temperatures below the specified limit temperature for creep. Components operating below this limit temperature have not been designed with due regard to creep cracking and have accordingly not been subjected to inspection for creep damage. This work has surveyed the extent of these cases of creep damage by reviewing earlier failures and performed metallographic studies of damaged components and made parametric calculations of creep crack growth below the limit temperature. The following critical parameters have been determined for power plants: Creep damage below the transition temperature does not usually occur until operating times above 200.000 hours. Time to rupture differs from ordinary creep crack growth because these cracks have substantially longer incubation time of 20-30 years, with relative low creep deformation, and after that a rapid creep crack growth with only some few years to the creep rupture. Operation at 470-480 deg C, i.e. up to some 10 deg C below the transition temperature for a material like EN 13CrMo4-5, can be expected to result in severe creep damages comparable with ordinary creep failures at stressed locations. Operation at a temperature of 450-460 deg C can give rise to creep damage, however, this damage shows a more sparse occurrence. Creep damaged welds occurring below the limit temperature show cracks at the melting junction of the weld bead in opposite to ordinary creep damages. System stresses can also cause a more rapid crack growth. An international survey also shows that the variation of creep strength values between individual steel batches are just as wide as for ordinary creep. Based on this work, the following complementary recommendations can be issued: Elastic stress analysis (based on expansion calculations) can also be recommended for the identification of areas with intensified stresses. One should also perform a complete

  11. Low ductility creep failure in austenitic weld metals

    International Nuclear Information System (INIS)

    Thomas, R.G.

    Creep tests have been carried out for times of up to approx. 22,000 hrs on three austenitic weld metals of nominal composition 17Cr-8Ni-2Mo, 19Cr-12Ni-3Mo+Nb and 17Cr-10Ni-2Mo. The two former deposits were designed to produce delta-ferrite contents in the range 3-9% while the latter was designed to be fully austenitic. The common feature of all three weld metals was that they all gave very low strains at failure, typically approx. 1%. The microstructures of the failed creep specimens have been studied using optical and electron microscopy and the precipitate structures related to the occurrence of low creep strains. Creep deformation and fracture mechanisms in austenitic materials in general have been reviewed and this has been used as a basis for discussion of the observations of the present work. Finally, some of the factors that can be controlled to improve long-term creep ductility have been appraised

  12. Creep test observation of viscoelastic failure of edible fats

    Energy Technology Data Exchange (ETDEWEB)

    Vithanage, C R; Grimson, M J; Wills, P R [Department of Physics, University of Auckland, Private Bag 92019 (New Zealand); Smith, B G, E-mail: cvit002@aucklanduni.ac.nz [Food Science Programmes, Department of Chemistry, University of Auckland, Private Bag 92019 (New Zealand)

    2011-03-01

    A rheological creep test was used to investigate the viscoelastic failure of five edible fats. Butter, spreadable blend and spread were selected as edible fats because they belong to three different groups according to the Codex Alimentarius. Creep curves were analysed according to the Burger model. Results were fitted to a Weibull distribution representing the strain-dependent lifetime of putative fibres in the material. The Weibull shape and scale (lifetime) parameters were estimated for each substance. A comparison of the rheometric measurements of edible fats demonstrated a clear difference between the three different groups. Taken together the results indicate that butter has a lower threshold for mechanical failure than spreadable blend and spread. The observed behaviour of edible fats can be interpreted using a model in which there are two types of bonds between fat crystals; primary bonds that are strong and break irreversibly, and secondary bonds, which are weaker but break and reform reversibly.

  13. Comparison of stress-based and strain-based creep failure criteria for severe accident analysis

    International Nuclear Information System (INIS)

    Chavez, S.A.; Kelly, D.L.; Witt, R.J.; Stirn, D.P.

    1995-01-01

    We conducted a parametic analysis of stress-based and strain-based creep failure criteria to determine if there is a significant difference between the two criteria for SA533B vessel steel under severe accident conditions. Parametric variables include debris composition, system pressure, and creep strain histories derived from different testing programs and mathematically fit, with and without tertiary creep. Results indicate significant differences between the two criteria. Stress gradient plays an important role in determining which criterion will predict failure first. Creep failure was not very sensitive to different creep strain histories, except near the transition temperature of the vessel steel (900K to 1000K). Statistical analyses of creep failure data of four independent sources indicate that these data may be pooled, with a spline point at 1000K. We found the Manson-Haferd parameter to have better failure predictive capability than the Larson-Miller parameter for the data studied. (orig.)

  14. 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

  15. Creep property testing of energy power plant component material

    International Nuclear Information System (INIS)

    Nitiswati, Sri; Histori; Triyadi, Ari; Haryanto, Mudi

    1999-01-01

    Creep testing of SA213 T12 boiler piping material from fossil plant, Suralaya has been done. The aim of the testing is to know the creep behaviour of SA213 T12 boiler piping material which has been used more than 10 yeas, what is the material still followed ideal creep curve (there are primary stage, secondary stage, and tertiary stage). This possibility could happened because the material which has been used more than 10 years usually will be through ageing process because corrosion. The testing was conducted in 520 0C, with variety load between 4% until 50% maximum allowable load based on strength of the material in 520 0C

  16. Modelling the impact of creep on the probability of failure of a solid oxidefuel cell stack

    DEFF Research Database (Denmark)

    Greco, Fabio; Frandsen, Henrik Lund; Nakajo, Arata

    2014-01-01

    In solid oxide fuel cell (SOFC) technology a major challenge lies in balancing thermal stresses from an inevitable thermal field. The cells are known to creep, changing over time the stress field. The main objective of this study was to assess the influence of creep on the failure probability of ...

  17. Evaluation of creep damage in power plant applications

    Energy Technology Data Exchange (ETDEWEB)

    Auerkari, P; Salonen, J. [VTT Manufacturing Technology, Espoo (Finland)] McNiven, U. [IVO Generation Services Ltd., Naantali (Finland)] Roennberg, J. [Imatran Voima Oy, Vantaa (Finland)] Borggreen, K. [FORCE Institute, Broendby (Germany)

    1998-12-31

    Metallographic inspection of creep cavitation damage provides routine support for maintenance scheduling of high temperature components in power plants. The available European inspection experience has been reviewed, particularly considering the performance of thick-section steam systems outside the boiler. Applications are highlighted with examples from plant. (orig.) 8 refs.

  18. Evaluation of creep damage in power plant applications

    Energy Technology Data Exchange (ETDEWEB)

    Auerkari, P.; Salonen, J. [VTT Manufacturing Technology, Espoo (Finland)] McNiven, U. [IVO Generation Services Ltd., Naantali (Finland)] Roennberg, J. [Imatran Voima Oy, Vantaa (Finland)] Borggreen, K. [FORCE Institute, Broendby (Germany)

    1997-12-31

    Metallographic inspection of creep cavitation damage provides routine support for maintenance scheduling of high temperature components in power plants. The available European inspection experience has been reviewed, particularly considering the performance of thick-section steam systems outside the boiler. Applications are highlighted with examples from plant. (orig.) 8 refs.

  19. Finite Element Creep-Fatigue Analysis of a Welded Furnace Roll for Identifying Failure Root Cause

    Science.gov (United States)

    Yang, Y. P.; Mohr, W. C.

    2015-11-01

    Creep-fatigue induced failures are often observed in engineering components operating under high temperature and cyclic loading. Understanding the creep-fatigue damage process and identifying failure root cause are very important for preventing such failures and improving the lifetime of engineering components. Finite element analyses including a heat transfer analysis and a creep-fatigue analysis were conducted to model the cyclic thermal and mechanical process of a furnace roll in a continuous hot-dip coating line. Typically, the roll has a short life, modeling heat convection from hot air inside the furnace. The creep-fatigue analysis was performed by inputting the predicted temperature history and applying mechanical loads. The analysis results showed that the failure was resulted from a creep-fatigue mechanism rather than a creep mechanism. The difference of material properties between the filler metal and the base metal is the root cause for the roll failure, which induces higher creep strain and stress in the interface between the weld and the HAZ.

  20. Influence of microstructural parameters on the deformation and failure behaviour of the ODS alloy PM 2000 under creep and creep-fatigue loading

    International Nuclear Information System (INIS)

    Bothe, K.; Kussmaul, K.; Maile, K.

    1999-01-01

    The influence of grain size, manufacturing type and specimen direction (anisotropy) with respect to deformation and failure behaviour under creep, fatigue and creep-fatigue load was investigated. Thus, a basis for the correlation between microstructure and mechanical behaviour has been established. The specific damage and failure behaviour could be explained by means of the different microstructures observed. (orig.)

  1. Micromechanics of intergranular creep failure under cyclic loading

    DEFF Research Database (Denmark)

    van der Giessen, Erik; Tvergaard, Viggo

    1996-01-01

    boundaries are modelled individually. The model incorporates power-law creep of the grains, viscous grain boundary sliding between grains as well as the nucleation and growth of grain boundary cavities until they coalesce and form microcracks. Study of a limiting case with a facet-size microcrack reveals....... The analyses provide some new understanding that helps to explain the sometimes peculiar behaviour under balanced cyclic creep. Copyright (C) 1996 Acta Metallurgica Inc....

  2. Estimations of creep behavior and failure life for a circumferentially notched specimen

    International Nuclear Information System (INIS)

    Kobayashi, Ken-ichi; Yokobori, Toshimitsu; Kikuchi, Kenji.

    1997-01-01

    No method with which to characterize and/or illustrate total creep behavior for specimens with notches, holes or cracks has been proposed. In this paper it is proposed that most creep curves can be drawn with a master curve for each creep test whenever test conditions and failure modes are similar to each other, and the lifetime ratio normalized by the rupture time is introduced. Using smooth and circumferentially notched specimens of 2.25 Cr-1 Mo steel, creep tests were performed at 600degC for examination of this concept. Furthermore, a θ projection method was used to describe creep curves for notched specimens and to extrapolate longer creep lives. Then, the whole creep curve shape for notched specimens could be easily drawn, except for that in the vicinity of the rupture point. However, longer creep lives of notched specimens were underestimated in comparison with a simple extrapolation of the experimental data. This resulted from the negative dependence of the parameter of θ 3 on the applied stress. (author)

  3. Study of the concrete tensile creep: application for the containment vessel of the nuclear power plants (PWR)

    International Nuclear Information System (INIS)

    Reviron, Nanthilde

    2009-01-01

    The aim of this work is to study experimentally and to conduct numerical simulations on the creep of concrete subjected to tensile stresses. The main purpose is to predict the behaviour of containment vessels of nuclear power plants (PWR) in the case of decennial test or accident. In order to satisfy to these industrial needs, it is necessary to characterize the behaviour of concrete under uniaxial tension. Thus, an important experimental study of tensile creep in concrete has been performed for different loading levels (50%, 70% and 90% of the tensile strength). In these tests, load was kept constant during 3 days. Several tests were performed: measurements of elastic properties and strength (in tension and in compression), monitoring of drying, shrinkage, basic creep and drying creep strains. Moreover, compressive creep tests were also performed and showed a difference with tensile creep. Furthermore, decrease of tensile strength and failure under tensile creep for large loading levels were observed. A numerical model has been proposed and developed in Cast3m finite element code. (author)

  4. Development of a Zircaloy creep and failure model for LOCA conditions

    International Nuclear Information System (INIS)

    Raff, S.; Meyder, R.

    1981-01-01

    The present status of NORA model for zircaloy-4 creep and failure in the high temperature region (from 600 deg C up to 1200 deg C) is described. Temperature dependence, strain hardening and oxygen content are found to be the most important features of the strain rate creep equation. The failure criterion is based on a modified strain fraction rule. Variables of this criterion are temperature, strain rate or applied stress respectively and oxygen content. Concerning the application of the deformation model, deduced from uniaxial tests, to tube deformation calculation the axial ballooning shape has to be taken into account. Its influence on the tube stress components and therefore on strain rate is discussed. A further improvement of the deformation model concerning yield drop and irregular creep behaviour aims at the enlargement of the range of applicability and reduction of the error band of the model

  5. Experimental and analytical studies on creep failure of reactor coolant piping

    Energy Technology Data Exchange (ETDEWEB)

    Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakamura, N.

    1999-07-01

    Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

  6. Experimental and analytical studies on creep failure of reactor coolant piping

    International Nuclear Information System (INIS)

    Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun; Nakamura, N.

    1999-01-01

    Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

  7. Investigation of Macroscopic Brittle Creep Failure Caused by Microcrack Growth Under Step Loading and Unloading in Rocks

    Science.gov (United States)

    Li, Xiaozhao; Shao, Zhushan

    2016-07-01

    The growth of subcritical cracks plays an important role in the creep of brittle rock. The stress path has a great influence on creep properties. A micromechanics-based model is presented to study the effect of the stress path on creep properties. The microcrack model of Ashby and Sammis, Charles' Law, and a new micro-macro relation are employed in our model. This new micro-macro relation is proposed by using the correlation between the micromechanical and macroscopic definition of damage. A stress path function is also introduced by the relationship between stress and time. Theoretical expressions of the stress-strain relationship and creep behavior are derived. The effects of confining pressure on the stress-strain relationship are studied. Crack initiation stress and peak stress are achieved under different confining pressures. The applied constant stress that could cause creep behavior is predicted. Creep properties are studied under the step loading of axial stress or the unloading of confining pressure. Rationality of the micromechanics-based model is verified by the experimental results of Jinping marble. Furthermore, the effects of model parameters and the unloading rate of confining pressure on creep behavior are analyzed. The coupling effect of step axial stress and confining pressure on creep failure is also discussed. The results provide implications on the deformation behavior and time-delayed rockburst mechanism caused by microcrack growth on surrounding rocks during deep underground excavations.

  8. Investigation of Rock Failure Pattern in Creep by Digital Speckle Correlation Method

    Directory of Open Access Journals (Sweden)

    Yunliang Tan

    2013-01-01

    Full Text Available In order to study the mechanical characteristics from creep deformation to failure of rock, the tests of uniaxial compression and pushing steel-plate anchored in rock were performed, by using RLJW-2000 servo test synchronizing with Digital Speckle Correlation Method (DSCM. The investigations showed that for a uniaxial compressive specimen, when load arrived at 0.5σc, displacement clusters orderly formed, which was ahead of the macrocreep strain occurring in a slight jump mode when load arrived at 0.7σc. When the load level arrived at 0.8σc, displacement clusters gathered to be a narrow band. After that, the specimen abruptly fractured in a shear mode. In the creep pushing steel-plate test, when pushing force arrived at 25 kN, crack began to occur, the horizontal displacement field as well as shear strain field concentrated continuously along the interface between steel-plate and rock, and a new narrow concentrating band gathered in the upper layer. When pushing force arrived at 27.5 kN, another new narrow shear deformation band formed in the lower layer. Then, the steel-plate was pushed out quickly accompanying strong creep deformation.

  9. An assessment of the linear damage summation method for creep-fatigue failure with reference to a cast of type 316 stainless steel tested at 570 deg. C

    International Nuclear Information System (INIS)

    Wareing, J.; Bretherton, I.

    This paper presents preliminary results from the programme for hold period tests on a cast BQ of type 316 stainless steel at 570 deg. C. The results of tensile hold period tests on a relatively low ductility cast of type 316 stainless steel have indicated that the failure mechanism changes from a creep-fatigue interaction failure to a creep dominated failure at low strain levels. An assessment of the linear damage summation approach for failure prediction indicates that it is inappropriate for creep-fatigue interaction failures. For creep dominated fracture, failure occurs when the accumulation relaxation strain exhausts the material ductility i.e. Nsub(f epsilon R)=D. The failure criterion based on a creep summation in terms of time to fracture underestimates life

  10. 4. Nuclear power plant component failures

    International Nuclear Information System (INIS)

    1990-01-01

    Nuclear power plant component failures are dealt with in relation to reliability in nuclear power engineering. The topics treated include classification of failures, analysis of their causes and impacts, nuclear power plant failure data acquisition and processing, interdependent failures, and human factor reliability in nuclear power engineering. (P.A.). 8 figs., 7 tabs., 23 refs

  11. Improved methods for prediction of creep-fatigue in next generation conventional and nuclear plant

    International Nuclear Information System (INIS)

    Payten, Warwick

    2012-01-01

    Materials technology poses a major challenge in the design and construction of next generation super critical/ultra super critical power plant (SC/USC) and Generation IV (GenIV) nuclear plant. New plant is expected to have in the order of a 60 year life-time, imposing complex design difficulties in areas of creep rupture and creep fatigue damage. For SC/USC plant, the main goal is the enhancement of performance by raising the steam pressure and temperatures. In order to achieve these goals materials with acceptable creep rupture strength at design temperatures and pressures must be used. In GenIV designs, the issue is more complex, with both low and high tempera-ture designs. A key requirement in the majority of the designs, however, will be acceptable resistance to creep rupture, fatigue cracking, creep fatigue interactions, with the additional effects of void swelling and irradiation creep. The accumulation of creep fatigue damage over time in both SC/USC and GenIV plant will be one of the principal damage mechanisms. This will eventually lead to crack initiation in critical high temperature equipment. Hence, improved knowledge of creep and fatigue interactions is a necessary development as components in power-generating plants move to operate at high temperature under cyclic conditions. The key to safe, reliable operation of these high-energy plants will depend on understanding the factors that affect damage initiation and propagation, as well as developing and validating technologies to predict the accumulation of damage in systems and components.

  12. Computational models for residual creep life prediction of power plant components

    International Nuclear Information System (INIS)

    Grewal, G.S.; Singh, A.K.; Ramamoortry, M.

    2006-01-01

    All high temperature - high pressure power plant components are prone to irreversible visco-plastic deformation by the phenomenon of creep. The steady state creep response as well as the total creep life of a material is related to the operational component temperature through, respectively, the exponential and inverse exponential relationships. Minor increases in the component temperature can thus have serious consequences as far as the creep life and dimensional stability of a plant component are concerned. In high temperature steam tubing in power plants, one mechanism by which a significant temperature rise can occur is by the growth of a thermally insulating oxide film on its steam side surface. In the present paper, an elegantly simple and computationally efficient technique is presented for predicting the residual creep life of steel components subjected to continual steam side oxide film growth. Similarly, fabrication of high temperature power plant components involves extensive use of welding as the fabrication process of choice. Naturally, issues related to the creep life of weldments have to be seriously addressed for safe and continual operation of the welded plant component. Unfortunately, a typical weldment in an engineering structure is a zone of complex microstructural gradation comprising of a number of distinct sub-zones with distinct meso-scale and micro-scale morphology of the phases and (even) chemistry and its creep life prediction presents considerable challenges. The present paper presents a stochastic algorithm, which can be' used for developing experimental creep-cavitation intensity versus residual life correlations for welded structures. Apart from estimates of the residual life in a mean field sense, the model can be used for predicting the reliability of the plant component in a rigorous probabilistic setting. (author)

  13. Irradiation creep of candidate materials for advanced nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J., E-mail: jiachao.chen@psi.ch; Jung, P.; Hoffelner, W.

    2013-10-15

    In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2–8 × 10{sup −6} dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

  14. A reference stress approach for the characterisation of the creep failure of dissimilar welds under isothermal conditions

    International Nuclear Information System (INIS)

    Nicholson, R.D.; Williams, J.A.

    1988-11-01

    In high temperature power plant, welds between austenitic and ferritic steels are required to operate under plant conditions for up to 250,000h. The experience and failure modes for such joints are briefly surveyed in this report. A semi-empirical reference stress approach is used to define the failure life of joints under isothermal conditions. The reference stress is based on a previously published form for multiaxial creep fracture of homogeneous materials but modified to include an additional factor to reflect the complex strains present close to the interface in a dissimilar weld. This reference stress can be modified to give approximate bounds characterised by the equivalent stress or the axial stress on the weld. The reference stress, when applied to the 21/4Cr1Mo:Type 316 welded component data base, gives conservative results for the test data available although conservatism is low for the 9Cr1Mo:Alloy 600 combination. The existing data base for welded components is limited. More data are needed covering a wider range of stress ratios and incorporating bending loads. (author)

  15. Influences of cyclic deformation on creep property and creep-fatigue life prediction considering them

    International Nuclear Information System (INIS)

    Takahashi, Yukio

    2009-01-01

    Evaluation of creep-fatigue is essential in design and life management of high-temperature components in power generation plants. Cyclic deformation may alter creep property of the materials and its consideration may improve predictability of creep-fatigue failure life. To understand them, creep tests were conducted for the materials subjected to cyclic loading and their creep rupture and deformation behaviors were compared with those of as-received materials. Both 316FR and modified 9Cr-1Mo steel were tested. (1) Creep rupture time and elongation generally tend to decrease with cyclic loading in both materials, and especially elongation of 316FR drastically decreases by being cyclically deformed. (2) Amount of primary creep deformation decreases by cyclic loading and the ways to improve its predictability were developed. (3) Use of creep rupture ductility after cyclic deformation, instead of that of as-received material, brought about clear improvement of life prediction in a modified ductility exhaustion approach. (author)

  16. Metallurgical considerations in the design of creep exposed, high temperature components for advanced power plants

    International Nuclear Information System (INIS)

    Schubert, F.

    1990-08-01

    Metallic components in advanced power generating plants are subjected to temperatures at which the material properties are significantly time-dependent, so that the creep properties become dominant for the design. In this investigation, methods by which such components are to be designed are given, taking into account metallurgical principles. Experimental structure mechanics testing of component related specimens carried out for representative loading conditions has confirmed the proposed methods. The determination of time-dependent design values is based on a scatterband evaluation of long-term testing data obtained for a number of different heats of a given alloy. The application of computer-based databank systems is recommendable. The description of the technically important secondary creep rate based on physical metallurgy principles can be obtained using the exponential relationship originally formulated by Norton, ε min = k.σ n . The deformation of tubes observed under internal pressure with a superimposed static or cyclic tensile stress and a torsion loading can be adequately described with the derived, three-dimensional creep equation (Norton). This is also true for the description of creep ratcheting and creep buckling phenomena. By superimposing a cyclic stress, the average creep rate is increased in one of the principal deformation axes. This is also true for the creep crack growth rate. The Norton equation can be used to derive this type of deformation behaviour. (orig.) [de

  17. In-pile creep strain and failure of CW 316 Ti pressurized tubes

    International Nuclear Information System (INIS)

    Boutard, J.L.; Maillard, A.; Carteret, Y.; Levy, V.; Meny, L.

    1984-06-01

    The in-pile creep and failure behavior of CW 316 Ti pressurized tubes irradiated in the same rig at 660-680 0 C and 81.4 dpaF max in Phenix is presented and compared to monitors of the same heat. The in-pile plastic strains are of the same order of what is expected from the monitors and are rather independent of the dose rate in the range 4 to 9 x 10 -3 dpaF/h. Such a behavior supports the assumption that the out-of-pile deformation mechanisms are operative in pile and a certain balance occurs between modification of the microstructure, dynamic hardening and deformation mechanisms due to irradiation. Examinations by fractography and optical micrography, show that the failures are intergranular either in-pile or out-of-pile. In both cases the damage consists in intergranular wedge cracks, and no cavitation can be observed by transmission electron microscopy. Then the in-pile embrittlement which gives lower failure strain and time is to be associated with a decrease of the surface energy of grain-boundaries rather then growth and coalescence of cavities

  18. Failure Investigation of Radiant Platen Superheater Tube of Thermal Power Plant Boiler

    Science.gov (United States)

    Ghosh, D.; Ray, S.; Mandal, A.; Roy, H.

    2015-04-01

    This paper highlights a case study of typical premature failure of a radiant platen superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement and chemical analysis, are conducted as part of the investigations. Apart from these, metallographic analysis and fractography are also conducted to ascertain the probable cause of failure. Finally it has been concluded that the premature failure of the super heater tube can be attributed to localized creep at high temperature. The corrective actions has also been suggested to avoid this type of failure in near future.

  19. Creep life prediction of super heater coils used in coal based thermal power plants subjected to fly ash erosion and oxide scale formation

    Science.gov (United States)

    Srinivasan, P.; Kushwaha, Shashank

    2018-04-01

    Super heater coils of the coal based thermal power plants and subjected to severe operating conditions from both steam side and gas side. Formation of oxide scale due to prolonged service lead to temperature raise of the tube and erosion due to fly ash present in the combusted gases leads to tube thinning. Both these factors lead to creep rupture of the coils much before the designed service life. Failure of super heater coils during service of the boiler leads to power loss and huge monitory loss to the power plants. An attempt is made to model the creep damage caused to the super heater coils using heat transfer analysis tube thinning due to erosive wear of the tubes. Combined effects of these parameters are taken into consideration to predict the life of the super heater coils. This model may be used to estimate the life of the coils operating under the severe operating conditions to prevent the unexpected failure of the coils.

  20. Overheating failure of superheater suspension tubes of a captive thermal power plant boiler

    International Nuclear Information System (INIS)

    Bhattacharya, Sova; Amir, Q.M.; Kannan, C.; Mahapatra, S.B.

    2000-01-01

    Failure of boiler tubes is the foremost cause of forced boiler outages. One of the predominant failure mechanism of boiler tubes is the stress rupture failure in the form of either short term overheating or long term overheating which are normally encountered in superheater and reheater sections working in the creep range. The strength of the boiler tube depends on the stress level as well on the temperature of exposure in the creep range. An increase in either can reduce the time to rupture. Time at the exposure temperature is an important factor based on which the failures are categorised as either short term or long term. Though there is no established time duration criteria demarcating the short or long term stress rupture failures, depending on the various manifestations on the failed samples, one can categorise the failure. This paper addresses one such stress rupture failure in the superheater section of a captive thermal power plant of a refinery. Multiple failures on the suspension coil of a superheater section was investigated for the cause of failure. Laboratory investigation of the failed sample involved visual inspection, dimensional measurements, chemical analysis of internal deposits and microstructural study. On the basis of these, the failure was attributed to deposition of trisodium phosphate carried over by the feed water into the superheater section resulting in chokage and increase in local operating hoop stresses of the tube. The ultimate failure was thus categorised as long term overheating failure. (author)

  1. Simulation of creep tests with French or German RPV-steel and investigation of a RPV-support against failure

    International Nuclear Information System (INIS)

    Willschuetz, H.-G.; Altstadt, E.; Sehgal, B.R.; Weiss, F.-P.

    2003-01-01

    Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be considered for a determination of the loadings on the containment. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed in which the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties is performed. For the consideration of the tertiary creep stage and for the evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in three levels: starting with the simulation of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called 'tube-failure-experiments' are modeled: the RUPTHER-14 and the 'MPA-Meppen'-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi5-5 RPV-steels, which are chemically nearly identical. Since these two steels show a similar behavior, it should be allowed to a limited extend to transfer experimental and numerical

  2. Example of an application of an assessment procedure for defects in plant operating in the creep range

    International Nuclear Information System (INIS)

    Ainsworth, R.; Coleman, M.C.

    1986-10-01

    A worked example is presented to illustrate the application of a CEGB procedure for the assessment of defects in plant that operates within the creep range. The example chosen is a large cylindrical pressure vessel with a fully circumferential crack for which experimental results are available. The example demonstrates that it is straightforward to perform the calculations required in the procedure for the time for structural failure by continuum damage mechanisms, the time for crack incubation and the time for crack growth. It is predicted that crack incubation in the pressure vessel occurs at about one-third of life with the remaining life being taken up by crack growth. The predictions are in close agreement with the experimentally observed incubation, growth and final failure times. (UK)

  3. Modeling Creep-Fatigue-Environment Interactions in Steam Turbine Rotor Materials for Advanced Ultra-supercritical Coal Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Chen [General Electric Global Research, Niskayuna, NY (United States)

    2014-04-01

    The goal of this project is to model creep-fatigue-environment interactions in steam turbine rotor materials for advanced ultra-supercritical (A-USC) coal power Alloy 282 plants, to develop and demonstrate computational algorithms for alloy property predictions, and to determine and model key mechanisms that contribute to the damages caused by creep-fatigue-environment interactions.

  4. Small punch creep test: A promising methodology for high temperature plant components life evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Tettamanti, S. [CISE SpA, Milan (Italy); Crudeli, R. [ENEL SpA, Milan (Italy)

    1998-12-31

    CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

  5. Small punch creep test: A promising methodology for high temperature plant components life evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Tettamanti, S [CISE SpA, Milan (Italy); Crudeli, R [ENEL SpA, Milan (Italy)

    1999-12-31

    CISE and ENEL are involved for years in a miniaturization creep methodology project to obtain similar non-destructive test with the same standard creep test reliability. The goal can be reached with `Small punch creep test` that collect all the requested characteristics; quasi nondestructive disk specimens extracted both on external or internal side of components, than accurately machined and tested on little and cheap apparatus. CISE has developed complete creep small punch procedure that involved peculiar test facility and correlation`s law comparable with the more diffused isostress methodology for residual life evaluation on ex-serviced high temperature plant components. The aim of this work is to obtain a simple and immediately applicable relationship useful for plant maintenance managing. More added work is need to validate the Small Punch methodology and for relationship calibration on most diffusion high temperature structural materials. First obtained results on a comparative work on ASTM A355 P12 ex-serviced pipe material are presented joint with a description of the Small Punch apparatus realized in CISE. (orig.) 6 refs.

  6. Predicting the creep life and failure mode of low-alloy steel weldments

    Energy Technology Data Exchange (ETDEWEB)

    Brear, J.M.; Middleton, C.J.; Aplin, P.F. [ERA Technology Ltd., Leatherhead (United Kingdom)

    1998-12-31

    This presentation reviews and consolidates experience gained through a number of research projects and practical plant assessments in predicting both the life and the likely failure mode and location in low alloy steel weldments. The approach adopted begins with the recognition that the relative strength difference between the microstructural regions is a key factor controlling both life and failure location. Practical methods based on hardness measurement and adaptable to differing weld geometries are presented and evidence for correlations between hardness ratio, damage accumulation and strain development is discussed. Predictor diagrams relating weld life and failure location to the service conditions and the hardness of the individual microstructural constituents are suggested and comments are given on the implications for identifying the circumstances in which Type IV cracking is to be expected. (orig.) 6 refs.

  7. Predicting the creep life and failure mode of low-alloy steel weldments

    Energy Technology Data Exchange (ETDEWEB)

    Brear, J M; Middleton, C J; Aplin, P F [ERA Technology Ltd., Leatherhead (United Kingdom)

    1999-12-31

    This presentation reviews and consolidates experience gained through a number of research projects and practical plant assessments in predicting both the life and the likely failure mode and location in low alloy steel weldments. The approach adopted begins with the recognition that the relative strength difference between the microstructural regions is a key factor controlling both life and failure location. Practical methods based on hardness measurement and adaptable to differing weld geometries are presented and evidence for correlations between hardness ratio, damage accumulation and strain development is discussed. Predictor diagrams relating weld life and failure location to the service conditions and the hardness of the individual microstructural constituents are suggested and comments are given on the implications for identifying the circumstances in which Type IV cracking is to be expected. (orig.) 6 refs.

  8. Creep tests on clean and argillaceous salt from the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Mellegard, K.D.; Pfeifle, T.W.

    1993-05-01

    Fifteen triaxial compression creep tests were performed on clean and argillaceous salt from the Waste Isolation Pilot Plant (WIPP). The temperatures in the tests were either 25 degrees C or 100 degrees C while the stress difference ranged from 3.5 MPa to 21.0 MPa. In all tests, the confining pressure was 15 MPa. Test duration ranged from 23 to 613 days with an average duration of 300 days. The results of the creep tests supplemented earlier testing and were used to estimate two parameters in the Modified Munson-Dawson constitutive law for the creep behavior of salt. The two parameters determined from each test were the steady-state strain rate and the transient strain limit. These estimates were combined with parameter estimates determined from previous testing to study the dependence of both transient and steady-state creep deformation on stress difference. The exponents on stress difference determined in this study were found to be consistent with revised estimates of the exponents reported by other investigators

  9. Creep-fatigue monitoring system for header ligaments of fossil power plants

    International Nuclear Information System (INIS)

    Chen, K.L.; Deardorf, A.F.; Copeland, J.F.; Pflasterer, R.; Beckerdite, G.

    1993-01-01

    The cracking of headers (primary and secondary superheater outlet, and reheater outlet headers) at ligament locations is an important issue for fossil power plants. A model for crack initiation and growth has been developed, based on creep-fatigue damage mechanisms. This cracking model is included in a creep-fatigue monitoring system to assess header structural integrity under high temperature operating conditions. The following principal activities are required to achieve this goal: (1) the development of transfer functions and (2) the development of a ligament cracking model. The first task is to develop stress transfer functions to convert measured (monitored) temperatures, pressures and flow rates into stresses to be used to compute damage. Elastic three-dimensional finite element analyses were performed to study transient thermal stress behavior. The sustained pressure stress redistribution due to high temperature creep was studied by nonlinear finite element analyses. The preceding results are used to derive Green's functions and pressure stress gradient transfer functions for monitoring at the juncture of the tube with the header inner surface, and for crack growth at the ligaments. The virtual crack closure method is applied to derive a stress intensity factor K solution for a corner crack at the tube/header juncture. Similarly, using the reference stress method, the steady state creep crack growth parameter C * is derived for a header corner crack. The C * solution for a small corner crack in a header can be inserted directed into the available C t solution, along with K to provide the complete transient creep solution

  10. Oxidation and creep failure of alloy 617 foils at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, S.K.; Ko, G.D.; Li, F.X. [Department of Mechanical Engineering, Chonnam National University, Gwangju 500 757 (Korea, Republic of); Kang, K.J. [Department of Mechanical Engineering, Chonnam National University, Gwangju 500 757 (Korea, Republic of)], E-mail: kjkang@chonnam.ac.kr

    2008-08-31

    The microstructure of thermally grown oxides (TGO) and the creep properties of alloy 617 were investigated. Oxidation and creep tests were performed on 100 {mu}m thick foils at 800-1000 deg. C in air environment, while the thickness of TGO was monitored in situ. According to energy dispersive X-ray (EDX) mapping micrographs observation, superficial dense oxides, chromia (Cr{sub 2}O{sub 3}), which was thermodynamically unstable at 1000 deg. C, and discrete internal oxides, alumina ({alpha}-Al{sub 2}O{sub 3}), were found. Consequently, the weight of the foil specimen decreased due to the spalling and volatilization of the Cr{sub 2}O{sub 3} oxide layer after an initial weight-gaining. Secondary and tertiary creeps were observed at 800 deg. C, while the primary, secondary and tertiary creeps were observed at 1000 deg. C. Dynamic recrystallization occurred at 800 deg. C and 900 deg. C, while partial dynamic recrystallization at 1000 deg. C. The apparent activation energy, Q{sub app}, for the creep deformation was 271 kJ/mol, which was independent of the applied stress.

  11. Review of creep resistant alloys for power plant applications

    Directory of Open Access Journals (Sweden)

    A. Nagode

    2011-01-01

    Full Text Available A paper describes the most popular alloys for power plant application as well as the most promising alloys for future application in that technology. The components in power plants operate in severe conditions (high temperatures and pressures and they are expected reliable service for 30 years and more. The correct choice of the material is, thus, of a very importance. The paper describes the development as well as advantages and disadvantages of convenient ferritic/martensitic steels, ferritic/bainitic steels, austenitic stainless steels and the new alloys for the application at temperatures of 650°C and more.

  12. Metallurgical principles of creep processes

    International Nuclear Information System (INIS)

    Bolton, C.J.

    1977-12-01

    A brief review is presented of current theories of a number of the physical processes which can be involved in deformation and fracture under creep conditions. The processes considered are power law creep, diffusion creep, grain boundary sliding, cavitation and other modes of failure, and creep crack growth. The note concludes with some suggestions for future work. (author)

  13. Failure diagnosis aiding device for plant equipment

    International Nuclear Information System (INIS)

    Uhara, Yoshihiko.

    1990-01-01

    The present invention intends to improve the efficiency of trouble shooting for equipments of industrial plants such as nuclear power plants. The device of the present invention comprises an intelligence base and an inference mechanism base. The intelligence base comprises a rule base, an information storing section having a part frame and a working frame and a user's frame. The parts frame contains the failure rate on every parts and data on related operations. The working frame contains the importance and frequency of working. The user's frame contains parameters showing the extent of user's skills. The rule base, the parts frame and the working frame can be selected in accordance with the extent of the user's skill in the inference mechanism. With such a constitution, failures can be checked with the intelligence base in accordance with the knowledges for the failures of the equipments and the extent of user's skill by way of the inference mechanism. (I.S.)

  14. LWR nuclear power plant component failures

    International Nuclear Information System (INIS)

    Schmidt, W.H.

    1980-10-01

    An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

  15. On the microstructural basis of creep strength and creep-fatigue interaction in 9-12 % Cr steels for application in power plants

    Energy Technology Data Exchange (ETDEWEB)

    Chilukuru, H

    2007-03-06

    As part of the efforts of preserving the environment it is necessary to reduce of the CO2 emissions from power plants. This can be done by increasing the plant efficiency. Research groups around the world are engaged in developing new steels capable of sustaining higher stresses and temperatures envisaged for high-efficiency power plants. Research carried out in Europe is organized within the COST Programme (Co-Operation in Science and Technology) aiming at replacing the conventional steels of type X20CrMoV121 by the new class of 9-12% Cr-steels with modified composition. The resistance of materials against deformation at elevated temperatures depends on their microstructure. Frequently in 9-12% Cr-steels improved short-term creep properties do not persist in the long-term service [1, 2, 3, 4, 5, 6]. This is related with insufficient microstructural stability. Hardening contributions in 9-12% Cr-steels come from solute atoms of the ferritic matrix, from dislocations, and from precipitates of foreign phases within the matrix. The term ''carbide stabilized substructure hardening'' of 9-12% Cr steels [7, 8] indicates that the hardening contributions are interdependent. The dislocations are the carriers of plastic deformation. They interact with each other, with solute atoms and with precipitates. The dislocation-dislocation interaction leads to formation of planar dislocation networks constituting low-angle boundaries. They form a subgrain structure within the grains. At present, a full and detailed understanding of the effects exerted by the different components of microstructure on creep strength is still lacking. The present work makes a contribution to the efforts of understanding the microstructural basis of creep strength and of creep-fatigue interaction by transmission electron microscopic structure investigations coupled with creep tests. Investigations by transmission electron microscopy (TEM) were carried out with regard to hardening by subgrain boundaries

  16. On the microstructural basis of creep strength and creep-fatigue interaction in 9-12 % Cr steels for application in power plants

    Energy Technology Data Exchange (ETDEWEB)

    Chilukuru, H.

    2007-03-06

    As part of the efforts of preserving the environment it is necessary to reduce of the CO2 emissions from power plants. This can be done by increasing the plant efficiency. Research groups around the world are engaged in developing new steels capable of sustaining higher stresses and temperatures envisaged for high-efficiency power plants. Research carried out in Europe is organized within the COST Programme (Co-Operation in Science and Technology) aiming at replacing the conventional steels of type X20CrMoV121 by the new class of 9-12% Cr-steels with modified composition. The resistance of materials against deformation at elevated temperatures depends on their microstructure. Frequently in 9-12% Cr-steels improved short-term creep properties do not persist in the long-term service [1, 2, 3, 4, 5, 6]. This is related with insufficient microstructural stability. Hardening contributions in 9-12% Cr-steels come from solute atoms of the ferritic matrix, from dislocations, and from precipitates of foreign phases within the matrix. The term ''carbide stabilized substructure hardening'' of 9-12% Cr steels [7, 8] indicates that the hardening contributions are interdependent. The dislocations are the carriers of plastic deformation. They interact with each other, with solute atoms and with precipitates. The dislocation-dislocation interaction leads to formation of planar dislocation networks constituting low-angle boundaries. They form a subgrain structure within the grains. At present, a full and detailed understanding of the effects exerted by the different components of microstructure on creep strength is still lacking. The present work makes a contribution to the efforts of understanding the microstructural basis of creep strength and of creep-fatigue interaction by transmission electron microscopic structure investigations coupled with creep tests. Investigations by transmission electron microscopy (TEM) were carried out with regard to hardening by

  17. Advanced nondestructive evaluation for creep damage

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    As a result of operation at elevated temperatures, power plant components experience creep. Changes in metallurgical structure and microscopic cracking occur after periods of operation and lead to component failure. In order to detect the presence of creep and avoid creep-related failures, EPRI has just initiated a five year program entitled Advanced NDE for Creep Damage (RP 1856-7). The objective of this program is to develop NDE methods for detection and characterization of microscopic creep damage. Several NDE methods will be initially evaluated to determine their potential for detecting and characterizing such damage. These NDE methods include ultrasonics, eddy current, Barkhausen, positron annihilation, and thermal-wave imaging. A prototype system will be developed and tested for commercial applications in a follow-on project, utilizing characteristics of the best NDE method for creep detection. A brief description of the project and results of a theoretical investigation, to determine feasibility of ultrasonic NDE method, for detection of creep damage are presented

  18. Importance of creep failure of hard rock in the near field of a nuclear waste repository

    Energy Technology Data Exchange (ETDEWEB)

    Blacic, J D [Los Alamos National Laboratory, NM, (USA)

    1982-12-31

    Potential damage resulting from slow creep deformation intuitively seems unlikely for a high-level nuclear waste repository excavated in hard rock. However, recent experimental and modeling results indicate that the processes of time-dependent microcracking and water-induced stress corrosion can lead to significant reductions in strength and alteration of other key rock properties in the near-field region of a repository. We review the small data base supporting these conclusions and stress the need for an extensive laboratory program to obtain the new data that will be required for design of a repository.

  19. Corrosion-related failures in power plant condensers. Final report

    International Nuclear Information System (INIS)

    Beavers, J.A.; Agrawal, A.K.; Berry, W.E.

    1980-08-01

    A survey of the literature has been conducted for the Electric Power Research Institute on corrosion failures in surface condensers. The survey was directed toward condenser failures in pressurized water reactor (PWR) power plants but includes pertinent literature related to fossil and to other nuclear power plants. It includes literature on reported service failures and on experimental studies that impact on these failures

  20. The transverse creep deformation and failure characteristics of SCS-6/Ti-6Al-4V metal matrix composites at 482 C

    International Nuclear Information System (INIS)

    Eggleston, M.R.; Ritter, A.M.

    1995-01-01

    While continuous fiber, unidirectional composites are primarily evaluated for their longitudinal properties, the behavior transverse to the fibers often limits their application. In this study, the tensile and creep behaviors of SCS-6/Ti-6Al-4V composites in the transverse direction at 482 C were evaluated. Creep tests were performed in air and argon environments over the stress range of 103 to 276 MPa. The composite was less creep resistant than the matrix when tested at stress values larger than 150 MPa. Below 150 MPa, the composite was ore creep resistant than the unreinforced matrix. Failure of the composite occurred by the ductile propagation of racks emanating from separated fiber interfaces. The environment in which the test was performed affected the creep behavior. At 103 MPa, the creep rate in argon was 4 times slower than the creep rate in air. The SCS-6 silicon-carbide fiber's graphite coating oxidized in the air environment and encouraged the separation of the fiber-matrix interface. However, at high stress levels, the difference in behavior between air- and argon-tested specimens was small. At these stresses, separation of the interface occurred during the initial loading of the composite and the subsequent degradation of the interface did not affect the creep behavior. Finally, the enrichment of the composite's surface by molybdenum during fabrication resulted in an alloyed surface layer that failed in a brittle fashion during specimen elongation. Although this embrittled layer did not appear to degrade the properties of the composite, the existence of a similar layer on a composite with a more brittle matrix might be very detrimental

  1. Failure analysis of a boiler tube in USC coal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, N.H.; Kim, S.; Choe, B.H.; Yoon, K.B.; Kwon, D.I. [Kangnung National University, Kangnung (Republic of Korea)

    2009-10-15

    This paper presents failure analysis of final superheater tube in ultra-supercritical (USC) coal power plant. Visual inspection was performed to find out the characteristics of fracture of the as-received material. And the micro-structural changes such as grain growth and carbide coarsening was examined by scanning electron microscope. Detailed microscopic studies were made to find out the behavior of the scale exfoliation on the waterside of tubes. From those investigations, the creep rupture may be caused by the softened structure induced by carbide coarsening and accelerated by the metal temperature increase by the impediment of heat transfer due to voids.

  2. Creep life assessment of Mod.9Cr-1Mo steel. Pt. 2. Quantitative evaluation of microstructural damage in creep-interrupted specimens

    International Nuclear Information System (INIS)

    Sawada, Kota; Maruyama, Kouichi; Komine, Ryuji; Nagae, Yuji

    1998-02-01

    Mod.9Cr-1Mo steel has a martensitic lath structure. Recovery of the lath structure takes place in the course of creep. Microstructural degradation due to the recovery results in the acceleration of creep rate and the subsequent failure of a specimen. Change of lath width during creep of the steel was quantitatively investigated to propose a residual life assessment methodology based on the recovery process. Since the steel was tempered at 1053K, the lath structure is thermally stable at the testing temperatures (848K-923K). However, recovery of lath structure readily takes place during creep, indicating that the recovery is induced by creep deformation. Lath width d increases with creep strain and saturates to a value d s determined by creep stress. The increase of d is faster at a higher stress and temperature. A normalized change in lath width, Δd/Δd s , was introduced to explain the variation of lath growth rate with creep stress and temperature. Δd is the change in lath width from the initial value d 0 , and Δd s is the difference between d s and d 0 . Δd/Δd s is uniquely related to creep strain ε and the relationship is independent of creep stress as well as creep temperature. This Δd/Δd s -ε relationship obtained by an accelerated creep test at a higher temperature or stress is applicable to any creep condition including service conditions of engineering plants. Creep strain can be evaluated from the measurement of Δd/Δd s based on the Δd/Δd s -ε relationship. A creep curve under any creep condition can readily be calculated by creep data of the steel. Combining these information one can assess residual life of a structural component made of the steel. (author)

  3. Influenced prior loading on the creep fatigue damage accumulation of heat resistant steels

    International Nuclear Information System (INIS)

    Kloos, K.H.; Granacher, J.; Scholz, A.

    1990-01-01

    On two heat resistant power plant steels the influence of prior strain cycling on the creep rupture behaviour and the influence of prior creep loading on the strain cycling behaviour is investigated. These influences concern the number of cycles to failure and the rupture time being the reference values of the generalized damage accumulation rule and they are used for a creep fatigue analysis of the results of long term service-type strain cycling tests. (orig.) [de

  4. Analysis and prevention of human failure in nuclear power plants

    International Nuclear Information System (INIS)

    Liu Xinshuan

    2001-01-01

    Based on the performances in Daya Bay Nuclear Power Plant and the common experience from the world nuclear industry, the features and usual kinds of human failures in nuclear power plants are highlighted and the prominent factors on the personal, external and decision problems which might cause the human failures are analyzed. Effective preventive measures have been proposed respectively. Some successful human-failure-prevention practices applied in the Daya Bay Nuclear Power Plant are illustrated specifically

  5. Compaction creep of sands due to time-dependent grain failure : Effects of chemical environment, applied stress, and grain size

    NARCIS (Netherlands)

    Brzesowsky, R. H.; Hangx, S. J. T.|info:eu-repo/dai/nl/30483579X; Brantut, N.; Spiers, C. J.|info:eu-repo/dai/nl/304829323

    2014-01-01

    Time-dependent brittle creep plays a role in controlling compaction of sands and sandstones under upper crustal conditions, influencing phenomena such as production-induced reservoir compaction, surface subsidence, and induced seismicity. Brittle creep also plays a role in determining the mechanical

  6. RUPTHER - an original experimental approach for creep failure study of RPV steel

    International Nuclear Information System (INIS)

    Sainte Catherine, C.; Mongabure, Ph.; Cotoni, V.; Nicolas, L.; Devos, J.

    1998-01-01

    Rupter (Rupture Under Thermal Conditions) experiment is designed in order to get validated models for the degradation of RPV (Reactor Pressure Vessel) bottom head in case of a severe accident with corium flow. A simple experimental testing device has been designed in order to perform realistic thermo-mechanical loading on a cylinder. It is externally heated in its central part by induction (max. 1300 deg C) giving an axial thermal gradient. The cylinder is then mechanically loaded by internal pressure (max. 100 bars) until failure occurrence. (authors)

  7. Creep properties of heat-resistant superalloys for nuclear plants in helium

    International Nuclear Information System (INIS)

    Shimizu, Shigeki; Satoh, Keisuke; Matsuda, Shozo; Murase, Hirokazu; Fujioka, Junzo.

    1979-01-01

    In order to estimate the creep and rupture strengths of candidate alloys for the intermediate heat exchanger of VHTR, creep and stress rupture tests in impure helium were conducted on Hastelloy X, Inconel 617, Inconel 625, Incoloy 800 and Incoloy 807 at 900 0 C. The results were discussed in comparison with those in air and the alloys were examined from the point of view of the elevated temperature structural design. The main results obtained are summarized as follows: (1) No appreciable decrease in creep and rupture strengths in helium as compared with those in air is observed on Hastelloy X and Inconel 625. On the contrary, the creep and rupture strengths of Inconel 617 in helium decrease slightly as compared with those in air. In the case of Incoloy 807, the creep strength to cause 1 percent total strain and that to initiate secondary creep increase remarkably in helium as compared with those in air. However, the creep strength to cause initiation of tertiary creep and the rupture strength in helium remarkably decrease as compared with those in air. (2) The order of magnitude of the S 0 value for each material in helium is as follows; Hastelloy X > Inconel 617 > Incoloy 807 > Inconel 625 > Incoloy 800 Meanwhile, that of the S sub(t) value in helium is; Inconel 617 > Hastelloy X > Incoloy 807 > Inconel 625 > Incoloy 800. (author)

  8. Investigation of valve failure problems in LWR power plants

    International Nuclear Information System (INIS)

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

  9. Experimental study of a macrocrack propagation in a concrete specimen subjected to creep loading

    Science.gov (United States)

    Rossi, P.; Boulay, C.; Tailhan, J.-L.; Martin, E.

    2013-07-01

    Structures managers need a better prediction of the delayed failure of concrete structures, especially for very important structures like nuclear power plant encasement. Sustained loadings at high level (above 75% of loading capacity of the structure), can lead to structure failure after some time. In this study, a series of 4-point bending tests were performed in order to characterize the creep behaviour of pre-cracked beams under high load level. The specimens were made of normal strength concrete. A power law relationship is observed between the secondary deflection creep rate and the failure time. It is also shown that when crack propagation occurs during the creep loading, the creep deflection rate increases with the creep loading level and with the crack propagation rate.

  10. Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

    Science.gov (United States)

    Magnusson, Per; Chen, Jiachao; Hoffelner, Wolfgang

    2009-12-01

    Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm3 were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 104 hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.

  11. A Critical Analysis of the Conventionally Employed Creep Lifing Methods.

    Science.gov (United States)

    Abdallah, Zakaria; Gray, Veronica; Whittaker, Mark; Perkins, Karen

    2014-04-29

    The deformation of structural alloys presents problems for power plants and aerospace applications due to the demand for elevated temperatures for higher efficiencies and reductions in greenhouse gas emissions. The materials used in such applications experience harsh environments which may lead to deformation and failure of critical components. To avoid such catastrophic failures and also increase efficiency, future designs must utilise novel/improved alloy systems with enhanced temperature capability. In recognising this issue, a detailed understanding of creep is essential for the success of these designs by ensuring components do not experience excessive deformation which may ultimately lead to failure. To achieve this, a variety of parametric methods have been developed to quantify creep and creep fracture in high temperature applications. This study reviews a number of well-known traditionally employed creep lifing methods with some more recent approaches also included. The first section of this paper focuses on predicting the long-term creep rupture properties which is an area of interest for the power generation sector. The second section looks at pre-defined strains and the re-production of full creep curves based on available data which is pertinent to the aerospace industry where components are replaced before failure.

  12. Creep properties of heat-resistant superalloys for nuclear plants in helium

    International Nuclear Information System (INIS)

    Shimizu, Shigeki; Satoh, Keisuke; Honda, Yoshio; Matsuda, Shozo; Murase, Hirokazu

    1979-01-01

    Creep properties of candidate superalloys for VHTR components in a helium environment at both temperatures of 800 0 C and 900 0 C were compared with those of the same alloys in the atmospheric condition, and the superalloys were contrasted with each other from the viewpoint of high temperature structural design. At 800 0 C, no significant effect of a helium environment on creep properties of the superalloys is observed. At 900 0 C, however, creep strength of Inconel 617, Incoloy 800 and Incoloy 807 in the helium environment decrease more than in the atmospheric environment. In Hastelloy X and Inconel 625, there is no significant difference between creep strengths in helium and those in the atmospheric condition. Concerning So and St values in helium at 900 0 C, Inconel 617 and Hastelloy X are clearly superior to other superalloys. (author)

  13. Evaluation of creep-fatigue life prediction methods for low-carbon/nitrogen-added SUS316

    International Nuclear Information System (INIS)

    Takahashi, Yukio

    1998-01-01

    Low-carbon/medium nitrogen 316 stainless steel called 316FR is a principal candidate for the high-temperature structural materials of a demonstration fast reactor plant. Because creep-fatigue damage is a dominant failure mechanism of the high-temperature materials subjected to thermal cycles, it is important to establish a reliable creep-fatigue life prediction method for this steel. Long-term creep tests and strain-controlled creep-fatigue tests have been conducted at various conditions for two different heats of the steel. In the constant load creep tests, both materials showed similar creep rupture strength but different ductility. The material with lower ductility exhibited shorter life under creep-fatigue loading conditions and correlation of creep-fatigue life with rupture ductility, rather than rupture strength, was made clear. Two kinds of creep-fatigue life prediction methods, i.e. time fraction rule and ductility exhaustion method were applied to predict the creep-fatigue life. Accurate description of stress relaxation behavior was achieved by an addition of 'viscous' strain to conventional creep strain and only the latter of which was assumed to contribute to creep damage in the application of ductility exhaustion method. The current version of the ductility exhaustion method was found to have very good accuracy in creep-fatigue life prediction, while the time fraction rule overpredicted creep-fatigue life as large as a factor of 30. To make a reliable estimation of the creep damage in actual components, use of ductility exhaustion method is strongly recommended. (author)

  14. Residual creep life assessment by change of martensitic lath structure in modified 9Cr-1Mo steels

    International Nuclear Information System (INIS)

    Sawada, Kota; Takeda, Masaaki; Maruyama, Kouichi; Komine, Ryuji; Nagae, Yuji

    1998-01-01

    Mod.9Cr-1Mo steel has a martensitic lath structure. Recovery of the lath structure takes place in the course of creep. Microstructural degradation due to the recovery results in the acceleration of creep rate and the subsequent failure of a specimen. Change of lath width during creep of the steel was quantitatively investigated to propose a residual life assessment methodology based on the recovery process. Since the steel was tempered at 1053 K, the lath structure is thermally stable at the testing temperatures (848 K - 923 K). However, recovery of lath structure readily takes place during creep, indicating that the recovery is induced by creep deformation. Lath width d increases with creep strain and saturates to a value d s determined by creep stress. The increase of d is faster at a higher stress and temperature. A normalized change in lath width, Δd/Δd s , was introduced to explain the variation of lath growth rate with creep stress and temperature. Δd is the change in lath width from the initial value d 0 , and Δd s is the difference between d s , and d 0 . Δd/Δd s is uniquely related to creep strain ε and the relationship is independent of creep stress as well as creep temperature. This Δd/Δd s -ε relationship obtained by an accelerated creep test at a higher temperature or stress is applicable to any creep condition including service conditions of engineering plants. Creep strain can be evaluated from the measurement of Δd/Δd s based on the Δd/Δd s -ε relationship. A creep curve under any creep condition can readily be calculated by creep data of the steel. Combining these information one can assess residual life of a structural component made of the steel. (author)

  15. Tensile Creep and Fatigue of Sylramic-iBN Melt-Infiltrated SiC Matrix Composites: Retained Properties, Damage Development, and Failure Mechanisms

    Science.gov (United States)

    Morscher, Greg; Gowayed, yasser; Miller, Robert; Ojard, Greg; Ahmad, Jalees; Santhosh, Unni; John, Reji

    2008-01-01

    An understanding of the elevated temperature tensile creep, fatigue, rupture, and retained properties of ceramic matrix composites (CMC) envisioned for use in gas turbine engine applications are essential for component design and life-prediction. In order to quantify the effect of stress, time, temperature, and oxidation for a state-of-the-art composite system, a wide variety of tensile creep, dwell fatigue, and cyclic fatigue experiments were performed in air at 1204 C for the SiC/SiC CMC system consisting of Sylramic-iBN SiC fibers, BN fiber interphase coating, and slurry-cast melt-infiltrated (MI) SiC-based matrix. Tests were either taken to failure or interrupted. Interrupted tests were then mechanically tested at room temperature to determine the residual properties. The retained properties of most of the composites subjected to tensile creep or fatigue were usually within 20% of the as-produced strength and 10% of the as-produced elastic modulus. It was observed that during creep, residual stresses in the composite are altered to some extent which results in an increased compressive stress in the matrix upon cooling and a subsequent increased stress required to form matrix cracks. Microscopy of polished sections and the fracture surfaces of specimens which failed during stressed-oxidation or after the room-temperature retained property test was performed on some of the specimens in order to quantify the nature and extent of damage accumulation that occurred during the test. It was discovered that the distribution of stress-dependent matrix cracking at 1204 C was similar to the as-produced composites at room temperature; however, matrix crack growth occurred over time and typically did not appear to propagate through thickness except at final failure crack. Failure of the composites was due to either oxidation-induced unbridged crack growth, which dominated the higher stress regime (> 179 MPa) or controlled by degradation of the fibers, probably caused by

  16. Failure cause and failure rate evaluation on pumps of BWR plants in PSA. Hypothesis testing for typical or plant specific failure rate of pumps

    International Nuclear Information System (INIS)

    Sanada, Takahiro; Nakamura, Makoto

    2009-01-01

    In support of domestic nuclear industry effort to gather and analyze failure data of components concerning nuclear power plants, Nuclear Information Archives (NUCIA) are published for useful information to help PSA. This report focuses on NUCIA pertaining to pumps in domestic nuclear power plants, and provides the reliable estimation on failure rate of pumps resulting from failure cause analysis and hypothesis testing of classified and plant specific failure rate of pumps for improving quality in PSA. The classified and plant specific failure rate of pumps are estimated by analyzing individual domestic nuclear power plant's data of 26 Boiling Water Reactors (BWRs) concerning functionally structurally classified pump failures reported from beginning of commercial operation to March 31, 2007. (author)

  17. Failure investigation of super heater tubes of coal fired power plant

    Directory of Open Access Journals (Sweden)

    A.K. Pramanick

    2017-10-01

    Full Text Available Cause of failure of two adjacent super heater tubes made of Cr-Mo steel of a coal based 60 MW thermal power plant has been portrayed in present investigation. Oxide deposits were found on internal surface of tubes. Deposits created significant resistance to heat transfer and resulted in undesirable rise in component temperature. This situation, in turn, aggravated the condition of gas side that was exposed to high temperature. Localized heating coarsened carbides as well as propelled precipitation of new brittle phases along grain boundary resulting in embrittlement of tube material. Continuous exposure to high temperature softened the tube material and tube wall was thinned down with bulging toward outside. Creep void formation along grain boundary was observed and steered intergranular cracking. All these effects contributed synergistically and tubes were failed ultimately due to overload under high Hoop stress.

  18. Interfacial failure in dissimilar weld joint of high boron 9% chromium steel and nickel-based alloy under high-temperature creep condition

    Energy Technology Data Exchange (ETDEWEB)

    Matsunaga, Tetsuya, E-mail: MATSUNAGA.Tetsuya@nims.go.jp; Hongo, Hiromichi, E-mail: HONGO.Hiromichi@nims.go.jp; Tabuchi, Masaaki, E-mail: TABUCHI.Masaaki@nims.go.jp

    2017-05-17

    The advanced ultra-supercritical (A-USC) power generation system is expected to become the next-generation base-load power station in Japan. Dissimilar weld joints between high-Cr heat-resistant steels and nickel-based alloys with a nickel-based filler metal (Alloy 82) will need to be adopted for this purpose. However, interfacial failure between the steels and weld metal has been observed under high-temperature creep conditions. Fractography and microstructure observations showed the failure initiated in a brittle manner by an oxide notch at the bottom of the U-groove. The fracture then proceeded along the bond line in a ductile manner with shallow dimples, where micro-Vickers hardness tests showed remarkable softening in the steel next to the bond line. In addition, the steel showed a much larger total elongation and reduction of area than the weld metal at low stresses under long-term creep conditions, leading to mismatch deformation at the interface. According to the results, it can be concluded that the interfacial failure between the 9Cr steels and Alloy 82 weld metal is initiated by an oxide notch and promoted by softening and the difference in the plasticity of the steels and weld metal.

  19. Failure investigation of a secondary super heater tube in a 140 MW thermal power plant

    Directory of Open Access Journals (Sweden)

    Atanu Saha

    2017-04-01

    Full Text Available This article describes the findings of a detailed investigation into the failure of a secondary super heater tube in a 140 MW thermal power plant. Preliminary macroscopic examinations along with visual examination, dimensional measurement and chemical analysis were carried out to deduce the probable cause of failure. In addition optical microscopy was a necessary supplement to understand the cause of failure. It was concluded that the tube had failed due to severe creep damage caused by high metal temperature during service. The probable causes of high metal temperature may be in sufficient flow of steam due to partial blockage, presence of thick oxide scale on ID surface, high flue gas temperature etc. rupture.

  20. Investigation of valve failure problems in LWR power plants

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

  1. Generic nuclear power plant component failure data bank

    International Nuclear Information System (INIS)

    Araujo Goes, A.G. de; Gibelli, S.M.O.

    1988-11-01

    This report consist in the development of a generic nuclear power plant component failure data bank. This data bank was implemented in a PC-XT microcomputer, IBM compatible, using the Open Access II program. Generic failure data tables for Westinghouse nuclear power plants and for general PWR power plants are presented. They are the final product of a research which included a preselection and a selection of data collected from the available sources in the library of CNEN (National Nuclear Energy Commission) and from the CIN/CNEN (Neclear Information Center). Futhermore, a proposal of evaluating models of average failure rates of pumps and valves are also presented. Through the electronic data bank one can easily have a generic view of failure rate ranges as well as failure models foe a certain component. It is very importante to develop procedures to collect and store generic failure data that can be quickly accessed, in order to update the Probabilistic Safety Study of Angra-1 and to used in studies which may have component failures of nuclear power plant safety systems. In the future, data specialization can be achieved by means of statistical calculations involving specific data collected from the operational experience of Angra-1 nuclear power plant and the generic data bank. (author) [pt

  2. Automated derivation of failure symptoms for diagnosis of nuclear plant

    International Nuclear Information System (INIS)

    Washio, T.; Kitamura, M.; Kotajima, K.; Sugiyama, K.

    1986-01-01

    A method of automated derivation of failure symptoms was developed as an approach to computer-aided failure diagnosis in a nuclear power plant. The automated derivation is realized using a knowledge representation called the semantic network (S-net). The purpose of this paper is to demonstrate the applicability of the S-net representation as a basic tool for deriving failure symptoms. If one can generate symptoms automatically, the computer-aided plant safety analysis and diagnosis can be performed easily by evaluating the influence of the failures on the whole plant. A specific description format called a 'network list' was introduced to implement the knowledge of the structure of the plant. The failure symptoms are derived automatically, based on the knowledge of the structure of the plant, using a PROLOG-based database handling system. This approach allows us to derive the failure symptoms of the plant without using conventional event-chain models (e.g. a cause-consequence tree) which are subject to human errors in their design and implementation. Applicability of this method was evaluated with a simulation model of the dynamics of the secondary system of a PWR. (author)

  3. Analysis of reactor trips involving balance-of-plant failures

    International Nuclear Information System (INIS)

    Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

    1986-01-01

    The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

  4. Radiation effects on time-dependent deformation: Creep and growth

    International Nuclear Information System (INIS)

    Simonen, E.P.

    1989-03-01

    Observations of irradiation creep strain as well as irradiation growth strain and related microstructures are reviewed and compared to mechanisms for radiation effects on time-dependent deformation. Composition, microstructure, stress and temperature affect irradiation creep less than thermal creep. Irradiation creep rates can often dominate thermal creep rates, particularly at low temperatures and low stresses. Irradiation creep mechanisms are classified in two general categories: (1) stress-induced preferential absorption and (2) climb-glide. In the former, creep results from dislocation climb, whereas in the latter, creep results from dislocation glide. The effects of irradiation creep on failure modes in nuclear environments are discussed. 53 refs., 18 figs., 1 tab

  5. Creep buckling of shells

    International Nuclear Information System (INIS)

    Stone, C.M.; Nickell, R.E.

    1977-01-01

    Because of the characteristics of LMFBR primary piping components (thin-walled, low pressure, high temperature), the designer must guard against creep buckling as a potential failure mode for certain critical regions, such as elbows, where structural flexibility and inelastic response may combine to concentrate deformation and cause instability. The ASME Boiler and Pressure Vessel Code, through its elevated temperature Code Case 1592 (Section III, Division 1) provides design rules for Class 1 components aimed at preventing creep buckling during the design life. A similar set of rules is being developed for Class 2 and 3 components at this time. One of the original concepts behind the creep buckling rules was that the variability in creep properties (especially due to the effects of prior heat treatment), the uncertainty about initial imperfections, and the lack of confirmed accuracy of design analysis meant that conservatism would be difficult to assure. As a result, a factor of ten on service life was required (i.e. analysis must show that, under service conditions that extrapolate the life of the component by ten times, creep buckling does not occur). Two obvious problems with this approach are that: first, the creep behavior must also be extrapolated (since most creep experiments are terminated at a small fraction of the design life, extrapolation of creep data is already an issue, irrespective of the creep buckling question); second the nonlinear creep analysis, which is very nearly prohibitively expensive for design life histograms, becomes even more costly. Analytical results for an aluminum cylindrical shell subjected to axial loads at elevated temperatures are used to examine the supposed equivalence of two types of time-dependent buckling safety factors - a factor of ten on service life and a factor of 1.5 on loading

  6. The effect of microstructural stability on long-term creep behaviour of 11 %Cr steels for steam power plants with operating steam temperatures up to 650 C

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Y.; Scholz, A.; Berger, C. [Technische Univ. Darmstadt (DE). Inst. fuer Werkstoffkunde (IFW); Kauffmann, F.; Maile, K. [Stuttgart Univ. (DE). Materialpruefungsanstalt (MPA); Mayer, K.H. [Alstom Power, Nuernberg (Germany)

    2010-07-01

    The investigations of advanced ferritic/martensitic steels for 650 C power plant components focus on the improvement of high-temperature creep properties with respect to chemical composition. This study deals with the development of new heat resistant 11-12%Cr ferritic-martensitic steels with sufficient creep and oxidation resistance up to 650 application by using basic principles and concepts of physical metallurgy. The highest creep strength could be achieved with a 0.04% Nb alloyed 11%CrWCoMoVB melt, which is in addition alloyed with a higher C and B content as well as with lower W and Co portions. The microstructure evolution during creep of this newly developed steel was investigated in comparison to a sister alloy which comprises 0.06% Ta instead of the Nb. (orig.)

  7. Creep/Stress Rupture Behavior and Failure Mechanisms of Full CVI and Full PIP SiC/SiC Composites at Elevated Temperatures in Air

    Science.gov (United States)

    Bhatt, R. T.; Kiser, J. D.

    2017-01-01

    SiC/SiC composites fabricated by melt infiltration are being considered as potential candidate materials for next generation turbine components. However these materials are limited to 2400 F application because of the presence of residual silicon in the SiC matrix. Currently there is an increasing interest in developing and using silicon free SiC/SiC composites for structural aerospace applications above 2400 F. Full PIP or full CVI or CVI + PIP hybrid SiC/SiC composites can be fabricated without excess silicon, but the upper temperature stress capabilities of these materials are not fully known. In this study, the on-axis creep and rupture properties of the state-of-the-art full CVI and full PIP SiC/SiC composites with Sylramic-iBN fibers were measured at temperatures to 2700 F in air and their failure modes examined. In this presentation creep rupture properties, failure mechanisms and upper temperature capabilities of these two systems will be discussed and compared with the literature data.

  8. Trend analysis of cables failure events at nuclear power plants

    International Nuclear Information System (INIS)

    Fushimi, Yasuyuki

    2007-01-01

    In this study, 152 failure events related with cables at overseas nuclear power plants are selected from Nuclear Information Database, which is owned by The Institute of Nuclear Safety System, and these events are analyzed in view of occurrence, causal factor, and so on. And 15 failure events related with cables at domestic nuclear power plants are selected from Nuclear Information Archives, which is owned by JANTI, and these events are analyzed by the same manner. As a result of comparing both trends, it is revealed following; 1) A cable insulator failure rate is lower at domestic nuclear power plants than at foreign ones. It is thought that a deterioration diagnosis is performed broadly in Japan. 2) Many buried cables failure events have been occupied a significant portion of cables failure events during work activity at overseas plants, however none has been occurred at domestic plants. It is thought that sufficient survey is conducted before excavating activity in Japan. 3) A domestic age related cables failure rate in service is lower than the overseas one and domestic improper maintenance rate is higher than the overseas one. Maintenance worker' a skill improvement is expected in order to reduce improper maintenance. (author)

  9. Diagnostic technology of PWR plant equipment failures

    International Nuclear Information System (INIS)

    Nakamura, Tetsuo; Tanaka, Mamoru; Okamachi, Masao; Taguchi, Shozo; Nagashima, Kazuhiro; Ishikawa, Satoshi

    1985-01-01

    To confirm the soundness of the important facilities in a nuclear power plant contributes to the reliability of the plant operations and improvement of its operation rate. For this purpose, the following diagnostic techniques have been developed. (1). Vibration and loose parts monitoring: Detection of abnormal structural vibrations in the reactor, estimation of its mode, detection of loose parts in the primary system, and estimation of the position and energy of their collisions against the reactor vessel or the like. (2). Valve leak monitoring: Detection of leaks from primary valves in the primary cooling boundary, such as the pressurizer relief valve and safety valve, and estimation of the form of the leaks. (3). Detector noise response diagnosis: Diagnosis of degradation of principal process detectors during plant operation. Furthermore, a diagnostic system incorporating the above diagnostic technology applicable to actual plants has been experimentally manufactured and successfully verified. (author)

  10. Cast Steels for Creep-Resistant Parts Used in Heat Treatment Plants

    Directory of Open Access Journals (Sweden)

    Drotlew A.

    2012-12-01

    Full Text Available Creep-resistant parts of heat treatment furnaces are in most cases made from high-alloyed chromium-nickel and nickel-chromium iron alloys, both cast and wrought. This paper presents the types of casting alloys used for this particular purpose, since the majority of furnace components are made by the casting process. Standards were cited which give symbols of alloy grades used in technical specifications by the domestic industry. It has been indicated that castings made currently are based on a wider spectrum of the creep-resistant alloy grades than the number of alloys covered by the standards. Alloy grades recommended by the technical literature for individual parts of the furnace equipment were given. The recommendations reflect both the type of the technological process used and the technical tasks performed by individual parts of the furnace equipment. Comments were also made on the role of individual alloying elements in shaping the performance properties of castings.

  11. Cast Steels for Creep-resistant Parts Used in Heat Treatment Plants

    Directory of Open Access Journals (Sweden)

    A. Drotlew

    2012-12-01

    Full Text Available Creep-resistant parts of heat treatment furnaces are in most cases made from high-alloyed chromium-nickel and nickel-chromium ironalloys, both cast and wrought. This paper presents the types of casting alloys used for this particular purpose, since the majority of furnace components are made by the casting process. Standards were cited which give symbols of alloy grades used in technical specifications by the domestic industry. It has been indicated that castings made currently are based on a wider spectrum of the creep-resistant alloy grades than the number of alloys covered by the standards. Alloy grades recommended by the technical literature for individual parts of the furnace equipment were given. The recommendations reflect both the type of the technological process used and the technical tasks performed by individual parts of the furnace equipment. Comments were also made on the role of individual alloying elements in shaping the performance properties of castings.

  12. Further evaluation of creep-fatigue life prediction methods for low-carbon nitrogen-added 316 stainless steel

    International Nuclear Information System (INIS)

    Takahashi, Y.

    1999-01-01

    Low-carbon, medium-nitrogen 316 stainless steel is a principal candidate for a main structural material of a demonstration fast breeder reactor plant in Japan. A number of long-term creep tests and creep-fatigue tests have been conducted for four products of this steel. Two representative creep-fatigue life prediction methods, i.e., time fraction rule and ductility exhaustion method were applied. Total stress relaxation behavior was simulated well by an addition of a viscous strain term to the conventional (primary plus secondary) creep strain, but only the letter was assumed to contribute to creep damage in the ductility exhaustion method. The present ductility exhaustion approach was found to have very good accuracy in creep-fatigue life prediction for all materials tested, while the time fraction rule tended to overpredict failure life as large as a factor of 30. Discussion was made on the reason for this notable difference

  13. Operational failure at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Szatmary, Z.

    2003-01-01

    NPP failures are ranked according to the International Nuclear Event Scale. To rank the failure first a presentation of the pressurized water plant is given, including fuel change, maintenance cleaning and decontamination process. The failure has been produced with fuel bars in the cleaning container. Consequences of the failure are small, negligible environmental pollution with radioactive material and significant financial outfall due to inactivity of block 2. Among the causes of the failure are design errors of the cleaning container, the pure chemical approach to cleaning, unknown risk factors for some of the cleaning staff, cleaning container has not been verified and approved by responsible authorities, the prevalence of economic and quantitative indicators of the plant on the detriment of safety. Organisational factors also contribute to the possibility of nuclear failures. Specialist training in Germany (where the container has been produced) is significantly reduced, while in Hungary the political tide has caused a permanent change in the higher echelons of the plant management, where nuclear specialists were not included. (Gy.M.)

  14. Probabilistic residual life assessment of high temperature pipings in nuclear power plants against creep fatigue damage: final report

    International Nuclear Information System (INIS)

    Gupta, Sanjay

    2014-02-01

    Residual life assessment of components of nuclear power plants is essential for their operational safety, reliability and financial viability. The high risks involved in the event of failures in nuclear power plants have led to the development of design philosophies that incorporate extreme conservatism in design. The implications of such conservatism in design leads to more frequent maintenance operations than necessary

  15. Development of Probability Evaluation Methodology for High Pressure/Temperature Gas Induced RCS Boundary Failure and SG Creep Rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Chul; Hong, Soon Joon; Lee, Jin Yong; Lee, Kyung Jin; Lee, Kuh Hyung [FNC Tech. Co., Seoul (Korea, Republic of)

    2008-04-15

    Existing MELCOR 1.8.5 model was improved in view of severe accident natural circulation and MELCOR 1.8.6 input model was developed and calculation sheets for detailed MELCOR 1.8.6 model were produced. Effects of natural circulation modeling were found by simulating SBO accident by comparing existing model with detailed model. Major phenomenon and system operations which affect on natural circulation by high temperature and high pressure gas were investigated and representative accident sequences for creep rupture model of RCS pipeline and SG tube were selected.

  16. Importance of creep failure of hard rock in the near field of a nuclear-waste repository

    International Nuclear Information System (INIS)

    Blacic, J.D.

    1981-01-01

    Potential damage resulting from slow creep deformation intuitively seems unlikely for a high-level nuclear waste repository excavated in hard rock. However, recent experimental and modeling results indicate that the processes of time-dependent microcracking and water-induced stress corrosion can lead to significant reductions in strength and alteration of other key rock properties in the near-field region of a repository. We review the small data base supporting these conclusions and stress the need for an extensive laboratory program to obtain the new data that will be required for design of a repository

  17. Development of creep damage assessment system for aged thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, Isamu [IshikawaJima-Harima Heavy Industries Co., Ltd., Tokyo (Japan); Umaki, Hideo [Ishikawajima-Harima Heavy Industries Co., Ltd., Tokyo (Japan); Nishida, Hidetalca [The Chugoku Electric Power Co., Inc., Hiroshima (Japan); Yamaguchi, Hiroshi [The Chugoku Electric Power Co., Inc., Hiroshima (Japan)

    1998-12-31

    IHI has developed the Creep Damage Assessment System to identify voids by processing an image observed by a small laser microscope with an advanced image processing technique jointly with Chugoku Electric Power Co., Inc. The result can be obtained immediately on the spot. Application tests of the system at the Unit No.3 boiler of the Kudamatsu Power Station showed good operability, adaptability to the environment, and accuracy. The new system can easily indicate damage conditions in parts during the periodical inspection, allowing rapid maintenance. Time reduction required for assessment and increased reliability of equipment can be also achieved. (orig.)

  18. Development of creep damage assessment system for aged thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, Isamu [IshikawaJima-Harima Heavy Industries Co., Ltd., Tokyo (Japan); Umaki, Hideo [Ishikawajima-Harima Heavy Industries Co., Ltd., Tokyo (Japan); Nishida, Hidetalca [The Chugoku Electric Power Co., Inc., Hiroshima (Japan); Yamaguchi, Hiroshi [The Chugoku Electric Power Co., Inc., Hiroshima (Japan)

    1999-12-31

    IHI has developed the Creep Damage Assessment System to identify voids by processing an image observed by a small laser microscope with an advanced image processing technique jointly with Chugoku Electric Power Co., Inc. The result can be obtained immediately on the spot. Application tests of the system at the Unit No.3 boiler of the Kudamatsu Power Station showed good operability, adaptability to the environment, and accuracy. The new system can easily indicate damage conditions in parts during the periodical inspection, allowing rapid maintenance. Time reduction required for assessment and increased reliability of equipment can be also achieved. (orig.)

  19. Image-based creep-fatigue damage mechanism investigation of Alloy 617 at 950 °C

    Energy Technology Data Exchange (ETDEWEB)

    Tahir, Fraaz; Dahire, Sonam; Liu, Yongming, E-mail: yongming.liu@asu.edu

    2017-01-02

    Alloy 617 is a primary candidate material to be used in the next generation of nuclear power plants. Structural materials for these plants are expected to undergo creep and fatigue at temperatures as high as 950 °C. This study uses a hybrid-control creep-fatigue loading profile, as opposed to the traditional strain-controlled loading, to generate creep dominated failure. Qualitative and quantitative image analysis through SEM, EDS, and EBSD, is used to show that hybrid control testing is capable of producing creep dominated failure and that time fraction approach is not a valid indicator of creep or fatigue dominated damage. The focus of image analysis is on surface fatigue cracks and internal creep voids. A creep-fatigue damage interaction diagram based on these micro-scale features is plotted. It is shown that the classical time fraction approach suggested by the ASME code does not agree with the experimental findings and has a poor correlation with observed microscale damage features. A new definition of creep damage fraction based on an effective hold time is found to correlate well with the micro-scale image analysis.

  20. Detection of sensor failures in nuclear plants using analytic redundancy

    International Nuclear Information System (INIS)

    Kitamura, M.

    1980-01-01

    A method for on-line, nonperturbative detection and identification of sensor failures in nuclear power plants was studied to determine its feasibility. This method is called analytic redundancy, or functional redundancy. Sensor failure has traditionally been detected by comparing multiple signals from redundant sensors, such as in two-out-of-three logic. In analytic redundancy, with the help of an assumed model of the physical system, the signals from a set of sensors are processed to reproduce the signals from all system sensors

  1. Influence of creep ductility on creep-fatigue behaviour of 20%Cr/25%Ni/Nb stainless steel

    International Nuclear Information System (INIS)

    Gladwin, D.; Miller, D.A.

    1985-01-01

    The influence of creep ductility on creep-fatigue endurance of 20%Cr/25%Ni/Nb stainless steel has been examined. In order to induce different creep ductilities in the 20/25/Nb stainless steel, three different thermo-mechanical routes were employed. These resulted in a range of ductilities (3-36%) being obtained at the strain rates of interest. Strain controlled slow-fast creep-fatigue cycles were used with strain rates of 10 -6 s -1 , 10 -7 s -1 in tension and 10 -3 s -1 in compression. It was found that creep ductility strongly influenced the creep-fatigue endurance of the 20/25/Nb stainless steel. When failure was creep dominated endurance was found to be directly proportional to the creep ductility. A ductility exhaustion model has been used to successfully predict creep-fatigue endurance when failure was creep dominated. (author)

  2. Failure data collection from a Swedish nuclear power plant

    International Nuclear Information System (INIS)

    Andersson, B.; Bhattacharyya, A.; Hilding, S.

    1975-01-01

    The Swedish nuclear utilities have formed a joint working group in the field of reliability data of thermal power plants, nuclear and fossil fuelled. The primary task of the working group is to create a standard procedure of collecting failure data from the Swedish nuclear power plants in operation. The failure data will be stored in a joint data bank. A first test collection of such data has been implemented on Oskarshamn I, and the experience with this work is discussed in this report. Reliability analysis of an engineering system is based on the availability of pertinent information on the system components. Right from the beginning within the Swedish nuclear industry the consensus has been that such data can be suitably obtained by monitoring the operating power stations. This has led to a co-operative arrangement between the vendor, ASEA-ATOM and a utility, Oskarshamnsverkets Kraftgrupp AB (OKG) to utilize information from component malfunctions in the reliability analysis. The utility prepares component failure reports which are sent to the vendor for further treatment. Experience gathered to date indicates that this arrangement is effective although many persons are involved in this process of information transmittal. The present set-up is flexible enough to accommodate necessary changes in view of problems which arise now and then in monitoring a complex system like a nuclear power station. This report briefly describes the structure of the failure data collection system. The way in which the raw data collection is done in the station by the owner and the subsequent data processing by the vendor is discussed. A brief status report of the information collected since 1971 is given. It can be concluded that valuable reliability data can be obtained by monitoring component failure reports from an operating power plant. Two requirements are, however, that all the parties involved in the arrangement follow given instructions carefully and that the assumed

  3. Penstock failure detection system at the 'Valsan' hydro power plant

    International Nuclear Information System (INIS)

    Georgescu, A M; Coşoiu, C I; Alboiu, N; Hlevca, D; Tataroiu, R; Popescu, O

    2012-01-01

    'Valsan' is a small Hydro Power Plant, 5 MW, situated at about 160 km north of Bucharest, Romania, on the small 'Valsan' river in a remote mountainous area. It is equipped with a single Francis turbine. The penstock is located in the access shaft of the HPP. 'Hidroelectrica', the Romanian company that operates the HPP, was trying to implement a remote penstock failure detection system. Starting from a classic hydraulic problem, the authors of the paper derived a method for failure detection and localization on the pipe. The method assumes the existence of 2 flow meters and 2 pressure transducers at the inlet and outlet of the pressurized pipe. Calculations have to be based on experimental values measured in a permanent regime for different values of the flow rate. The method was at first tested on a pipe, in the Hydraulic Laboratory of the Technical University of Civil Engineering Bucharest. Pipe failure was modelled by opening of a valve on a tee branch of the analyzed pipe. Experimental results were found to be in good agreement with theoretical ones. The penstock of the 'Valsan' HPP, was modelled in EPANET, in order to: i) test the method at a larger scale; ii) get the right flow and pressure transducers that are needed to implement it. At the request of 'Hidroelectrica' a routine that computes the efficiency of the turbine was added to the monitoring software. After the system was implemented, another series of measurements were performed at the site in order to validate it. Failure was modelled by opening an existing valve on a branch of the penstock. Detection of the failure was correct and almost instantaneous, while failure location was accurate within 5% of the total penstock length.

  4. Penstock failure detection system at the "Valsan" hydro power plant

    Science.gov (United States)

    Georgescu, A. M.; Coşoiu, C. I.; Alboiu, N.; Hlevca, D.; Tataroiu, R.; Popescu, O.

    2012-11-01

    "Valsan" is a small Hydro Power Plant, 5 MW, situated at about 160 km north of Bucharest, Romania, on the small "Valsan" river in a remote mountainous area. It is equipped with a single Francis turbine. The penstock is located in the access shaft of the HPP. "Hidroelectrica", the Romanian company that operates the HPP, was trying to implement a remote penstock failure detection system. Starting from a classic hydraulic problem, the authors of the paper derived a method for failure detection and localization on the pipe. The method assumes the existence of 2 flow meters and 2 pressure transducers at the inlet and outlet of the pressurized pipe. Calculations have to be based on experimental values measured in a permanent regime for different values of the flow rate. The method was at first tested on a pipe, in the Hydraulic Laboratory of the Technical University of Civil Engineering Bucharest. Pipe failure was modelled by opening of a valve on a tee branch of the analyzed pipe. Experimental results were found to be in good agreement with theoretical ones. The penstock of the "Valsan" HPP, was modelled in EPANET, in order to: i) test the method at a larger scale; ii) get the right flow and pressure transducers that are needed to implement it. At the request of "Hidroelectrica" a routine that computes the efficiency of the turbine was added to the monitoring software. After the system was implemented, another series of measurements were performed at the site in order to validate it. Failure was modelled by opening an existing valve on a branch of the penstock. Detection of the failure was correct and almost instantaneous, while failure location was accurate within 5% of the total penstock length.

  5. Common Cause Failure Analysis for the Digital Plant Protection System

    International Nuclear Information System (INIS)

    Kagn, Hyun Gook; Jang, Seung Cheol

    2005-01-01

    Safety-critical systems such as nuclear power plants adopt the multiple-redundancy design in order to reduce the risk from the single component failure. The digitalized safety-signal generation system is also designed based on the multiple-redundancy strategy which consists of more redundant components. The level of the redundant design of digital systems is usually higher than those of conventional mechanical systems. This higher redundancy would clearly reduce the risk from the single failure of components, but raise the importance of the common cause failure (CCF) analysis. This research aims to develop the practical and realistic method for modeling the CCF in digital safety-critical systems. We propose a simple and practical framework for assessing the CCF probability of digital equipment. Higher level of redundancy causes the difficulty of CCF analysis because it results in impractically large number of CCF events in the fault tree model when we use conventional CCF modeling methods. We apply the simplified alpha-factor (SAF) method to the digital system CCF analysis. The precedent study has shown that SAF method is quite realistic but simple when we consider carefully system success criteria. The first step for using the SAF method is the analysis of target system for determining the function failure cases. That is, the success criteria of the system could be derived from the target system's function and configuration. Based on this analysis, we can calculate the probability of single CCF event which represents the CCF events resulting in the system failure. In addition to the application of SAF method, in order to accommodate the other characteristics of digital technology, we develop a simple concept and several equations for practical use

  6. AsHSP17, a creeping bentgrass small heat shock protein modulates plant photosynthesis and ABA-dependent and independent signalling to attenuate plant response to abiotic stress.

    Science.gov (United States)

    Sun, Xinbo; Sun, Chunyu; Li, Zhigang; Hu, Qian; Han, Liebao; Luo, Hong

    2016-06-01

    Heat shock proteins (HSPs) are molecular chaperones that accumulate in response to heat and other abiotic stressors. Small HSPs (sHSPs) belong to the most ubiquitous HSP subgroup with molecular weights ranging from 12 to 42 kDa. We have cloned a new sHSP gene, AsHSP17 from creeping bentgrass (Agrostis stolonifera) and studied its role in plant response to environmental stress. AsHSP17 encodes a protein of 17 kDa. Its expression was strongly induced by heat in both leaf and root tissues, and by salt and abscisic acid (ABA) in roots. Transgenic Arabidopsis plants constitutively expressing AsHSP17 exhibited enhanced sensitivity to heat and salt stress accompanied by reduced leaf chlorophyll content and decreased photosynthesis under both normal and stressed conditions compared to wild type. Overexpression of AsHSP17 also led to hypersensitivity to exogenous ABA and salinity during germination and post-germinative growth. Gene expression analysis indicated that AsHSP17 modulates expression of photosynthesis-related genes and regulates ABA biosynthesis, metabolism and ABA signalling as well as ABA-independent stress signalling. Our results suggest that AsHSP17 may function as a protein chaperone to negatively regulate plant responses to adverse environmental stresses through modulating photosynthesis and ABA-dependent and independent signalling pathways. © 2015 John Wiley & Sons Ltd.

  7. Damage development - effects of multiaxial loads on creep pore formation and fatigue damage in typical power plant steels. Final report

    International Nuclear Information System (INIS)

    Lenk, P.; Proft, D.; Kussmaul, A.; Fischer, R.

    2000-01-01

    The influence of multiaxial stress on creep pore formation in the steels 14MoV6-3 10CrMo9-10 and X10CrMoVNb9-1 was investigated on the basis of internal pressure experiments on smooth and notched hollow cylinders. In some cases, additional axial forces were applied in order to reproduce component-relevant multiaxial stresses. Local elongation during loading was investigated and analyzed using applied HT-DMS. When different strain levels had been reached, the samples were removed, analyzed, and characterized with regard to different damage parameters. It was found that no interdependence between the surface damage pattern and the deep damage pattern can be derived across the wall thickness if no information on the load state is available. Parallel to the experiments, inelastic FEA were carried out using the ABAQUS program system. The creep law of Graham and Walles was used for calculating flow and creep via a user-defined subroutine CREEP. The parameters of the creep law could be identified by adaptation to monoaxial creep tests [de

  8. Development of nondestructive evaluation of creep-fatigue damage in SUS316 stainless steel

    International Nuclear Information System (INIS)

    Shoji, Tetsuo; Kawahara, Tetsuji; Awano, Masakazu; Sato, Yasumoto

    1999-01-01

    Creep-fatigue is a fatal failure mode of high temperature structural materials. It is recognized that the law of linear damage, according to which creep-fatigue damage is expressed by the sum of the creep damage and the fatigue damage, is inadequate to evaluate creep-fatigue damage. This is due to the fact that the law of linear damage does not include the effect of interaction between the creep damage and the fatigue damage. Consequently, development of direct measurement of damage accumulation on the sample of interest is required for plant life evaluation. In this study, the induced current focusing potential drop (ICFPD) technique was used to evaluate the depth of small surface cracks in SUS316FR stainless steel which was subjected to creep-fatigue damage. It is shown that the potential drop increased during the micro-crack initiation and propagation. Correspondingly, the ICFPD technique applied to estimate micro-crack depth changes was used to accurately evaluate the residual life of creep-fatigue damaged structural materials. (author)

  9. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

    Schwab, P.R.

    1981-01-01

    Creep rupture is a long term failure mode in structural materials that occurs at high temperatures and moderate stress levels. The deterioration of the material preceding rupture, termed creep damage, manifests itself in the formation of small cavities on grain boundaries. To measure creep damage, sometimes uniaxial tests are performed, sometimes density measurements are made, and sometimes the grain boundary cavities are measured by microscopy techniques. The purpose of the present research is to explore a new method of measuring creep rupture, which involves measuring the curvature of eccentric pressurized tubes. Theoretical investigations as well as the design, construction, and operation of an experimental apparatus are included in this research

  10. Creep rupture behavior of unidirectional advanced composites

    Science.gov (United States)

    Yeow, Y. T.

    1980-01-01

    A 'material modeling' methodology for predicting the creep rupture behavior of unidirectional advanced composites is proposed. In this approach the parameters (obtained from short-term tests) required to make the predictions are the three principal creep compliance master curves and their corresponding quasi-static strengths tested at room temperature (22 C). Using these parameters in conjunction with a failure criterion, creep rupture envelopes can be generated for any combination of in-plane loading conditions and ambient temperature. The analysis was validated experimentally for one composite system, the T300/934 graphite-epoxy system. This was done by performing short-term creep tests (to generate the principal creep compliance master curves with the time-temperature superposition principle) and relatively long-term creep rupture tensile tests of off-axis specimens at 180 C. Good to reasonable agreement between experimental and analytical results is observed.

  11. Cleavage and creep fracture of rock salt

    International Nuclear Information System (INIS)

    Chan, K.S.; Munson, D.E.; Bodner, S.R.

    1996-01-01

    The dominant failure mechanism in rock salt at ambient temperature is either cleavage or creep fracture. Since the transition of creep fracture to cleavage in a compressive stress field is not well understood, failure of rock salt by cleavage and creep fracture is analyzed in this paper to elucidate the effect of stress state on the competition between these two fracture mechanisms. For cleavage fracture, a shear crack is assumed to cause the formation and growth of a symmetric pair of wing cracks in a predominantly compressive stress field. The conditions for wing-crack instability are derived and presented as the cleavage fracture boundary in the fracture mechanism map. Using an existing creep fracture model, stress conditions for the onset of creep fracture and isochronous failure curves of specified times-to-rupture are calculated and incorporated into the fracture mechanism map. The regimes of dominance by cleavage and creep fracture are established and compared with experimental data. The result indicates that unstable propagation of cleavage cracks occurs only in the presence of tensile stress. The onset of creep fracture is promoted by a tensile stress, but can be totally suppressed by a high confining pressure. Transition of creep fracture to cleavage occurs when critical conditions of stress difference and tensile stress for crack instability are exceeded

  12. Nondestructive evaluation of creep-fatigue damage: an interim report

    International Nuclear Information System (INIS)

    Nickell, R.E.

    1977-02-01

    In view of the uncertainties involved in designing against creep-fatigue failure and the consequences of such failures in Class 1 nuclear components that operate at elevated temperature, the possibility of intermittent or even continuous non-destructive examination of these components has been considered. In this interim report some preliminary results on magnetic force and ultrasonic evaluation of creep-fatigue damage in an LMFBR steam generator material are presented. These results indicate that the non-destructive evaluation of pure creep damage will be extremely difficult. A set of biaxial creep-fatigue tests that are designed to discriminate between various failure theories is also described

  13. Determination of safety margins for creep loaded primary circuit components in case of loss of pressure accidents of a HTR plant (SR 383)

    International Nuclear Information System (INIS)

    Breitbach, G.; Ahmed, K.; Over, H.; Schubert, F.; Nickel, H.

    1991-10-01

    The wall thickness of tubes in high temperature plants must be limited in such a way that pressure differences can not produce unadmissible deformations. For HTR (PNP-Plant) the postulated loss of secondary pressure is one of the considered accidents. In that case the tubes of the heat exchangers are loaded by the outher pressure of the primary coolant. So the risk of a creep collapse is given. The report is related to experimental and theoretical work for the creep collapse phenomena. HTR relevant tube geometries of the high temperature alloys NiCr22Co12Mo (INCONEL 617) and X10NiCrAlTi 32 20 (INCOLOY 800) were tested at temperatures of 900 and 950deg C and outer pressure loads in the range 40 bars. The experimental results are compared with theoretically computed values and discussed. The problem of safety margin is treated. Further, simplified procedures are developed for the estimation of the collapse time. (orig.) [de

  14. A real-time expert system for nuclear power plant failure diagnosis and operational guide

    International Nuclear Information System (INIS)

    Naito, N.; Sakuma, A.; Shigeno, K.; Mori, N.

    1987-01-01

    A real-time expert system (DIAREX) has been developed to diagnose plant failure and to offer a corrective operational guide for boiling water reactor (BWR) power plants. The failure diagnosis model used in DIAREX was systematically developed, based mainly on deep knowledge, to cover heuristics. Complex paradigms for knowledge representation were adopted, i.e., the process representation language and the failure propagation tree. The system is composed of a knowledge base, knowledge base editor, preprocessor, diagnosis processor, and display processor. The DIAREX simulation test has been carried out for many transient scenarios, including multiple failures, using a real-time full-scope simulator modeled after the 1100-MW(electric) BWR power plant. Test results showed that DIAREX was capable of diagnosing a plant failure quickly and of providing a corrective operational guide with a response time fast enough to offer valuable information to plant operators

  15. Influence of creep and cyclic oxidation in thermal barrier coatings

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, Philipp; Baeker, Martin; Roesler, Joachim [Technische Univ. Braunschweig (Germany). Inst. fuer Werkstoffe

    2012-01-15

    The lifetime of thermal barrier coating systems is limited by cracks close to the interfaces, causing delamination. To study the failure mechanisms, a simplified model system is analysed which consists of a bond-coat bulk material, a thermally grown oxide, and an yttria-stabilised zirconia topcoat. The stresses in the model system are calculated using a finite element model which covers the simulation of full thermal cycles, creep in all layers, and the anisotropic oxidation during dwelling. Creep in the oxide and the thermal barrier coating is varied with the use of different creep parameter sets. The influence of creep in the bondcoat is analysed by using two different bond-coat materials: fast creeping Fecralloy and slow creeping oxide dispersion strengthened MA956. It is shown that creep in the bondcoat influences the lifetime of the coatings. Furthermore, a fast creeping thermally grown oxide benefits the lifetime of the coating system. (orig.)

  16. Risk analysis of geothermal power plants using Failure Modes and Effects Analysis (FMEA) technique

    International Nuclear Information System (INIS)

    Feili, Hamid Reza; Akar, Navid; Lotfizadeh, Hossein; Bairampour, Mohammad; Nasiri, Sina

    2013-01-01

    Highlights: • Using Failure Modes and Effects Analysis (FMEA) to find potential failures in geothermal power plants. • We considered 5 major parts of geothermal power plants for risk analysis. • Risk Priority Number (RPN) is calculated for all failure modes. • Corrective actions are recommended to eliminate or decrease the risk of failure modes. - Abstract: Renewable energy plays a key role in the transition toward a low carbon economy and the provision of a secure supply of energy. Geothermal energy is a versatile source as a form of renewable energy that meets popular demand. Since some Geothermal Power Plants (GPPs) face various failures, the requirement of a technique for team engineering to eliminate or decrease potential failures is considerable. Because no specific published record of considering an FMEA applied to GPPs with common failure modes have been found already, in this paper, the utilization of Failure Modes and Effects Analysis (FMEA) as a convenient technique for determining, classifying and analyzing common failures in typical GPPs is considered. As a result, an appropriate risk scoring of occurrence, detection and severity of failure modes and computing the Risk Priority Number (RPN) for detecting high potential failures is achieved. In order to expedite accuracy and ability to analyze the process, XFMEA software is utilized. Moreover, 5 major parts of a GPP is studied to propose a suitable approach for developing GPPs and increasing reliability by recommending corrective actions for each failure mode

  17. Design of heat treatments for 9-12%Cr steels to optimise creep resistance for power plant applications

    Energy Technology Data Exchange (ETDEWEB)

    Morris, P.F.; Sachadel, U.A.; Clarke, P.D. [Tata Steel Europe, Rotherham (United Kingdom). Swinden Technology Centre; CRD and T, IJmuiden (Netherlands)

    2010-07-01

    Optimisation of the creep rupture properties of Steel 92 (9%Cr, 0.5%Mo, 2%W) by modification of heat treatment and C:N ratio has been studied. It was shown that a higher austenization temperature and double tempering at lower temperature can significantly extend creep life of the standard composition. The increase in austenization temperature from 1060 C and double tempering at 660 C/3h instead of single tempering at 780 C/2h resulted in the increase of stress rupture life from 1,734 to 6,179h at 650 C/110MPa. Even greater improvement in creep life was achieved by the combination of the modified heat treatment and decreased C:N ratio. In this case the creep life was extended to 10,255 h at 650 C/110MPa. A further increase in austenitization temperature to 1200 C for the decreased C:N ratio variant extended the rupture life to 17,118h. Initial results indicate that this modified heat treatment schedule does not result in notch brittle behaviour and most of the improvement in creep strength remains after a simulated post weld heat treatment at 740 C. The stress rupture programme is continuing and at 600 C test durations are approaching 60,000h. To explain the effect on rupture life thermodynamic calculations, microscopic investigations and a literature study were performed. Electron metallography investigations revealed that the lower tempering temperature resulted in a finer distribution of nano-size particles. Calculations show that increasing the austenitization temperature gave more dissolved B, N, C, Nb and V. The lower C:N ratio resulted in a higher atomic fraction of N in nano-size particles on subsequent tempering. Dissolved B should stabilize M{sub 23}M{sub 6} and dissolved N, C, Nv, and V should allow precipitation of a higher volume fraction of nano-size carbo-nitrides during tempering. Literature data suggest that lower tempering temperatures could also change their type from MX to M{sub 2}X. (orig.)

  18. Statistical analysis of nuclear power plant pump failure rate variability: some preliminary results

    International Nuclear Information System (INIS)

    Martz, H.F.; Whiteman, D.E.

    1984-02-01

    In-Plant Reliability Data System (IPRDS) pump failure data on over 60 selected pumps in four nuclear power plants are statistically analyzed using the Failure Rate Analysis Code (FRAC). A major purpose of the analysis is to determine which environmental, system, and operating factors adequately explain the variability in the failure data. Catastrophic, degraded, and incipient failure severity categories are considered for both demand-related and time-dependent failures. For catastrophic demand-related pump failures, the variability is explained by the following factors listed in their order of importance: system application, pump driver, operating mode, reactor type, pump type, and unidentified plant-specific influences. Quantitative failure rate adjustments are provided for the effects of these factors. In the case of catastrophic time-dependent pump failures, the failure rate variability is explained by three factors: reactor type, pump driver, and unidentified plant-specific influences. Finally, point and confidence interval failure rate estimates are provided for each selected pump by considering the influential factors. Both types of estimates represent an improvement over the estimates computed exclusively from the data on each pump

  19. Analysis of localized damage in creep rupture

    International Nuclear Information System (INIS)

    Wang Zhengdong; Wu Dongdi

    1992-01-01

    Continuum Damage Mechanics studies the effect of distributed defects, whereas the failure of engineering structures is usually caused by local damage. In this paper, an analysis of localized damage in creep rupture is carried out. The material tested is a 2 1/4Cr-1Mo pressure vessel steel and the material constants necessary for damage analysis are evaluated. Notched specimens are used to reflect localized damage in creep rupture and the amount of damage is measured using DCPD method. Through FEM computation, stress components and effective stress in the region of notch root are evaluated and it is found that the von Mises effective stress can represent the damage effective stress in the analysis of localized creep damage. It is possible to develop a method for the assessment of safety of pressure vessels under creep through localized creep damage analysis. (orig.)

  20. Laboratory creep and mechanical tests on salt data report (1975-1996): Waste Isolation Pilot Plant (WIPP) thermal/structural interactions program

    Energy Technology Data Exchange (ETDEWEB)

    Mellegard, K.D. [RE/SPEC Inc., Rapid City, SD (United States); Munson, D.E. [Sandia National Labs., Albuquerque, NM (United States)

    1997-02-01

    The Waste Isolation Pilot Plant (WIPP), a facility located in a bedded salt formation in Carlsbad, New Mexico, is being used by the U.S. Department of Energy to demonstrate the technology for safe handling and disposal of transuranic wastes produced by defense activities in the United States. In support of that demonstration, mechanical tests on salt were conducted in the laboratory to characterize material behavior at the stresses and temperatures expected for a nuclear waste repository. Many of those laboratory test programs have been carried out in the RE/SPEC Inc. rock mechanics laboratory in Rapid City, South Dakota; the first program being authorized in 1975 followed by additional testing programs that continue to the present. All of the WIPP laboratory data generated on salt at RE/SPEC Inc. over the last 20 years is presented in this data report. A variety of test procedures were used in performance of the work including quasi-static triaxial compression tests, constant stress (creep) tests, damage recovery tests, and multiaxial creep tests. The detailed data is presented in individual plots for each specimen tested. Typically, the controlled test conditions applied to each specimen are presented in a plot followed by additional plots of the measured specimen response. Extensive tables are included to summarize the tests that were performed. Both the tables and the plots contain cross-references to the technical reports where the data were originally reported. Also included are general descriptions of laboratory facilities, equipment, and procedures used to perform the work.

  1. Characterisation of creep cavitation damage in a stainless steel pressure vessel using small angle neutron scattering

    CERN Document Server

    Bouchard, P J; Treimer, W

    2002-01-01

    Grain-boundary cavitation is the dominant failure mode associated with initiation of reheat cracking, which has been widely observed in austenitic stainless steel pressure vessels operating at temperatures within the creep range (>450 C). Small angle neutron scattering (SANS) experiments at the LLB PAXE instrument (Saclay) and the V12 double-crystal diffractometer of the HMI-BENSC facility (Berlin) are used to characterise cavitation damage (in the size range R=10-2000 nm) in a variety of creep specimens extracted from ex-service plant. Factors that affect the evolution of cavities and the cavity-size distribution are discussed. The results demonstrate that SANS techniques have the potential to quantify the development of creep damage in type-316H stainless steel, and thereby link microstructural damage with ductility-exhaustion models of reheat cracking. (orig.)

  2. Common cause failure rate estimates for diesel generators in nuclear power plants

    International Nuclear Information System (INIS)

    Steverson, J.A.; Atwood, C.L.

    1982-01-01

    Common cause fault rates for diesel generators in nuclear power plants are estimated, using Licensee Event Reports for the years 1976 through 1978. The binomial failure rate method, used for obtaining the estimates, is briefly explained. Issues discussed include correct classification of common cause events, grouping of the events into homogeneous data subsets, and dealing with plant-to-plant variation

  3. Development of failure diagnosis method based on transient information of nuclear power plant

    International Nuclear Information System (INIS)

    Washio, Takashi; Kitamura, Masaharu; Sugiyama, Kazusuke

    1987-01-01

    This paper proposes a new method of failure diagnosis of nuclear power plant (NPP). Transient behavior of the NPP includes ample failure information even for a limited period of time from the failure onset. We tried to develop a diagnosis system with high capability of identifying the failure cause and of estimating failure severeness. The Walsh function transformation of transient time series data and the reduction of the Walsh coefficients into ternary valued amplitude indicators were utilized to extract the essential characteristics of failure. The correspondences of the transient characteristics and causes were summarized in a failure symptom database. A method of ternary tree search using an information measure as a heuristic strategy was adopted to conduct the efficient retrieval of failure causes in the database. Through numerical experiments using a simulation model of a NPP, the diagnostic capability of the system was proved to be satisfactory. (author)

  4. An analysis of human maintenance failures of a nuclear power plant

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-01-01

    In the report, a study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components appeared to be especially prone to human failures. Many human failures were found in safety related systems. Several failures also remained latent from outages to power operation. However, the safety significance of failures was generally small. Modifications were an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more in the future. (orig.)

  5. Constitutive Expression of a miR319 Gene Alters Plant Development and Enhances Salt and Drought Tolerance in Transgenic Creeping Bentgrass1[W][OA

    Science.gov (United States)

    Zhou, Man; Li, Dayong; Li, Zhigang; Hu, Qian; Yang, Chunhua; Zhu, Lihuang; Luo, Hong

    2013-01-01

    MicroRNA319 (miR319) is one of the first characterized and conserved microRNA families in plants and has been demonstrated to target TCP (for TEOSINTE BRANCHED/CYCLOIDEA/PROLIFERATING CELL FACTORS [PCF]) genes encoding plant-specific transcription factors. MiR319 expression is regulated by environmental stimuli, suggesting its involvement in plant stress response, although experimental evidence is lacking and the underlying mechanism remains elusive. This study investigates the role that miR319 plays in the plant response to abiotic stress using transgenic creeping bentgrass (Agrostis stolonifera) overexpressing a rice (Oryza sativa) miR319 gene, Osa-miR319a. We found that transgenic plants overexpressing Osa-miR319a displayed morphological changes and exhibited enhanced drought and salt tolerance associated with increased leaf wax content and water retention but reduced sodium uptake. Gene expression analysis indicated that at least four putative miR319 target genes, AsPCF5, AsPCF6, AsPCF8, and AsTCP14, and a homolog of the rice NAC domain gene AsNAC60 were down-regulated in transgenic plants. Our results demonstrate that miR319 controls plant responses to drought and salinity stress. The enhanced abiotic stress tolerance in transgenic plants is related to significant down-regulation of miR319 target genes, implying their potential for use in the development of novel molecular strategies to genetically engineer crop species for enhanced resistance to environmental stress. PMID:23292790

  6. Evaluation of long term creep-fatigue life for type 304 stainless steel

    International Nuclear Information System (INIS)

    Kawasaki, Hirotsugu; Ueno, Fumiyoshi; Aoto, Kazumi; Ichimiya, Masakazu; Wada, Yusaku

    1992-01-01

    The long term creep-fatigue life of type 304 stainless steel was evaluated by the creep-fatigue life prediction method based on a linear damage fraction rule. The displacement controlled creep-fatigue tests were carried out, and the time to failure of longer than 10000 hours was obtained. The creep damage of long term creep-fatigue was evaluated by taking into account the stress relaxation behavior with elastic follow-up during the hold period. The relationship between life reduction of creep-fatigue and fracture mode was provided by the creep cavity growth. The results of this study are summarized as follows; (1) The long term creep-fatigue data can be reasonably evaluated by the present method. The predicted lives were within a factor of 3 of the observed ones. (2) The present method provides the capability to predict the long term creep-fatigue life at lower temperatures as well as that at the creep dominant temperature. (3) The value of creep damage for the long term creep-fatigue data increased by elastic follow-up. The creep-fatigue damage diagram intercepted between 0.3 and 1 can represent the observed creep-fatigue damages. (4) The cavity growth depends on the hold time. The fracture of long term creep-fatigue is caused by the intergranular cavity growth. The intergranular fracture of creep-fatigue is initiated by the cavity growth and followed by the microcrack propagation along grain boundaries starting from creep cavities. (author)

  7. Monitoring microstructural evolution of alloy 617 with non-linear acoustics for remaining useful life prediction; multiaxial creep-fatigue and creep-ratcheting

    Energy Technology Data Exchange (ETDEWEB)

    Lissenden, Cliff [Pennsylvania State Univ., State College, PA (United States); Hassan, Tasnin [North Carolina State Univ., Raleigh, NC (United States); Rangari, Vijaya [Tuskegee Univ., Tuskegee, AL (United States)

    2014-10-30

    The research built upon a prior investigation to develop a unified constitutive model for design-­by-­analysis of the intermediate heat exchanger (IHX) for a very high temperature reactor (VHTR) design of next generation nuclear plants (NGNPs). Model development requires a set of failure data from complex mechanical experiments to characterize the material behavior. Therefore uniaxial and multiaxial creep-­fatigue and creep-­ratcheting tests were conducted on the nickel-­base Alloy 617 at 850 and 950°C. The time dependence of material behavior, and the interaction of time dependent behavior (e.g., creep) with ratcheting, which is an increase in the cyclic mean strain under load-­controlled cycling, are major concerns for NGNP design. This research project aimed at characterizing the microstructure evolution mechanisms activated in Alloy 617 by mechanical loading and dwell times at elevated temperature. The acoustic harmonic generation method was researched for microstructural characterization. It is a nonlinear acoustics method with excellent potential for nondestructive evaluation, and even online continuous monitoring once high temperature sensors become available. It is unique because it has the ability to quantitatively characterize microstructural features well before macroscale defects (e.g., cracks) form. The nonlinear acoustics beta parameter was shown to correlate with microstructural evolution using a systematic approach to handle the complexity of multiaxial creep-­fatigue and creep-­ratcheting deformation. Mechanical testing was conducted to provide a full spectrum of data for: thermal aging, tensile creep, uniaxial fatigue, uniaxial creep-­fatigue, uniaxial creep-ratcheting, multiaxial creep-fatigue, and multiaxial creep-­ratcheting. Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), and Optical Microscopy were conducted to correlate the beta parameter with individual microstructure mechanisms. We researched

  8. Monitoring microstructural evolution of alloy 617 with non-linear acoustics for remaining useful life prediction; multiaxial creep-fatigue and creep-ratcheting

    International Nuclear Information System (INIS)

    Lissenden, Cliff; Hassan, Tasnin; Rangari, Vijaya

    2014-01-01

    The research built upon a prior investigation to develop a unified constitutive model for design-@by-@analysis of the intermediate heat exchanger (IHX) for a very high temperature reactor (VHTR) design of next generation nuclear plants (NGNPs). Model development requires a set of failure data from complex mechanical experiments to characterize the material behavior. Therefore uniaxial and multiaxial creep-@fatigue and creep-@ratcheting tests were conducted on the nickel base Alloy 617 at 850 and 950°C. The time dependence of material behavior, and the interaction of time dependent behavior (e.g., creep) with ratcheting, which is an increase in the cyclic mean strain under load-@controlled cycling, are major concerns for NGNP design. This research project aimed at characterizing the microstructure evolution mechanisms activated in Alloy 617 by mechanical loading and dwell times at elevated temperature. The acoustic harmonic generation method was researched for microstructural characterization. It is a nonlinear acoustics method with excellent potential for nondestructive evaluation, and even online continuous monitoring once high temperature sensors become available. It is unique because it has the ability to quantitatively characterize microstructural features well before macroscale defects (e.g., cracks) form. The nonlinear acoustics beta parameter was shown to correlate with microstructural evolution using a systematic approach to handle the complexity of multiaxial creep-@fatigue and creep-@ratcheting deformation. Mechanical testing was conducted to provide a full spectrum of data for: thermal aging, tensile creep, uniaxial fatigue, uniaxial creep-@fatigue, uniaxial creep-ratcheting, multiaxial creep-fatigue, and multiaxial creep-@ratcheting. Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), and Optical Microscopy were conducted to correlate the beta parameter with individual microstructure mechanisms. We researched application of the

  9. Single-failure-proof cranes for nuclear power plants

    International Nuclear Information System (INIS)

    Porse, L.

    1979-05-01

    NRC has licensed reactors on the basis that the safe handling of critical loads can be accomplished by adding safety features to the handling equipment, by adding special features to the structures and areas over which the critical load is carried, or by a combination of the two. When reliance for the safe handling of critical loads is placed on the crane system itself, the system should be designed so that a single failure will not result in the loss of the capability of the system to safely retain the load. Features of the design, fabrication, installation, inspection, testing, and operation of single-failure-proof overhead crane handling systems that are used for handling critical loads are identified. These features are limited to the hoisting system and to braking systems for trolley and bridge. Other load-bearing items such as girders should be conservatively designed but need not be considered single failure proof

  10. Failure analysis and success analysis: roles in plant aging assessments

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1985-06-01

    Component aging investigations are an important element in NRC's Nuclear Plant Aging Research (NPAR) strategy. Potential sources of components include plants in decommissioning and commercial plant, both for in situ tests and for examination of equipment removed from service. Nuclear utilities currently have voluntary programs addressing aspects of equipment reliability, such as root cause analysis for safety-related equipment that malfunctions, and trending analysis to follow the course of both successful and abnormal equipment performance. Properly coordinated, the NPAR and utility programs offer an important approach to establish the data base necessary for life extension of nuclear electrical generating plants

  11. Phenomenological approach to precise creep life prediction by means of quantitative evaluation of strain rate acceleration in secondary creep

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Miyano, Takaya

    2010-01-01

    A method of creep life prediction by means of Strain-Acceleration-Parameter (SAP), α, is presented. The authors show that the shape of creep curve can be characterized by SAP that reflects magnitude of strain-rate change in secondary creep. The SAP-values, α are evaluated on magnesium-aluminium solution hardened alloys. Reconstruction of creep curves by combinations of SAP and minimum-creep rates are successfully performed, and the curves reasonably agree with experiments. The advantage of the proposed method is that the required parameters evaluated from individual creep curves are directly connected with the minimum creep rate. The predicted times-to-failure agree well with that obtained by experiments, and possibility of precise life time prediction by SAP is pronounced.

  12. Failure analysis of boiler tubes in lakhra coal power plant

    International Nuclear Information System (INIS)

    Shah, A.; Baluch, M.M.; Ali, A.

    2010-01-01

    Present work deals with the failure analysis of a boiler tube in Lakhra fluidized bed combustion power station. Initially, visual inspection technique was adopted to analyse the fractured surface. Detailed microstructural investigations of the busted boiler tube were carried out using light optical microscope and scanning electron microscope. The hardness tests were also performed. A 50 percent decrease in hardness of intact portion of the tube material and from area adjacent to failure was measured, which was found to be in good agreement with the wall thicknesses measured of the busted boiler tube i.e. 4 mm and 2 mm from unaffected portion and ruptured area respectively. It was concluded that the major cause of failure of boiler tube is erosion of material which occurs due the coal particles strike at the surface of the tube material. Since the temperature of boiler is not maintained uniformly. The variations in boiler temperature can also affect the material and could be another reason for the failure of the tube. (author)

  13. Statistical analysis of human maintenance failures of a nuclear power plant

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-01-01

    In this paper, a statistical study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant safety and system availability. More than 4400 maintenance history reports from the years 1992-1994 of Olkiluoto BWR nuclear power plant (NPP) were analysed together with the maintenance personnel. The human action induced faults were classified, e.g., according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components are especially prone to human failures. Many human failures were found in safety related systems. Similarly, several failures remained latent from outages to power operation. The safety significance was generally small. Modifications are an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more, in future. (orig.)

  14. Simultaneous consolidation and creep

    DEFF Research Database (Denmark)

    Krogsbøll, Anette

    1997-01-01

    Materials that exhibit creep under constant effective stress typically also show rate dependent behavior. The creep deformations and the rate sensitive behavior is very important when engineering and geological problems with large time scales are considered. When stress induced compaction...

  15. Creep of crystals

    International Nuclear Information System (INIS)

    Poirier, J.-P.

    1988-01-01

    Creep mechanisms for metals, ceramics and rocks, effect of pressure and temperature on deformation processes are considered. The role of crystal defects is analysed, different models of creep are described. Deformation mechanisms maps for different materials are presented

  16. Seismically induced common cause failures in PSA of nuclear power plants

    International Nuclear Information System (INIS)

    Ravindra, M.K.; Johnson, J.J.

    1991-01-01

    In this paper, a research project on the seismically induced common cause failures in nuclear power plants performed for Toshiba Corp. is described. The objective of this research was to develop the procedure for estimating the common cause failure probabilities of different nuclear power plant components using the combination of seismic experience data, the review of sources of dependency, sensitivity studies and engineering judgement. The research project consisted of three tasks: the investigation of damage instances in past earthquakes, the analysis of multiple failures and their root causes, and the development of the methodology for assessing seismically induced common cause failures. The details of these tasks are explained. In this paper, the works carried out in the third task are described. A methodology for treating common cause failures and the correlation between component failures is formulated; it highlights the modeling of event trees taking into account common cause failures and the development of fault trees considering the correlation between component failures. The overview of seismic PSA, the quantification methods for dependent failures and Latin Hypercube sampling method are described. (K.I.)

  17. Modelling of creep damage development in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    The physical creep damage, which is observed in fossil-fired power plants, is mainly due to the formation of cavities and their interaction. It has previously been demonstrated that both the nucleation and growth of creep cavities can be described by power functions in strain for low alloy and 12 % CrMoV creep resistant steels. It possible to show that the physical creep damage is proportional to the product of the number of cavities and their area. Hence, the physical creep damage can also be expressed in terms of the creep strain. In the presentation this physical creep damage is connected to the empirical creep damage classes (1-5). A creep strain-time function, which is known to be applicable to low alloy and 12 % CrMoV creep resistant steels, is used to describe tertiary creep. With this creep strain - time model the residual lifetime can be predicted from the observed damage. For a given damage class the remaining life is directly proportional to the service time. An expression for the time to the next inspection is proposed. This expression is a function of fraction of the total allowed damage, which is consumed till the next inspection. (orig.) 10 refs.

  18. Modelling of creep damage development in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, R [Swedish Institute for Metals Research, Stockholm (Sweden)

    1999-12-31

    The physical creep damage, which is observed in fossil-fired power plants, is mainly due to the formation of cavities and their interaction. It has previously been demonstrated that both the nucleation and growth of creep cavities can be described by power functions in strain for low alloy and 12 % CrMoV creep resistant steels. It possible to show that the physical creep damage is proportional to the product of the number of cavities and their area. Hence, the physical creep damage can also be expressed in terms of the creep strain. In the presentation this physical creep damage is connected to the empirical creep damage classes (1-5). A creep strain-time function, which is known to be applicable to low alloy and 12 % CrMoV creep resistant steels, is used to describe tertiary creep. With this creep strain - time model the residual lifetime can be predicted from the observed damage. For a given damage class the remaining life is directly proportional to the service time. An expression for the time to the next inspection is proposed. This expression is a function of fraction of the total allowed damage, which is consumed till the next inspection. (orig.) 10 refs.

  19. Expert environment for the development of nuclear power plants failure diagnosis systems

    International Nuclear Information System (INIS)

    Guido, P.N.; Oggianu, S.; Etchepareborda, A.; Fernandez, O.

    1996-01-01

    The present work explores some of the developing stages of an Expert Environment for plant failures Diagnosis Systems starting from Knowledge Based Systems. We present a prototype that carries out an inspection of anomalous symptoms and a diagnosis process based on a Plant Abnormality Model of a PHWR secondary system

  20. The effect of uncertainties in nuclear reactor plant-specific failure data on core damage frequency

    International Nuclear Information System (INIS)

    Martz, H.F.

    1995-05-01

    It is sometimes the case in PRA applications that reported plant-specific failure data are, in fact, only estimates which are uncertain. Even for detailed plant-specific data, the reported exposure time or number of demands is often only an estimate of the actual exposure time or number of demands. Likewise the reported number of failure events or incidents is sometimes also uncertain because incident or malfunction reports may be ambiguous. In this report we determine the corresponding uncertainty in core damage frequency which can b attributed to such uncertainties in plant-specific data using a simple but typical nuclear power reactor example

  1. Analysis of failure and maintenance experiences of motor operated valves in a Finnish nuclear power plant

    International Nuclear Information System (INIS)

    Simola, K.; Laakso, K.

    1992-01-01

    Operating experiences from 1981 up to 1989 of totally 104 motor operated closing valves (MOV) in different safety systems at TVO I and II nuclear power units were analysed in a systematic way. The qualitative methods used were failure mode and effects analysis (FMEA) and maintenance effects and criticality analysis (MECA). The failure descriptions were obtained from power plant's computerized failure reporting system. The reported 181 failure events were reanalysed and sorted according to specific classifications developed for the MOV function. Filled FMEA and MECA sheets on individual valves were stored in a microcomputer data base for further analyses. Analyses were performed for the failed mechanical and electrical valve parts, ways of detection of failure modes, failure effects, and repair and unavailability times

  2. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1995-01-01

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  3. Creep-fatigue of low cobalt superalloys

    Science.gov (United States)

    Halford, G. R.

    1982-01-01

    Testing for the low cycle fatigue and creep fatigue resistance of superalloys containing reduced amounts of cobalt is described. The test matrix employed involves a single high temperature appropriate for each alloy. A single total strain range, again appropriate to each alloy, is used in conducting strain controlled, low cycle, creep fatigue tests. The total strain range is based upon the level of straining that results in about 10,000 cycles to failure in a high frequency (0.5 Hz) continuous strain-cycling fatigue test. No creep is expected to occur in such a test. To bracket the influence of creep on the cyclic strain resistance, strain hold time tests with ore minute hold periods are introduced. One test per composition is conducted with the hold period in tension only, one in compression only, and one in both tension and compression. The test temperatures, alloys, and their cobalt compositions that are under study are given.

  4. Performance Based Failure Criteria of the Base Isolation System for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Jung Han; Kim, Min Kyu; Choi, In Kil

    2013-01-01

    The realistic approach to evaluate the failure state of the base isolation system is necessary. From this point of view, several concerns are reviewed and discussed in this study. This is the preliminary study for the performance based risk assessment of a base isolated nuclear power plant. The items to evaluate the capacity and response of an individual base isolator and a base isolation system were briefly outlined. However, the methodology to evaluate the realistic fragility of a base isolation system still needs to be specified. For the quantification of the seismic risk for a nuclear power plant structure, the failure probabilities of the structural component for the various seismic intensity levels need to be calculated. The failure probability is evaluated as the probability when the seismic response of a structure exceeds the failure criteria. Accordingly, the failure mode of the structural system caused by an earthquake vibration should be defined first. The type of a base isolator appropriate for a nuclear power plant structure is regarded as an elastometric rubber bearing with a lead core. The failure limit of the lead-rubber bearing (LRB) is not easy to be predicted because of its high nonlinearity and a complex loading condition by an earthquake excitation. Furthermore, the failure mode of the LRB system installed below the nuclear island cannot be simply determined because the basemat can be sufficiently supported if the number of damaged isolator is not much

  5. Creep theories compared by means of high sensitivity tensile creep data

    International Nuclear Information System (INIS)

    Salim, A.

    1987-01-01

    Commonly used creep theories include time-hardening, strain-hardening and Rabotnov's modified strain-hardening. In the paper they are examined by using high sensitivity tensile creep data produced on 1% CrMoV steel at a temperatue of 565 0 C. A special creep machine designed and developed by the author is briefly described and is compared with other existing machines. Tensile creep data reported cover a stress range of 100-260 MN m -2 ; four variable-creep tests each in duplicate are also reported. Test durations are limited to 3000 h, or failure, whichever occurs earlier. The strain-hardening theory and Rabotnov's modified strain-hardening theory are found to give good prediction of creep strain under variable stress conditions. The time-hardening theory shows a relatively poor agreement and considerably underestimates the accumulated inelastic strain under increasing stress condition. This discrepancy increases with the increased stress rate. The theories failed to predict the variable stress results towards the later part of the test where tertiary effects were significant. The use of creep equations which could account for creep strain at higher stress levels seems to improve the situation considerably. Under conditions of variable stress, it is suggested that a theory based on continuous damage mechanics concepts might give a better prediction. (author)

  6. A creep life assessment method for boiler pipes using small punch creep test

    International Nuclear Information System (INIS)

    Izaki, Toru; Kobayashi, Toshimi; Kusumoto, Junichi; Kanaya, Akihiro

    2009-01-01

    The small punch creep (SPC) test is considered as a highly useful method for creep life assessment for high temperature plant components. SPC uses miniature-sized specimens and does not cause any serious sampling damages, and its assessment accuracy is at a high level. However, in applying the SPC test to the residual creep life assessment of the boiler in service, there are some issues to be studied. In order to apply SPC test to the residual creep life assessment of the 2.25Cr-1Mo steel boiler pipe, the relationship between uniaxial creep stress and the SPC test load has been studied. The virgin material, pre-crept, weldment and service aged samples of 2.25Cr-1Mo steel were tested. It was confirmed that the relationship between uniaxial creep stress and the SPC test load at the same rupture time can be described as a single straight line independent of test conditions and materials. Therefore a life assessment is possible by using SPC test in place of uniaxial creep tests. The creep life assessment using SPC was applied to actual thermal power plant components which are in service.

  7. Creep in ceramics

    CERN Document Server

    Pelleg, Joshua

    2017-01-01

    This textbook is one of its kind, since there are no other books on Creep in Ceramics. The book consist of two parts: A and B. In part A general knowledge of creep in ceramics is considered, while part B specifies creep in technologically important ceramics. Part B covers creep in oxide ceramics, carnides and nitrides. While covering all relevant information regarding raw materials and characterization of creep in ceramics, the book also summarizes most recent innovations and developments in this field as a result of extensive literature search.

  8. Expression of a novel antimicrobial peptide Penaeidin4-1 in creeping bentgrass (Agrostis stolonifera L. enhances plant fungal disease resistance.

    Directory of Open Access Journals (Sweden)

    Man Zhou

    Full Text Available BACKGROUND: Turfgrass species are agriculturally and economically important perennial crops. Turfgrass species are highly susceptible to a wide range of fungal pathogens. Dollar spot and brown patch, two important diseases caused by fungal pathogens Sclerotinia homoecarpa and Rhizoctonia solani, respectively, are among the most severe turfgrass diseases. Currently, turf fungal disease control mainly relies on fungicide treatments, which raises many concerns for human health and the environment. Antimicrobial peptides found in various organisms play an important role in innate immune response. METHODOLOGY/PRINCIPAL FINDINGS: The antimicrobial peptide - Penaeidin4-1 (Pen4-1 from the shrimp, Litopenaeus setiferus has been reported to possess in vitro antifungal and antibacterial activities against various economically important fungal and bacterial pathogens. In this study, we have studied the feasibility of using this novel peptide for engineering enhanced disease resistance into creeping bentgrass plants (Agrostis stolonifera L., cv. Penn A-4. Two DNA constructs were prepared containing either the coding sequence of a single peptide, Pen4-1 or the DNA sequence coding for the transit signal peptide of the secreted tobacco AP24 protein translationally fused to the Pen4-1 coding sequence. A maize ubiquitin promoter was used in both constructs to drive gene expression. Transgenic turfgrass plants containing different DNA constructs were generated by Agrobacterium-mediated transformation and analyzed for transgene insertion and expression. In replicated in vitro and in vivo experiments under controlled environments, transgenic plants exhibited significantly enhanced resistance to dollar spot and brown patch, the two major fungal diseases in turfgrass. The targeting of Pen4-1 to endoplasmic reticulum by the transit peptide of AP24 protein did not significantly impact disease resistance in transgenic plants. CONCLUSION/SIGNIFICANCE: Our results

  9. Detection of instrument or component failures in a nuclear plant by Luenberger observers

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Colley, R.W.; Alexandro, F.J.; Clark, R.N.

    1985-01-01

    A diagnostic system, which will distinguish between instrument failures (flowmeters, etc.) and component failures (valves, filters, etc.) that show the same symptoms, has been developed for nuclear Plants using Luenberger observers. Luenberger observers are online computer based modules constructed following the technology of Clark [3]. A seventh order model of an FFTF subsystem was constructed using the Advanced Continuous Simulation Language (ACSL) and was used to show through simulation that Luenberger observers can be applied to nuclear systems

  10. Process pump operating problems and equipment failures, F-Canyon Reprocessing Facility, Savannah River Plant

    International Nuclear Information System (INIS)

    Durant, W.S.; Starks, J.B.; Galloway, W.D.

    1987-02-01

    A compilation of operating problems and equipment failures associated with the process pumps in the Savannah River Plant F-Canyon Fuel Reprocessing Facility is presented. These data have been collected over the 30-year operation of the facility. An analysis of the failure rates of the pumps is also presented. A brief description of the pumps and the data bank from which the information was sorted is also included

  11. Analysis of GMO Plum Plant Culture in System Operations Failure

    Science.gov (United States)

    Mercado, Dianne

    2017-01-01

    GMO plum trees are being evaluated at the Kennedy Space Center as a possible candidate for future space crops. Previously conducted horticultural testing compared the performance of several plum genotypes in controlled environment chambers, resulting in a down-selection to the NASA-11 genotype. Precursory studies determined the water use requirements to sustain the plants as well as the feasibility of grafting non-GMO plum scions onto GMO plum rootstocks of NASA-5, NASA-10, and NASA-11 genotypes. This study follows the growth and horticultural progress of plum trees and in-vitro cultures from August 2017 to November 2017, and provides supplemental support for future GMO plum studies. The presence of Hurricane Irma in early September 2017 resulted in the plants undergoing material deterioration from major changes to their overall horticultural progress.

  12. A quantitative impact analysis of sensor failures on human operator's decision making in nuclear power plants

    International Nuclear Information System (INIS)

    Seong, Poong Hyun

    2004-01-01

    In emergency or accident situations in nuclear power plants, human operators take important roles in generating appropriate control signals to mitigate accident situation. In human reliability analysis (HRA) in the framework of probabilistic safety assessment (PSA), the failure probabilities of such appropriate actions are estimated and used for the safety analysis of nuclear power plants. Even though understanding the status of the plant is basically the process of information seeking and processing by human operators, it seems that conventional HRA methods such as THERP, HCR, and ASEP does not pay a lot of attention to the possibilities of providing wrong information to human operators. In this paper, a quantitative impact analysis of providing wrong information to human operators due to instrument faults or sensor failures is performed. The quantitative impact analysis is performed based on a quantitative situation assessment model. By comparing the situation in which there are sensor failures and the situation in which there are not sensor failures, the impact of sensor failures can be evaluated quantitatively. It is concluded that the impact of sensor failures are quite significant at the initial stages, but the impact is gradually reduced as human operators make more and more observations. Even though the impact analysis is highly dependent on the situation assessment model, it is expected that the conclusions made based on other situation assessment models with be consistent with the conclusion made in this paper. (author)

  13. Experimental study and modelling of high temperature creep flow and damage behaviour of 9Cr1Mo-NbV steel weldments

    International Nuclear Information System (INIS)

    Gaffard, V.

    2004-12-01

    Chromium martensitic stainless steels are under development since the 70's with the prospect of using them as structural components in thermal and nuclear power plants. The modified 9Cr1Mo-NbV steel is already used, especially in England and Japan, as a material for structural components in thermal power plants where welding is a commonly used joining technique. New generations of chromium martensitic stainless steels with improved mechanical properties for high pressure and temperature use are currently under development. However, observations of several in-service premature failures of welded components in 9Cr1Mo-NbV steel, outline a strong need for understanding the high temperature creep flow and damage behaviour of 9Cr1Mo-NbV steels and weldments. The present study aimed at experimentally determining and then modelling the high temperature creep flow and damage behaviour of both 9Cr1Mo-NbV steels and weldments (typically in the temperature range from 450 C to 650 C). The base metal was first studied as the reference material. It was especially evidenced that tempered chromium martensitic steels exhibit a change in both creep flow and damage behaviour for long term creep exposure. As a consequence, the classically performed extrapolation of 1,000 hours creep data to 100,000 hours creep lifetime predictions might be very hazardous. Based on experimental observations, a new model, integrating and coupling multiple creep flow and damage mechanisms, was developed in the framework of the mechanics of porous media. It was then successfully used to represent creep flow and damage behaviour of the base metal from high to low stress levels even for complex multiaxial loading conditions. Although the high temperature creep properties of the base metal are quite good, the occurrence of premature failure in weldments in high temperature creep conditions largely focused the attention of the scientific community. The lower creep strength of the weld component was also

  14. Advances in the assessment of creep data

    Energy Technology Data Exchange (ETDEWEB)

    Holdsworth, S.R.

    2010-07-01

    Many of the classical models representing the creep and rupture behaviour of metals were developed prior to and during the 1950s and 1960s, and their subsequent exploitation, in particular for the assessment of large creep property datasets, was initially limited by the capability of the analytical tools available at the time. The formation of ECCC (the European Creep Collaborative Committee) in 1991, with a main objective of providing reliable peer reviewed long-time creep property values for European Design and Product Standards, led to the development of rigorous assessment procedures such as PD6605 and DESA incorporating post assessment tests to verify: physical realism, effectiveness of model-fit within the range of the source experimental data, and extrapolation credibility. The first ECCC assessment recommendations published in 1996 undoubtedly provided a catalyst for others to exploit the availability of low cost, powerful desktop computers to develop rigorous methodologies for the physically realistic analysis of uniaxial and multi-axial data for the reliable and accurate characterisation of creep strain, and rupture strength and ductility properties. More recent improvements in data assessment methodologies have been driven by the need to effectively model the creep deformation and rupture characteristics of the complex new generation alloys and fabrications being designed to cater for the continually evolving requirements of modern advanced power plant. These advances in the assessment of creep data are reviewed. (orig.)

  15. Preliminary Analysis of the Common Cause Failure Events for Domestic Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kang, Daeil; Han, Sanghoon

    2007-01-01

    It is known that the common cause failure (CCF) events have a great effect on the safety and probabilistic safety assessment (PSA) results of nuclear power plants (NPPs). However, the domestic studies have been mainly focused on the analysis method and modeling of CCF events. Thus, the analysis of the CCF events for domestic NPPs were performed to establish a domestic database for the CCF events and to deliver them to the operation office of the international common cause failure data exchange (ICDE) project. This paper presents the analysis results of the CCF events for domestic nuclear power plants

  16. Failure of Titanium Condenser Tubes after 24 Years Power Plant Service

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Enemark, Allan; Hangaard, Anders

    2014-01-01

    The titanium condenser has been in operation for 24 years at Amager unit 3 power plant. In February 2012, the plant was contaminated by seawater due to a failed condenser tube and some tubes were plugged. A month later, the plant tripped again. Small leaks were found again and finally approx. 200...... a plant trip. In addition, small amounts of titanium hydride were revealed to be present in the tubes within the tubesheet indicating that the carbon steel tubesheet was corroding due to ingress of salt water. Although this was not the reason for the failure, it indicated the need for repair of the epoxy...

  17. Creep and creep fatigue crack behavior of 1Cr- and 9Cr-steels

    International Nuclear Information System (INIS)

    Maile, K.; Klenk, A.; Schellenberg, G.; Granacher, J.; Tramer, M.

    2000-01-01

    A large database for creep crack initiation and propagation under constant load conditions is available on conventional power plant steels of types 1%Cr and 12%Cr. Modern plants are often used in the medium and peak load regime, thus the dominant loading situation in high temperature components is creep fatigue. For life assessment data about crack initiation and growth under creep fatigue loading are required. These characteristics can not be substituted by pure fatigue or creep crack data. Therefore, a comprehensive test programme was started to investigate the creep fatigue crack behaviour of a 1%CrMoNiV turbine rotor steel (30CrMoNiV 4 11) at 550 C and a new 9%CrMoVNb pipe steel (type P 9 1) at 600 C. DENT-specimen with 15 and 60 mm thickness as well as side grooved CT-specimen with 25 and 50 mm thickness have been tested to determine possible influences of geometry and thus to check the transferability of the data to components. The creep fatigue crack growth results of tests with dwell times between t H = 0,32h and 10 h lie in the scatterbands given by creep crack growth results. Nevertheless a higher crack growth rate under creep fatigue conditions can be stated. An increase in crack growth rate due to creep fatigue is clearly visible. Loading situations with frequencies higher than 1.10 -4 Hz should be not assessed with pure creep crack results or sufficient safety margins have to be applied. (orig.)

  18. Tube failures due to cooling process problem and foreign materials in power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, J. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia); Purbolaksono, J., E-mail: judha@uniten.edu.m [Department of Mechanical Engineering, Universiti Tenaga Nasional, Km 7 Jalan Kajang-Puchong, Kajang 43009, Selangor (Malaysia); Beng, L.C. [Kapar Energy Ventures Sdn Bhd, Jalan Tok Muda, Kapar 42200 (Malaysia)

    2010-07-15

    Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

  19. Tube failures due to cooling process problem and foreign materials in power plants

    International Nuclear Information System (INIS)

    Ahmad, J.; Purbolaksono, J.; Beng, L.C.

    2010-01-01

    Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented.

  20. Creep of Posidonia Shale at Elevated Pressure and Temperature

    Science.gov (United States)

    Rybacki, E.; Herrmann, J.; Wirth, R.; Dresen, G.

    2017-12-01

    The economic production of gas and oil from shales requires repeated hydraulic fracturing operations to stimulate these tight reservoir rocks. Besides simple depletion, the often observed decay of production rate with time may arise from creep-induced fracture closure. We examined experimentally the creep behavior of an immature carbonate-rich Posidonia shale, subjected to constant stress conditions at temperatures between 50 and 200 °C and confining pressures of 50-200 MPa, simulating elevated in situ depth conditions. Samples showed transient creep in the semibrittle regime with high deformation rates at high differential stress, high temperature and low confinement. Strain was mainly accommodated by deformation of the weak organic matter and phyllosilicates and by pore space reduction. The primary decelerating creep phase observed at relatively low stress can be described by an empirical power law relation between strain and time, where the fitted parameters vary with temperature, pressure and stress. Our results suggest that healing of hydraulic fractures at low stresses by creep-induced proppant embedment is unlikely within a creep period of several years. At higher differential stress, as may be expected in situ at contact areas due to stress concentrations, the shale showed secondary creep, followed by tertiary creep until failure. In this regime, microcrack propagation and coalescence may be assisted by stress corrosion. Secondary creep rates were also described by a power law, predicting faster fracture closure rates than for primary creep, likely contributing to production rate decline. Comparison of our data with published primary creep data on other shales suggests that the long-term creep behavior of shales can be correlated with their brittleness estimated from composition. Low creep strain is supported by a high fraction of strong minerals that can build up a load-bearing framework.

  1. Estimation of component failure rates for PSA on nuclear power plants 1982-1997

    International Nuclear Information System (INIS)

    Kirimoto, Yukihiro; Matsuzaki, Akihiro; Sasaki, Atsushi

    2001-01-01

    Probabilistic safety assessment (PSA) on nuclear power plants has been studied for many years by the Japanese industry. The PSA methodology has been improved so that PSAs for all commercial LWRs were performed and used to examine for accident management.On the other hand, most data of component failure rates in these PSAs were acquired from U.S. databases. Nuclear Information Center (NIC) of Central Research Institute of Electric Power Industry (CRIEPI) serves utilities by providing safety- , and reliability-related information on operation and maintenance of the nuclear power plants, and by evaluating the plant performance and incident trends. So, NIC started a research study on estimating the major component failure rates at the request of the utilities in 1988. As a result, we estimated the hourly-failure rates of 47 component types and the demand-failure rates of 15 component types. The set of domestic component reliability data from 1982 to 1991 for 34 LWRs has been evaluated by a group of PSA experts in Japan at the Nuclear Safety Research Association (NSRA) in 1995 and 1996, and the evaluation report was issued in March 1997. This document describes the revised component failure rate calculated by our re-estimation on 49 Japanese LWRs from 1982 to 1997. (author)

  2. Analysis of failure events for expansion joints in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Masahiro [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    Although a large number of expansion joints are used in nuclear power plants with light water reactors, their failure events have not been paid as much attention as those of vessels and pipes. However, as the operation period of nuclear power plants becomes longer, it is necessary to pay attention to their failure events as well as those of vessels and pipes, because aging problems and latent troubles originated in design or fabrication of expansion joints may appear during their long period operation. In this work, we investigated failure event reports of expansion joints in nuclear power plants both in Japan and in U.S.A. and analyzed (1) the influence to output power level, (2) the position and (3) the cause of each failure. It is revealed that the failure events of expansion joints have continuously occurred, some of which have exerted influence upon power level and have caused fatal or injury accidents of personnel, and hence the importance of corrective actions to prevent the recurrence of such events is pointed out. The importance of countermeasures to the following individual events is also pointed out: (1) corrosion of expansion joints in service water systems, (2) degradation of rubber expansion joints in main condensers, (3) vibration and fatigue of expansion joints in extraction steam lines and (4) transgranular stress corrosion cracking of penetration bellows of containments. (author)

  3. Containment failure modes preliminary analysis for Atucha-I nuclear power plant during severe accidents

    International Nuclear Information System (INIS)

    Baron, J.; Caballero, C.; Zarate, S.M.

    1997-01-01

    The present work has the objective to analyze the containment behavior of the Atucha-I nuclear power plant during a severe accident, as part of a probabilistic safety assessment (PSA). Initially, a generic description of the containment failure modes considered in other PSAs is performed. Then, the possible containment failure modes for Atucha I are qualitatively analyzed, according to it design peculiarities. These failure modes involve some substantial differences from other PSAs, due to the particular design of Atucha I. Among others, it is studied the influence of: moderator/coolant separation, existence of cooling Zircaloy channels, existence of filling bodies inside the pressure vessel, reactor cavity geometry, on-line refueling mode, and existence of a double shell containment (steel and concrete) with an annular separation room. As a functions of the before mentioning analysis, a series of parameters to be taken into account is defined, on a preliminary basis, for definition of the plant damage states. (author) [es

  4. Recent Operating Experience involving Power Electronics Failure in Korea Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Jaedo

    2015-01-01

    Recently, modern power electronics devices for electrical component were steadily increased in electrical systems which used for main power control and protection. To upgrade the system reliability we recommended the redundancy for electrical equipment trip system. The past several years, Korean Nuclear power plants have changed the electrical control and protection systems (Auto Voltage Regulator, Power Protection Relay) for main generator and main power protection relay systems. In this paper we deal with operating experience involving modern solid state power electronics failure in Korean nuclear power plants. One of the failures we will discuss the degraded phenomenon of power electronics device for CEDMCS(Control Element Drive Mechanism Control System). As the result of the failure we concerned about the modification for trip source of main generator excitation systems and others. We present an interesting issue for modern solid state devices (IGBT, Thyristors). (authors)

  5. Trend evaluation of incident and failure data from japanese nuclear power plants

    International Nuclear Information System (INIS)

    Kondo, S.; Hada, M.; Mikami, Y.

    1990-01-01

    Major incident and failure at nuclear power plants in Japan have to be reported to the regulatory agency i.e. Ministry of International Trade and Industry (MITI). Nuclear Power Safety Information Research Center (NUSIRC) has established a system for the collection, classification and analysis of this report under the contract to MITI. In this paper, the authors give several results of trend analyses of the incidents related to electric and instrumentation and control (I and C) systems reported, especially, the trend of the contribution of troubles in I and C system to the operation states, analysis of dominant contributors to the failure of I and C systems. Also, the relations of failure frequency of these systems with the plant age and effect of periodic inspections of it are discussed in some detail

  6. Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

    International Nuclear Information System (INIS)

    Azarm, M.A.; Boccio, J.L.; Mitra, S.

    1985-12-01

    This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs

  7. Correlation of Creep Behavior of Domal Salts

    International Nuclear Information System (INIS)

    Munson, D.E.

    1999-01-01

    The experimentally determined creep responses of a number of domal salts have been reported in, the literature. Some of these creep results were obtained using standard (conventional) creep tests. However, more typically, the creep data have come from multistage creep tests, where the number of specimens available for testing was small. An incremental test uses abrupt changes in stress and temperature to produce several time increments (stages) of different creep conditions. Clearly, the ability to analyze these limited data and to correlate them with each other could be of considerable potential value in establishing the mechanical characteristics of salt domes, both generally and specifically. In any analysis, it is necessary to have a framework of rules to provide consistency. The basis for the framework is the Multimechanism-Deformation (M-D) constitutive model. This model utilizes considerable general knowledge of material creep deformation to supplement specific knowledge of the material response of salt. Because the creep of salt is controlled by just a few micromechanical mechanisms, regardless of the origin of the salt, certain of the material parameters are values that can be considered universal to salt. Actual data analysis utilizes the methodology developed for the Waste Isolation Pilot Plant (WIPP) program, and the response of a bedded pure WIPP salt as the baseline for comparison of the domal salts. Creep data from Weeks Island, Bryan Mound, West Hackberry, Bayou Choctaw, and Big Hill salt domes, which are all sites of Strategic Petroleum Reserve (SPR) storage caverns, were analyzed, as were data from the Avery Island, Moss Bluff, and Jennings salt domes. The analysis permits the parameter value sets for the domal salts to be determined in terms of the M-D model with various degrees of completeness. In turn this permits detailed numerical calculations simulating cavern response. Where the set is incomplete because of the sparse database, reasonable

  8. Operational planning optimization of steam power plants considering equipment failure in petrochemical complex

    International Nuclear Information System (INIS)

    Luo, Xianglong; Zhang, Bingjian; Chen, Ying; Mo, Songping

    2013-01-01

    Highlights: ► We develop a systematic programming methodology to address equipment failure. ► We classify different operation conditions into real periods and virtual periods. ► The formulated MILP models guarantee cost reduction and enough operation safety. ► The consideration of reserving operation redundancy is effective. - Abstract: One or more interconnected steam power plants (SPPs) are constructed in a petrochemical complex to supply utility energy to the process. To avoid large economic penalties or process shutdowns, these SPPs should be flexible and reliable enough to meet the process energy requirement under varying conditions. Unexpected utility equipment failure is inevitable and difficult to be predicted. Most of the conventional methods are based on the assumption that SPPs do not experience any kind of equipment failure. Unfortunately, a process shutdown cannot be avoided when equipment fails unexpectedly. In this paper, a systematic methodology is presented to minimize the total cost under normal conditions while reserving enough flexibility and safety for unexpected equipment failure conditions. The proposed method transforms the different conditions into real periods to indicate normal scenarios and virtual periods to indicate unexpected equipment failure scenarios. The optimization strategy incorporating various operation redundancy scheduling, the transition constraints from equipment failure conditions to normal conditions, and the boiler load increase behavior modeling are presented to save cost and guarantee operation safety. A detailed industrial case study shows that the proposed systematic methodology is effective and practical in coping with equipment failure conditions with only few additional cost penalties

  9. Defense-in-depth for common cause failure of nuclear power plant safety system software

    International Nuclear Information System (INIS)

    Tian Lu

    2012-01-01

    This paper briefly describes the development of digital I and C system in nuclear power plant, and analyses the viewpoints of NRC and other nuclear safety authorities on Software Common Cause Failure (SWCCF). In view of the SWCCF issue introduced by the digitized platform adopted in nuclear power plant safety system, this paper illustrated a diversified defence strategy for computer software and hardware. A diversified defence-in-depth solution is provided for digital safety system of nuclear power plant. Meanwhile, analysis on problems may be faced during application of nuclear safety license are analyzed, and direction of future nuclear safety I and C system development are put forward. (author)

  10. Creep in commercially pure metals

    International Nuclear Information System (INIS)

    Nabarro, F.R.N.

    2006-01-01

    The creep of commercially pure polycrystalline metals under constant stress has four stages: a virtually instantaneous extension, decelerating Andrade β creep, almost steady-state Andrade κ creep, and an acceleration towards failure. Little is known about the first stage, and the fourth stage has been extensively reviewed elsewhere. The limited experimental evidence on the physical mechanism of the second stage is reviewed and a critical discussion is given of various theories of this stage. The dependence of strain rate on stress in the third, steady-state, period seems to fall into two regimes, a power law with an exponent of about 4-5, and a rather closely exponential law. The limits of the parameters within which a simple theory of the exponential dependence can be expected to be valid are discussed, and found to be compatible with experiments. Theories of the power-law dependence are discussed, and, appear to be unconvincing. The theoretical models do not relate closely to the metallographic and other physical observations. In view of the weakness of theory, experiments which may indicate the physical processes dominant in steady-state creep are reviewed. It is usually not clear whether they pertain to the power-law or the exponential regime. While the theories all assume that most of the deformation occurs homogeneously within the grains, most experimental observations point strongly to a large deformation at or close to the grain boundaries. However, a detailed study of dislocation processes in a single grain of polycrystalline foil strained in the electron microscope shows that most of the observed strain can be accounted for by the motion of single dislocations through the subgrain structure. There is no clear reconciliation of these two sets of observations. Grain-boundary sliding cannot occur without intragranular deformation. One or other process may dominate the overall deformation; the geometrically dominant process may not be the rate

  11. Phenomenological uncertainty analysis of early containment failure at severe accident of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Su Won

    2011-02-15

    The severe accident has inherently significant uncertainty due to wide range of conditions and performing experiments, validation and practical application are extremely difficult because of its high temperature and pressure. Although internal and external researches were put into practice, the reference used in Korean nuclear plants were foreign data of 1980s and safety analysis as the probabilistic safety assessment has not applied the newest methodology. Also, it is applied to containment pressure formed into point value as results of thermal hydraulic analysis to identify the probability of containment failure in level 2 PSA. In this paper, the uncertainty analysis methods for phenomena of severe accident influencing early containment failure were developed, the uncertainty analysis that apply Korean nuclear plants using the MELCOR code was performed and it is a point of view to present the distribution of containment pressure as a result of uncertainty analysis. Because early containment failure is important factor of Large Early Release Frequency(LERF) that is used as representative criteria of decision-making in nuclear power plants, it was selected in this paper among various modes of containment failure. Important phenomena of early containment failure at severe accident based on previous researches were comprehended and methodology of 7th steps to evaluate uncertainty was developed. The MELCOR input for analysis of the severe accident reflected natural circulation flow was developed and the accident scenario for station black out that was representative initial event of early containment failure was determined. By reviewing the internal model and correlation for MELCOR model relevant important phenomena of early containment failure, the uncertainty factors which could affect on the uncertainty were founded and the major factors were finally identified through the sensitivity analysis. In order to determine total number of MELCOR calculations which can

  12. Noise diagnosis - a method for early detection of failures in a nuclear plant

    International Nuclear Information System (INIS)

    Brinckmann, H.F.

    1981-01-01

    Noise diagnosis constitutes one method for early detection of plant failures. The method is based on the fact that nearly all undesired processes in a nuclear power plant make a measurable contribution to the noise portion of signals. Well-known examples of undesired processes in pressurized water reactors include core-barrel movement, the vibration of control elements, the appearance of loose parts in the coolant flow, and the process of coolant boiling. Each of these processes has been implicated in past nuclear plant failures. In the German Democratic Republic (GDR) P. Liewers and his colleagues have introduced noise analysis systems into the primary circuit of WWER-440 pressurized water reactors (PWR). The most progressive version (RAS-II) has become a prototype for research and routine investigations. This system is described. (author)

  13. Creep strength and ductility of 9 to 12% chromium steels

    DEFF Research Database (Denmark)

    Hald, John

    2004-01-01

    The present paper focuses in on long-term creep properties of parent material of the new 9-12%Cr creep resistant steels, P91, E911 and P92 developed for use in advanced ultrasupercritical power plants. These steels have been at the center of activities in the ECCC Working Group 3A (WG3A) "Ferriti...

  14. microstructure change in 12 % Cr steel during creep

    International Nuclear Information System (INIS)

    Winatapura, D. S.; Panjaitan, E.; Arslan, A.; Sulistioso, G.S.

    1998-01-01

    The microstructure change in steel containing of 12% Cr or DIN X20CrMoV 12 1 during creep has been studied by means of optical microscopy and Transmission Electron Microscope (TEM). The creep testing at 650 o C was conducted under constant load of 650 Mpa. The heat treatment of the specimen before creep testing was austenization, followed by tempering for 2 hours. The obtained microstructure was tempered martensitic. This microstructure consisted of the martensite laths, and distributed randomly in the matrix. During tempering, chromium carbide particles of Cr 7 C 6 less than 0,2 μmin-size were precipitated on or and in the subgrain and lath martensite grain boundary. During creep testing those particles transformed and precipitated as chrome carbide precipitates of Cr 23 C 6 . At the secondary creep stage, the void formation occurred, and then it developed into the creep cracks. At tertiary creep stage for 3554 hours, the specimen was failure. The creep cracks were informs of transgranular and intergranular modes which propagated almost perpendicular to the stress axis. From this observation, it is suggested that tempering caused the ductility of martensitic microstructure, which increased the creep resistant or Cr 12% steel

  15. Creep properties and simulation of weld repaired low alloy heat resistant CrMo and Mo steels at 540 deg C. Sub project 1 - Ex-serviced parent metal and virgin weld metals

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Storesund, Jan; Borggreen, Kjeld; Weilin Zang

    2006-10-15

    Many existing power generating and process plants, where low alloy heat resistant CrMo(V) steels are extensively used for critical components, have exceeded their design lifetime of usually 100,000 hours. Assessment of residual lifetime and extension of economic life by weld repair have become increasingly important and attractive. This project aims at i) performing weld repair and determining the degree of mismatching, ii) evaluating the creep properties of weld repairs, iii) analysing creep behaviour of weld repair and providing necessary data for further reliable simulations of weld repair creep behaviour in long term service, and iv), simulating and assessing lifetime and creep damage evolution of weld repair. Weld repair using 10 CrMo 9 10, 13 CrMo 4 4 and 15 Mo 3 consumables has been carried out in a service-exposed 10 CrMo 9 10 pipe. Creep specimens have been extracted from the service-exposed 10 CrMo 9 10 parent metal (PM), from the virgin 10 CrMo 9 10 weld metal (WM), from the virgin 13 CrMo 4 4 WM as well as from the virgin 15 Mo 3 WM. Iso-thermal uniaxial creep tests have been performed at 540 deg C in air. Pre- and post-metallography are carried out on the selected samples. FEM simulations using obtained creep data are executed. Pre-test metallography shows normal and acceptable weld repairs at given welding conditions. Creep tests demonstrate that the virgin 10 CrMo 9 10, 13 CrMo 4 4 and 15 Mo 3 WMs have apparently longer creep lifetime than the service-exposed CrMo 9 10 PM at higher stresses than 110 MPa. Among the weld metals, the longest creep lifetime is found in 10 CrMo 9 10. Higher creep strength and lower creep strain rate in the weld metals indicate an overmatch weld. At 95 MPa, however, lifetime of 13 CrMo 4 4 WM is surprisingly short (factors which may shorten lifetime are discussed and one more test will start to verify creep strength at low stress) and tests are still running for other two weld metals. More results regarding low stress

  16. Creep in buffer clay

    International Nuclear Information System (INIS)

    Pusch, R.; Adey, R.

    1999-12-01

    The study involved characterization of the microstructural arrangement and molecular forcefields in the buffer clay for getting a basis for selecting suitable creep models. It is concluded that the number of particles and wide range of the particle bond spectrum require that stochastical mechanics and thermodynamics will be considered and they are basic to the creep model proposed for predicting creep settlement of the canisters. The influence of the stress level on creep strain of MX-80 clay is not well known but for the buffer creep is approximately proportional to stress. Theoretical considerations suggest a moderate impact for temperatures up to 90 deg C and this is supported by model experiments. It is believed that the assumption of strain being proportional to temperature is conservative. The general performance of the stochastic model can be illustrated in principle by use of visco-elastic rheological models implying a time-related increase in viscosity. The shear-induced creep settlement under constant volume conditions calculated by using the proposed creep model is on the order of 1 mm in ten thousand years and up to a couple of millimeters in one million years. It is much smaller than the consolidation settlement, which is believed to be on the order of 10 mm. The general conclusion is that creep settlement of the canisters is very small and of no significance to the integrity of the buffer itself or of the canisters

  17. Measuring irradiation creep

    International Nuclear Information System (INIS)

    Pelah, I.

    1981-03-01

    Simulation of fusion-neutron induced damage by beams of light ions is discussed. It is suggested that accelerated creep measurements to determine ''end of life'' of materials may be done by the application of thermal treatment and thermal creep measurements. (author)

  18. Failure modes of safety-related components at fires on nuclear power plants

    International Nuclear Information System (INIS)

    Aaslund, A.

    2000-03-01

    Probabilistic assessment methods can be used to identify specific plant vulnerabilities. Application of such methods can also facilitate selection among system design alternatives available for safety enhancements. The quality of assessment results is however strongly dependent on realistic and accurate input data for modelling of system component behaviour and failure modes during conditions to be assessed. Use of conservative input data may not lead to results providing guidance on safety upgrades. Adequate input data for probabilistic assessments seems to be lacking for at least failure modes of some electrical components when exposed to a fire. This report presents an attempt to improve the situation with respect to such input data. In order to take advantage of information in existing documentation of fire incident occurrences some of the lessons learned from the fire at Browns Ferry Nuclear Power Plant on March 22, 1975 are discussed in this report. Also a summary of results from different fire tests of electrical cables presented in a fire risk analysis report is a part of the references. The failure modes used to describe fire-induced damage are 'open circuit' and 'hot short' which seems to be commonly accepted terms within the branch. Definitions of the terms are included in the report. Effects of the failure modes when occurring in some of the channels of the reactor protection system are discussed with respect to the existing design of the reactor protection system at Ringhals 2 nuclear power unit. Experiences from the Browns Ferry fire and results from fire tests of electrical cables indicate that the dominating failure mode for electrical cables is 'open circuit'. An 'open circuit' failure leads to circuit disjunction and loss of continuity. The circuit can no longer transmit its signal or power. When affecting channels of the reactor protection system an 'open circuit' failure can cause extensive inadvertent actions of safety related equipment

  19. Filtering technique for detection and identification of measurement failures in nuclear power plants

    International Nuclear Information System (INIS)

    Racz, A.

    1989-11-01

    The basic requirement of the safe operation of nuclear power plants (NPP) is to have reliable information on all quantities that can be measured, monitored or controlled during the operation. Kalman filtering techniques have been applied for prompt detection and identification of failures in the measurement systems used in NPPs. Mathematical basis of Kalman filtering and various models applied to failure detection are overviewed. The applicability of some models are evaluated by real results of NPP measurements. A sample system for an NPP is suggested, based on several numerical tests. (R.P.) 23 refs.; 40 figs.; 2 tabs

  20. Failure analysis of motor bearing of sea water pump in nuclear power plant

    International Nuclear Information System (INIS)

    Bian Chunhua; Zhang Wei

    2015-01-01

    The motor bearing of sea water pump in Qinshan Phase II Nuclear Power plant broke after only one year's using. This paper introduces failure analysis process of the motor bearing. Chemical composition analysis, metallic phase analysis, micrographic examination, and hardness analysis, dimension analysis of each part of the bearing, as well as the high temperature and low temperature performance analysis of lubricating grease are performed. According to the analysis above mentioned, the failure mode of the bearing is wearing, and the reason of wearing is inappropriate installation of the bearing. (authors)

  1. Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil [Sogang Univ., Seoul, (Korea, Republic of); Lee, Jin Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-11-15

    Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve.

  2. Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

    International Nuclear Information System (INIS)

    Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil; Lee, Jin Haeng

    2013-01-01

    Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve

  3. Creep properties of discontinuous fibre composites with partly creeping fibres

    International Nuclear Information System (INIS)

    Bilde-Soerensen, J.B.; Lilholt, H.

    1977-05-01

    In a previous report (RISO-M-1810) the creep properties of discontinuous fibre composites with non-creeping fibres were analyzed. In the present report this analysis is extended to include the case of discontinuous composites with partly creeping fibres. It is shown that the creep properties of the composite at a given strain rate, epsilonsub(c), depend on the creep properties of the matrix at a strain rate higher than epsilonsub(c), and on the creep properties of the fibres at epsilonsub(c). The composite creep law is presented in a form which permits a graphical determination of the composite creep curve. This can be constructed on the basis of the matrix and the fibre creep curves by vector operations in a log epsilon vs. log sigma diagram. The matrix contribution to the creep strength can be evaluated by a simple method. (author)

  4. Investigations of the Failure in Boilers Economizer Tubes Used in Power Plants

    Science.gov (United States)

    Moakhar, Roozbeh Siavash; Mehdipour, Mehrad; Ghorbani, Mohammad; Mohebali, Milad; Koohbor, Behrad

    2013-09-01

    In this study, failure of a high pressure economizer tube of a boiler used in gas-Mazut combined cycle power plants was studied. Failure analysis of the tube was accomplished by taking into account visual inspection, thickness measurement, and hardness testing as well as microstructural observations using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and x-ray diffraction (XRD). Optical microscopy images indicate that there is no phase transformation during service, and ferrite-pearlite remained. The results of XRD also revealed Iron sulfate (FeSO4) and Iron hydroxide sulfate (FeOH(SO4)) phases formed on the steel surface. A considerable amount of Sulfur was also detected on the outer surface of the tube by EDS analysis. Dew-point corrosion was found to be the principal reason for the failure of the examined tube while it has been left out-of-service.

  5. The prediction of creep damage in type 347 weld metal. Part I: the determination of material properties from creep and tensile tests

    International Nuclear Information System (INIS)

    Spindler, M.W.

    2005-01-01

    Calculations of creep damage under conditions of strain control are often carried out using either a time fraction approach or a ductility exhaustion approach. In the case of the time fraction approach the rupture strength is used to calculate creep damage, whereas creep ductility is used in the ductility exhaustion approach. In part I of this paper the methods that are used to determine these material properties are applied to some creep and constant strain rate tests on a Type 347 weld metal. In addition, new developments to the ductility exhaustion approach are described which give improved predictions of creep damage at failure in these tests. These developments use reverse modelling to determine the most appropriate creep damage model as a function of strain rate, stress and temperature. Hence, the new approach is no longer a ductility exhaustion approach but is a true creep damage model

  6. Experimental research on creep characteristics of Nansha soft soil.

    Science.gov (United States)

    Luo, Qingzi; Chen, Xiaoping

    2014-01-01

    A series of tests were performed to investigate the creep characteristics of soil in interactive marine and terrestrial deposit of Pearl River Delta. The secondary consolidation test results show that the influence of consolidation pressure on coefficient of secondary consolidation is conditional, which is decided by the consolidation state. The ratio of coefficient of secondary consolidation and coefficient of compressibility (Ca/Cc) is almost a constant, and the value is 0.03. In the shear-box test, the direct sheer creep failure of soil is mainly controlled by shear stress rather than the accumulation of shear strain. The triaxial creep features are closely associated with the drainage conditions, and consolidation can weaken the effect of creep. When the soft soil has triaxial creep damage, the strain rate will increase sharply.

  7. Examination of the creep behaviour of microstructurally unstable ferritic steels

    International Nuclear Information System (INIS)

    Williams, K.R.

    1981-01-01

    The inherent microstructural instability of 1/2Cr 1/2Mo 1/4V; 21/4Cr 1Mo and carbon steels creep tested or service exposed at low stresses is demonstrated. Measurements of important dispersion parameters have been made during creep life and have been found to follow normal coarsening kinetics. Using the measured time dependent change of the dispersion parameters, a dislocation source controlled model for recovery creep is used and further developed. The model allows the calculation of the Manson-Haferd plot of log (time to failure) against temperature for unstable steels. In addition, a classification of material stability is proposed, based on the ratio of time to fracture, t(sub f), and time to tertiary creep, tsub(t). This classification enables estimates of remaining creep life to be based either on well established criteria for stable materials or modifications of these criteria for unstable steels. (author)

  8. Creep damage development in welded X20 and P91

    Energy Technology Data Exchange (ETDEWEB)

    Brett, Steve; Holmstrom, Stefan; Hald, John; Borg, Ulrik; Aakjaer Jensen, Soeren; Vulpen, Rijk Van; Degnan, Craig; Vinter Dahl, Kristian; Vilhelmsen, Tommy

    2011-03-15

    The Martensitic steel X20CrMoV121 (hereinafter called X20) and the modified 9Cr1Mo steel (hereinafter called P91) have been used for a number of years in high temperature applications since they posses superior creep strength compared to low alloyed steels. Due to the simple fact that very few failures were observed, almost no knowledge as to the evolution of creep damage in welds were available despite long operation times exceeding well over 100.000 hours. It has been suggested that X20 will develop creep damage in a different manner compared to low alloyed steel, i.e damage initiation should be slow followed by accelerated growth. The research work presented in this report included systematic investigations of the first components of X20, which has developed creep during long-term operation. All of the investigated components showed creep damage evolution similar to low alloy steels

  9. Experimental Research on Creep Characteristics of Nansha Soft Soil

    Directory of Open Access Journals (Sweden)

    Qingzi Luo

    2014-01-01

    Full Text Available A series of tests were performed to investigate the creep characteristics of soil in interactive marine and terrestrial deposit of Pearl River Delta. The secondary consolidation test results show that the influence of consolidation pressure on coefficient of secondary consolidation is conditional, which is decided by the consolidation state. The ratio of coefficient of secondary consolidation and coefficient of compressibility Ca/Cc is almost a constant, and the value is 0.03. In the shear-box test, the direct sheer creep failure of soil is mainly controlled by shear stress rather than the accumulation of shear strain. The triaxial creep features are closely associated with the drainage conditions, and consolidation can weaken the effect of creep. When the soft soil has triaxial creep damage, the strain rate will increase sharply.

  10. Experimental Research on Creep Characteristics of Nansha Soft Soil

    Science.gov (United States)

    Luo, Qingzi; Chen, Xiaoping

    2014-01-01

    A series of tests were performed to investigate the creep characteristics of soil in interactive marine and terrestrial deposit of Pearl River Delta. The secondary consolidation test results show that the influence of consolidation pressure on coefficient of secondary consolidation is conditional, which is decided by the consolidation state. The ratio of coefficient of secondary consolidation and coefficient of compressibility C a/C c is almost a constant, and the value is 0.03. In the shear-box test, the direct sheer creep failure of soil is mainly controlled by shear stress rather than the accumulation of shear strain. The triaxial creep features are closely associated with the drainage conditions, and consolidation can weaken the effect of creep. When the soft soil has triaxial creep damage, the strain rate will increase sharply. PMID:24526925

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. Multiple-failure signal validation in nuclear power plants using artificial neural networks

    International Nuclear Information System (INIS)

    Fantoni, P.F.; Mazzola, A.

    1996-01-01

    The possibility of using a neural network to validate process signals during normal and abnormal plant conditions is explored. In boiling water reactor plants, signal validation is used to generate reliable thermal limits calculation and to supply reliable inputs to other computerized systems that support the operator during accident scenarios. The way that autoassociative neural networks can promptly detect faulty process signal measurements and produce a best estimate of the actual process values even in multifailure situations is shown. A method was developed to train the network for multiple sensor-failure detection, based on a random failure simulation algorithm. Noise was artificially added to the input to evaluate the network's ability to respond in a very low signal-to-noise ratio environment. Training and test data sets were simulated by the real-time transient simulator code APROS

  13. Real-time sensor failure detection by dynamic modelling of a PWR plant

    International Nuclear Information System (INIS)

    Turkcan, E.; Ciftcioglu, O.

    1992-06-01

    Signal validation and sensor failure detection is an important problem in real-time nuclear power plant (NPP) surveillance. Although conventional sensor redundancy, in a way, is a solution, identification of faulty sensor is necessary for further preventive actions to be taken. A comprehensive solution for the system so that any sensory reading is verified by its model based estimated counterpart, in real-time. Such a realization is accomplished by means of dynamic system's states estimation methodology using Kalman filter modelling technique. The method is investigated by means of real-time data of the steam generator of Borssele nuclear power plant and the method has proved to be satisfactory for real-time sensor failure detection as well as model validation verification. (author). 5 refs.; 6 figs.; 1 tab

  14. A Study on Estimating the Next Failure Time of Compressor Equipment in an Offshore Plant

    Directory of Open Access Journals (Sweden)

    SangJe Cho

    2016-01-01

    Full Text Available The offshore plant equipment usually has a long life cycle. During its O&M (Operation and Maintenance phase, since the accidental occurrence of offshore plant equipment causes catastrophic damage, it is necessary to make more efforts for managing critical offshore equipment. Nowadays, due to the emerging ICTs (Information Communication Technologies, it is possible to send health monitoring information to administrator of an offshore plant, which leads to much concern on CBM (Condition-Based Maintenance. This study introduces three approaches for predicting the next failure time of offshore plant equipment (gas compressor with case studies, which are based on finite state continuous time Markov model, linear regression method, and their hybrid model.

  15. Steam explosions-induced containment failure studies for Swiss nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zuchuat, O.; Schmocker, U. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland); Esmaili, H.; Khatib-Rahbar, M.

    1998-01-01

    The assessment of the consequences of both in-vessel and ex-vessel energetic fuel-coolant interaction for Beznau (a Westinghouse pressurized water reactor with a large, dry containment), Goesgen (a Siemens/KWU pressurized water reactor with a large, dry containment) and Leibstadt (a General Electric boiling water reactor-6 with a free standing steel, MARK-III containment) nuclear power plants is presented in this paper. The Conditional Containment Failure Probability of the steel containment of these Swiss nuclear power plants is determined based on different probabilistic approaches. (author)

  16. Basic factors to forecast maintenance cost and failure processes for nuclear power plants

    International Nuclear Information System (INIS)

    Popova, Elmira; Yu, Wei; Kee, Ernie; Sun, Alice; Richards, Drew; Grantom, Rick

    2006-01-01

    Two types of maintenance interventions are usually administered at nuclear power plants: planned and corrective. The cost incurred includes the labor (manpower) cost, cost for new parts, or emergency order of expensive items. At the plant management level there is a budgeted amount of money to be spent every year for such operations. It is very important to have a good forecast for this cost since unexpected events can trigger it to a very high level. In this research we present a statistical factor model to forecast the maintenance cost for the incoming month. One of the factors is the expected number of unplanned (due to failure) maintenance interventions. We introduce a Bayesian model for the failure rate of the equipment, which is input to the cost forecasting model. The importance of equipment reliability and prediction in the commercial nuclear power plant is presented along with applicable governmental and industry organization requirements. A detailed statistical analysis is performed on a set of maintenance cost and failure data gathered at the South Texas Project Nuclear Operating Company (STPNOC) in Bay City, Texas, USA

  17. Evaluation of a Kalman filter based power pressurizer instrument failure detection system implemented on a nuclear power plant training simulator

    International Nuclear Information System (INIS)

    Seegmiller, D.S.

    1984-01-01

    The usefulness of a nuclear power plant training simulator for developing and testing modern estimation and control applications for nuclear power plants is demonstrated. A Kalman filter based instrument failure detection technique for a pressurized water reactor pressurizer is implemented on the Department of Energy N Reactor Training Simulator. This real-time failure detection method computes the first two moments (mean and variance) of each element of a normalized filter innovations vector. Failed pressurizer instrumentation can be detected by comparing these moments to the known statistical properties of the steady state, linear Kalman fitler innovations sequence. The capabilities of the detection system are evaluated using simulated plant transients and instrument failures

  18. Accelerated Creep Testing of High Strength Aramid Webbing

    Science.gov (United States)

    Jones, Thomas C.; Doggett, William R.; Stnfield, Clarence E.; Valverde, Omar

    2012-01-01

    A series of preliminary accelerated creep tests were performed on four variants of 12K and 24K lbf rated Vectran webbing to help develop an accelerated creep test methodology and analysis capability for high strength aramid webbings. The variants included pristine, aged, folded and stitched samples. This class of webbings is used in the restraint layer of habitable, inflatable space structures, for which the lifetime properties are currently not well characterized. The Stepped Isothermal Method was used to accelerate the creep life of the webbings and a novel stereo photogrammetry system was used to measure the full-field strains. A custom MATLAB code is described, and used to reduce the strain data to produce master creep curves for the test samples. Initial results show good correlation between replicates; however, it is clear that a larger number of samples are needed to build confidence in the consistency of the results. It is noted that local fiber breaks affect the creep response in a similar manner to increasing the load, thus raising the creep rate and reducing the time to creep failure. The stitched webbings produced the highest variance between replicates, due to the combination of higher local stresses and thread-on-fiber damage. Large variability in the strength of the webbings is also shown to have an impact on the range of predicted creep life.

  19. Dependence of irradiation creep on temperature and atom displacements in 20% cold worked type 316 stainless steel

    International Nuclear Information System (INIS)

    Gilbert, E.R.

    1976-04-01

    Irradiation creep studies with pressurized tubes of 20 percent cold worked Type 316 stainless steel were conducted in EBR-2. Results showed that as atom displacements are extended above 5 dpa and temperatures are increased above 375 0 C, the irradiation induced creep rate increases with both increasing atom displacements and increasing temperature. The stress exponent for irradiation induced creep remained near unity. Irradiation-induced effective creep strains up to 1.8 percent were observed without specimen failure. 13 figures

  20. Seismic Creep, USA Images

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Seismic creep is the constant or periodic movement on a fault as contrasted with the sudden rupture associated with an earthquake. It is a usually slow deformation...

  1. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  2. Biaxial Creep Specimen Fabrication

    International Nuclear Information System (INIS)

    JL Bump; RF Luther

    2006-01-01

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments

  3. Biaxial Creep Specimen Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    JL Bump; RF Luther

    2006-02-09

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.

  4. Micromechanical studies of cyclic creep fracture under stress controlled loading

    DEFF Research Database (Denmark)

    van der Giessen, Erik; Tvergaard, Viggo

    1996-01-01

    is based on numerical unit cell analyses for a planar polycrystal model with the grains and grain boundaries modeled individually, in order to investigate the interactions between the mechanisms involved and to account for the build-up of residual stress fields during cycling. The behaviour of a limiting......This paper deals with a study of intergranular failure by creep cavitation under stress-controlled cyclic loading conditions. Loading is assumed to be slow enough that diffusion and creep mechanisms (including grain boundary sliding) dominate, leading to intergranular creep fracture. This study...

  5. Creep Tests and Modeling Based on Continuum Damage Mechanics for T91 and T92 Steels

    Science.gov (United States)

    Pan, J. P.; Tu, S. H.; Zhu, X. W.; Tan, L. J.; Hu, B.; Wang, Q.

    2017-12-01

    9-11%Cr ferritic steels play an important role in high-temperature and high-pressure boilers of advanced power plants. In this paper, a continuum damage mechanics (CDM)-based creep model was proposed to study the creep behavior of T91 and T92 steels at high temperatures. Long-time creep tests were performed for both steels under different conditions. The creep rupture data and creep curves obtained from creep tests were captured well by theoretical calculation based on the CDM model over a long creep time. It is shown that the developed model is able to predict creep data for the two ferritic steels accurately up to tens of thousands of hours.

  6. Limit analysis via creep

    International Nuclear Information System (INIS)

    Taroco, E.; Feijoo, R.A.

    1981-07-01

    In this paper it is presented a variational method for the limit analysis of an ideal plastic solid. This method has been denominated as Modified Secundary Creep and enables to find the collapse loads through a minimization of a functional and a limit process. Given an ideal plastic material it is shown how to determinate the associated secundary creep constitutive equation. Finally, as an application, it is found the limit load in an pressurized von Mises rigid plastic sphere. (Author) [pt

  7. A dynamic failure evaluation of a simplified digital control system of a nuclear power plant pressurizer

    International Nuclear Information System (INIS)

    Pinto, J.M.O.; Melo, P.F. Frutuoso e; Saldanha, P.L.C.

    2010-01-01

    Given the increasing use of digital systems in nuclear power plants, a specific approach to reliability and risk analysis has been required. The digital system reflects many interactions between hardware, software, process variables, and human actions. At the same time, the software, does not have a reliability approach as well-defined as the one existing for the other physical components of the system. Then, its reliability analysis is still under development due to difficulties arising from the complexity, flexibility and interactions present in such systems.The traditional approach of using fault trees is static and does not approach the dynamic interactions in such systems, such as delays in capture and processing information, memory, logic loops, system states, etc. It is necessary to find a reliability methodology that takes into account these issues without violating the existing requirements concerning safety analysis, such as: ability to distinguish between common-cause failures, availability of relevant information to users, like minimal cut sets, and failure probabilities as long as the possibility of incorporating the results into existing probabilistic safety assessments (PSA).One approach is to trace all the possible errors of the digital system through dynamic methodologies. The DFM (Dynamic Flow-graph Methodology) is one of the methodologies that better meets the requirements for modeling dynamic systems. It discretizes the most relevant variables of the analyzed system in states that reflect their behavior, sets the logic that connects them through decision tables and finally performs a system analysis, aiming, for example, the root causes (prime implicants) of a given top event of failure. Three aspects have been addressed, the modeling of the system itself, the incorporation of results to probabilistic safety analyses and identification of software failures.To illustrate the DFM, a simplified digital control system of a typical PWR pressurizer

  8. Probability of failure of the waste hoist brake system at the Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    Greenfield, M.A.; Sargent, T.J.; Stanford Univ., CA

    1998-01-01

    In its most recent report on the annual probability of failure of the waste hoist brake system at the Waste Isolation Pilot Plant (WIPP), the annual failure rate is calculated to be 1.3E(-7)(1/yr), rounded off from 1.32E(-7). A calculation by the Environmental Evaluation Group (EEG) produces a result that is about 4% higher, namely 1.37E(-7)(1/yr). The difference is due to a minor error in the US Department of Energy (DOE) calculations in the Westinghouse 1996 report. WIPP's hoist safety relies on a braking system consisting of a number of components including two crucial valves. The failure rate of the system needs to be recalculated periodically to accommodate new information on component failure, changes in maintenance and inspection schedules, occasional incidents such as a hoist traveling out-of-control, either up or down, and changes in the design of the brake system. This report examines DOE's last two reports on the redesigned waste hoist system. In its calculations, the DOE has accepted one EEG recommendation and is using more current information about the component failures rates, the Nonelectronic Parts Reliability Data (NPRD). However, the DOE calculations fail to include the data uncertainties which are described in detail in the NPRD reports. The US Nuclear Regulatory Commission recommended that a system evaluation include mean estimates of component failure rates and take into account the potential uncertainties that exist so that an estimate can be made on the confidence level to be ascribed to the quantitative results. EEG has made this suggestion previously and the DOE has indicated why it does not accept the NRC recommendation. Hence, this EEG report illustrates the importance of including data uncertainty using a simple statistical example

  9. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    International Nuclear Information System (INIS)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho

    2016-01-01

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B 4 C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  10. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B{sub 4}C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  11. Theoretical and experimental investigations of creep buckling on NiCr 22 Co 12 Mo tubes

    International Nuclear Information System (INIS)

    Ahmed, K.; Breitbach, G.; Over, H.; Schubert, F.; Nickel, H.

    1988-08-01

    The postulated pressure loss of the secondary circuit is one of the hardest loading conditions for the heat exchanging components in a HTGR plant. It is to proof for the design that the heat exchanging metallic components (heat exchanger or reformer tubes of a PNP plant for instance) do not collapse under such an emergency condition. An external pressure p a stressed tubes or cylindric shells at a pressure loss of a secondary circuit side. This external pressure buckles the tubes in dependence of the fabrication implied out of roundness 0 (fabrication tolerances) by material creep in the high temperature region. This creep buckling ends in a failure (collapse) of the component after a critical time t cr . The aim of the work is the experimental verification of creep buckling behaviour for the heat exchanger components and the comparison with different constitutive equations. With these equations safety factors can be formulated against as well the critical collapse time and pressure as the permissible out of roundness from fabrication. (orig.) [de

  12. Dead or Alive? Using Membrane Failure and Chlorophyll a Fluorescence to Predict Plant Mortality from Drought.

    Science.gov (United States)

    Guadagno, Carmela R; Ewers, Brent E; Speckman, Heather N; Aston, Timothy Llewellyn; Huhn, Bridger J; DeVore, Stanley B; Ladwig, Joshua T; Strawn, Rachel N; Weinig, Cynthia

    2017-09-01

    Climate models predict widespread increases in both drought intensity and duration in the next decades. Although water deficiency is a significant determinant of plant survival, limited understanding of plant responses to extreme drought impedes forecasts of both forest and crop productivity under increasing aridity. Drought induces a suite of physiological responses; however, we lack an accurate mechanistic description of plant response to lethal drought that would improve predictive understanding of mortality under altered climate conditions. Here, proxies for leaf cellular damage, chlorophyll a fluorescence, and electrolyte leakage were directly associated with failure to recover from drought upon rewatering in Brassica rapa (genotype R500) and thus define the exact timing of drought-induced death. We validated our results using a second genotype (imb211) that differs substantially in life history traits. Our study demonstrates that whereas changes in carbon dynamics and water transport are critical indicators of drought stress, they can be unrelated to visible metrics of mortality, i.e. lack of meristematic activity and regrowth. In contrast, membrane failure at the cellular scale is the most proximate cause of death. This hypothesis was corroborated in two gymnosperms ( Picea engelmannii and Pinus contorta ) that experienced lethal water stress in the field and in laboratory conditions. We suggest that measurement of chlorophyll a fluorescence can be used to operationally define plant death arising from drought, and improved plant characterization can enhance surface model predictions of drought mortality and its consequences to ecosystem services at a global scale. © 2017 American Society of Plant Biologists. All Rights Reserved.

  13. Creep fatigue design of FBR components

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1997-01-01

    This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

  14. Creep feeding nursing beef calves.

    Science.gov (United States)

    Lardy, Gregory P; Maddock, Travis D

    2007-03-01

    Creep feeding can be used to increase calf weaning weights. However, the gain efficiency of free-choice, energy-based creep feeds is relatively poor. Generally, limit-feeding, high-protein creep feeds are more efficient, and gains may be similar to those produced by creep feeds offered free choice. Creep feeding can increase total organic matter intake and improve the overall energy status of the animal. Creep-fed calves tend to acclimate to the feedlot more smoothly than unsupplemented calves. Furthermore, provision of a high-starch creep feed may have a positive influence on subsequent carcass quality traits. Creep feeding can be applied to numerous environmental situations to maximize calf performance; however, beef cattle producers should consider their individual situations carefully before making the decision to creep feed.

  15. Correlated seed failure as an environmental veto to synchronize reproduction of masting plants.

    Science.gov (United States)

    Bogdziewicz, Michał; Steele, Michael A; Marino, Shealyn; Crone, Elizabeth E

    2018-07-01

    Variable, synchronized seed production, called masting, is a widespread reproductive strategy in plants. Resource dynamics, pollination success, and, as described here, environmental veto are possible proximate mechanisms driving masting. We explored the environmental veto hypothesis, which assumes that reproductive synchrony is driven by external factors preventing reproduction in some years, by extending the resource budget model of masting with correlated reproductive failure. We ran this model across its parameter space to explore how key parameters interact to drive seeding dynamics. Next, we parameterized the model based on 16 yr of seed production data for populations of red (Quercus rubra) and white (Quercus alba) oaks. We used these empirical models to simulate seeding dynamics, and compared simulated time series with patterns observed in the field. Simulations showed that resource dynamics and reproduction failure can produce masting even in the absence of pollen coupling. In concordance with this, in both oaks, among-year variation in resource gain and correlated reproductive failure were necessary and sufficient to reproduce masting, whereas pollen coupling, although present, was not necessary. Reproductive failure caused by environmental veto may drive large-scale synchronization without density-dependent pollen limitation. Reproduction-inhibiting weather events are prevalent in ecosystems, making described mechanisms likely to operate in many systems. © 2018 The Authors New Phytologist © 2018 New Phytologist Trust.

  16. Study on Flexural Creep Parameters of Overlayed Particleboard by Natural and Melaminated Veneers

    Directory of Open Access Journals (Sweden)

    Abdollah Najafi

    2012-06-01

    Full Text Available In this study, effects of natural and artificial veneer on flexural creep behavior of particleboard was investigated. Particleboard panels were prepared from Pars Neopan industries with 660 kg/m3 density and then overlaid by natural and melamine veneers. Their creep behavior was compared to control particleboard. For evaluating maximum bending load in static flexural test, specimens were cut from panels according to ASTM D 1037 with dimensions of 370×50×16 mm. Then, The flexural creep tests at 20% and 40% of failure bending load was applied to test specimens. Results of flexural tests indicated that the MOR and MOE values of veneered particleboard were highest. Results of creep showed that levels of stresses are effective on all creep parameters, but showed less effect on relative creep. Also, creep parameters less effective on specimens overlaid by natural veneer.

  17. Development of Web-Based Common Cause Failure (CCF) Database Module for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Hyun-Gyo; Hwang, Seok-Won; Shin, Tae-young

    2015-01-01

    Probabilistic safety assessment (PSA) has been used to identify risk vulnerabilities and derive the safety improvement measures from construction to operation stages of nuclear power plants. In addition, risk insights from PSA can be applied to improve the designs and operation requirements of plants. However, reliability analysis methods for quantitative PSA evaluation have essentially inherent uncertainties, and it may create a distorted risk profiles because of the differences among the PSA models, plant designs, and operation status. Therefore, it is important to ensure the quality of the PSA model so that analysts identify design vulnerabilities and utilize risk information. Especially, the common cause failure (CCF) has been pointed out as one of major issues to be able to cause the uncertainty related to the PSA analysis methods and data because CCF has a large influence on the PSA results. Organization for economic cooperation and development /nuclear energy agent (OECD/NEA) has implemented an international common cause failure data exchange (ICDE) project for the CCF quality assurance through the development of the detailed analysis methodologies and data sharing. However, Korea Hydro and Nuclear Power company (KHNP) does not have the basis for the data gathering and analysis for CCF analyses. In case of methodology, the Alpha Factor parameter estimation, which can analyze uncertainties and estimate an interface factor (Impact Vector) with an ease, is ready to be applied rather than the Multi Greek Letter (MGL) method. This article summarizes the development of the plant-specific CCF database (DB) module considering the raw data collection and the analysis procedure based on the CCF parameter calculation method of ICDE. Although the portion affected by CCF in the PSA model is quite a large, the development efforts of the tools to collect and analyze data were insufficient. Currently, KHNP intends to improve PSA quality and ensure CCF data reliability by

  18. Development of Web-Based Common Cause Failure (CCF) Database Module for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun-Gyo; Hwang, Seok-Won; Shin, Tae-young [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Probabilistic safety assessment (PSA) has been used to identify risk vulnerabilities and derive the safety improvement measures from construction to operation stages of nuclear power plants. In addition, risk insights from PSA can be applied to improve the designs and operation requirements of plants. However, reliability analysis methods for quantitative PSA evaluation have essentially inherent uncertainties, and it may create a distorted risk profiles because of the differences among the PSA models, plant designs, and operation status. Therefore, it is important to ensure the quality of the PSA model so that analysts identify design vulnerabilities and utilize risk information. Especially, the common cause failure (CCF) has been pointed out as one of major issues to be able to cause the uncertainty related to the PSA analysis methods and data because CCF has a large influence on the PSA results. Organization for economic cooperation and development /nuclear energy agent (OECD/NEA) has implemented an international common cause failure data exchange (ICDE) project for the CCF quality assurance through the development of the detailed analysis methodologies and data sharing. However, Korea Hydro and Nuclear Power company (KHNP) does not have the basis for the data gathering and analysis for CCF analyses. In case of methodology, the Alpha Factor parameter estimation, which can analyze uncertainties and estimate an interface factor (Impact Vector) with an ease, is ready to be applied rather than the Multi Greek Letter (MGL) method. This article summarizes the development of the plant-specific CCF database (DB) module considering the raw data collection and the analysis procedure based on the CCF parameter calculation method of ICDE. Although the portion affected by CCF in the PSA model is quite a large, the development efforts of the tools to collect and analyze data were insufficient. Currently, KHNP intends to improve PSA quality and ensure CCF data reliability by

  19. Metallurgical failure analysis for a blade failed in a gas-turbine engine of a power plant

    International Nuclear Information System (INIS)

    Huda, Zainul

    2009-01-01

    The failed gas-turbine blades (first stage blades) (type Siemens V94.2 KWU) were acquired from TNB Research Sdn. Bhd: a subsidiary of Malaysian power-generation industry (TNB, Malaysia). The blades were sectioned for metallographic investigations. The microstructural characterization involved use of both optical as well as electron microscopes including application of EPMA technique. The Microstructures were compared for three spots selection i.e. leading edge of the blade (transverse and longitudinal), trailing edge of the blade (transverse and longitudinal), and centre (near the platform of the blade) (transverse and longitudinal). The material properties and behavior at high temperature were interpreted on the basis of the observed microstructures and the phases present in the alloy. The interpretations were related to the operating conditions of the turbine blade; and main cause of failure was found to be creep damage. Recommendations have been made for improved material performance.

  20. Creep-fatigue evaluation method for type 304 and 316FR SS

    International Nuclear Information System (INIS)

    Wada, Y.; Aoto, K.; Ueno, F.

    1997-01-01

    For long-term creep-fatigue of Type 304SS, intergranular failure is dominant in the case of significant life reduction. It is considered that this phenomenon has its origin in the grain boundary sliding as observed in cavity-type creep-rupture. Accordingly a simplified procedure to estimate intergranular damages caused by the grain boundary sliding is presented in connection with the secondary creep. In the conventional ductility exhaustion method, failure ductility includes plastic strain, and damage estimation is based on the primary creep strain, which is recoverable during strain cycling. Therefore the accumulated creep strain becomes a very large value, and quite different from grain boundary sliding strain. As a new concept on ductility exhaustion, the product of secondary creep rate and time to rupture (Monkman-Grant product) is applied to fracture ductility, and grain boundary sliding strain is approximately estimated using the accumulated secondary creep strain. From the new concept it was shown that the time fraction rule and the conventional ductility exhaustion method can be derived analytically. Furthermore an advanced method on cyclic stress relaxation was examined. If cyclic plastic strain hardening is softened thermally during strain hold, cyclic creep strain behaviour is also softened. An unrecoverable accumulated primary creep strain causes hardening of the primary creep, and the reduction of deformation resistance to the secondary creep caused by thermal softening accelerates grain boundary sliding rate. As the results creep damages depend not on applied stress but on effective stress. The new concept ductility exhaustion method based on the above consideration leads up to simplified time fraction estimation method only by continuous cycling fatigue and monotonic creep which was already developed in PNC for Monju design guide. This method gave good life prediction for the intergranular failure mode and is convenient for design use on the elastic

  1. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre

    International Nuclear Information System (INIS)

    2007-05-01

    This note of information resumes the progress of the electric failure on the reactor n.3 of the nuclear power plant of Dampierre, the organization during the incident, it establishes then a comparison with the incident arisen to Forsmark in 2006 and reminds that it lead in an inspection on behalf of the Asn which noticed that all the procedures had been respected by the operators and did not noticed any abnormality in the maintenance. This event was classified at the level 1 of the international nuclear event scale (INES). (N.C.)

  2. Pilot program to identify valve failures which impact the safety and operation of light water nuclear power plants

    International Nuclear Information System (INIS)

    Tsacoyeanes, J.C.; Raju, P.P.

    1980-04-01

    The pilot program described has been initiated under the Department of Energy Light Water Reactor Safety Research and Development Program and has the following specific objectives: to identify the principal types and causes of failures in valves, valve operators and their controls and associated hardware, which lead to, or could lead to plant trip; and to suggest possible remedies for the prevention of these failures and recommend future research and development programs which could lead to minimizing these valve failures or mitigating their effect on plant operation. The data surveyed cover incidents reported over the six-year period, beginning 1973 through the end of 1978. Three sources of information on valve failures have been consulted: failure data centers, participating organizations in the nuclear power industry, and technical documents

  3. Deep knowledge expert system for diagnosis of multiple-failure severe transients in nuclear power plant

    International Nuclear Information System (INIS)

    Martin, R.P.; Nassersharif, B.

    1987-01-01

    TAMUS (Transient Analysis of MUltiple-failure Simulations) is a prototype expert system which is the result of a project investigating and implementing event confidence-levels (used by reactor safety experts in reactor transient analysis) in the form of an expert system. Currently, TAMUS is designed to diagnose reactor transients by analyzing simulated sensor and plant thermal hydraulic information from a system simulation. TAMUS uses a knowledge base of existing emergency nuclear plant operating guidelines and detailed thermal-hydraulic calculating results correlated to confidence-levels. TAMUS can diagnose a number of reactor transients (for example, loss-of-coolant accidents, steam-generator-tube ruptures, loss-of-offsite power, etc.). Future work includes the expansion of the knowledge base and improvement of the deep-knowledge qualitative models

  4. Evaluation of Component Failure Data of the Operating Nuclear Power Plants in Korea Based on NUREG/CR-6928

    International Nuclear Information System (INIS)

    Jeon, Hojun; Na, Janghwan; Shin, Taeyoung

    2014-01-01

    This paper focuses on ensuring the quality of component failure data. When performing data analysis in PSA, we have customized the component failure data based on Bayesian analysis using plant specific experiences and the generic data of Advanced Light Water Reactor Utility Requirements Document (ALWR URD). However, ALWR URD was established by collecting US nuclear power plant (NPP) practices from mid 1980s to early 1990s. We analyzed the component failure data using the raw data of component failures in Pressurized Water Reactor (PWR) plants by 2012. This paper presents the results from analyzing the component failure data based on the new generic data and the latest specific failure data. We also compare the new component failure data to the existing data of PSA models, and evaluate the risk impacts by applying the new data to the PSA models of reference NPPs in this paper. To apply the new generic data source to PSA models, we reviewed and compared NUREG/CR-6928 and the existing generic data source, ALWR URD. In addition, we analyzed the component failure data generated from 16 PWR plants by the end of 2012, and performed the Bayesian update with these raw data based on the new generic data source of NUREG/CR-6928. Also, we reviewed the PSA models of the reference NPP, and identified some important components to CDF. The failure data of the major components decreased in general by applying the new generic data and the latest plant specific data. As a result, the CDF of the reference NPP decreased over 30% compared to the value of the existing CDF

  5. Changes in creep of polymethylmetacrylate after irradiation

    International Nuclear Information System (INIS)

    Peschanskaya, N.N.; Smolyanskij, A.S.; Suvorova, V.Yu.

    1992-01-01

    A study was made on PMMA, irradiated by different doses of 60 Co γ-radiation in vacuum under creep during compression. It is shown that occurence of tendency to failure at +20 degC is observed at doses of D > 100 kGy (> 10 Mrad), whereas sufficient decrease of deformation before failure takes place at D > 350 kGy. Peculiarities of behaviour of irradiated and nonirradiated PMMA under compression and tension were correlated. It is noted that critical irradiation doses may differ sufficiently for different loading conditions, deformation and longevity characteristics

  6. Inductive analysis of failure patterns and of their impact on thermohydraulic circuits of nuclear power plants

    International Nuclear Information System (INIS)

    Limnios, N.

    1983-01-01

    The APACHE code (Automatic Analysis of Failures of Hydraulic and Thermohydraulic Circuits more particularly of Water) situates in an important program of computer codes development in the field of studies on reliability and safety of systems in nuclear power plants. APACHE is an automatic generation code of failure pattern and of their effects. After a presentation of the theoretical basis, the methodological principles of the theory of networks are developed. Then, the model of the code is developed: model of individual behavior of each classical model component of normal behavior and model of failure pattern with specifications. The global model of hydraulic systems and the resolution systems are then developed. More particularly, some aspects of the theory of graphs, and the algorithms developed for the automatic construction of the equation systems and especially the algorithm of the research of meshes are presented. The computer aspect of the code and the programming of the code with its limits and some specifications are described. The practical aspect of utilization is finally presented [fr

  7. A discrete dislocation dynamics model of creeping single crystals

    Science.gov (United States)

    Rajaguru, M.; Keralavarma, S. M.

    2018-04-01

    Failure by creep is a design limiting issue for metallic materials used in several high temperature applications. Current theoretical models of creep are phenomenological with little connection to the underlying microscopic mechanisms. In this paper, a bottom-up simulation framework based on the discrete dislocation dynamics method is presented for dislocation creep aided by the diffusion of vacancies, known to be the rate controlling mechanism at high temperature and stress levels. The time evolution of the creep strain and the dislocation microstructure in a periodic unit cell of a nominally infinite single crystal is simulated using the kinetic Monte Carlo method, together with approximate constitutive laws formulated for the rates of thermal activation of dislocations over local pinning obstacles. The deformation of the crystal due to dislocation glide between individual thermal activation events is simulated using a standard dislocation dynamics algorithm, extended to account for constant stress periodic boundary conditions. Steady state creep conditions are obtained in the simulations with the predicted creep rates as a function of stress and temperature in good agreement with experimentally reported values. Arrhenius scaling of the creep rates as a function of temperature and power-law scaling with the applied stress are also reproduced, with the values of the power-law exponents in the high stress regime in good agreement with experiments.

  8. Micro creep mechanisms of tungsten

    International Nuclear Information System (INIS)

    Levoy, R.; Hugon, I.; Burlet, H.; Baillin, X.; Guetaz, L.

    2000-01-01

    Due to its high melting point (3410 deg C), tungsten offers good mechanical properties at elevated temperatures for several applications in non-oxidizing environment. The creep behavior of tungsten is well known between 1200 and 2500 deg C and 10 -3 to 10 -1 strain. However, in some applications when dimensional stability of components is required, these strains are excessive and it is necessary to know the creep behavior of the material for micro-strains (between 10 -4 and 10 -6 ). Methods and devices used to measure creep micro-strains are presented, and creep equations (Norton and Chaboche laws) were developed for wrought, annealed and recrystallized tungsten. The main results obtained on tungsten under low stresses are: stress exponent 1, symmetry of micro-strains in creep-tension and creep-compression, inverse creep (threshold stress), etc. TEM, SEM and EBSD studies allow interpretation of the micro-creep mechanism of tungsten under low stresses and low temperature (∼0.3 K) like the Harper-Dorn creep. In Harper-Dorn creep, micro-strains are associated with the density and the distribution of dislocations existing in the crystals before creep. At 975 deg C, the initial dislocation structure moves differently whether or not a stress is applied. To improve the micro-creep behavior of tungsten, a heat treatment is proposed to create the optimum dislocation structure. (authors)

  9. Correlation of electrical reactor cable failure with materials degradation

    Energy Technology Data Exchange (ETDEWEB)

    Stuetzer, O.M.

    1986-03-01

    Complete circuit failure (shortout) of electrical cables typically used in nuclear power plant containments is investigated. Failure modes are correlated with the mechanical deterioration of the elastomeric cable materials. It is found that for normal reactor operation, electrical cables are reliable and safe over very long periods. During high temperature excursions, however, cables pulled across corners under high stress may short out due to conductor creep. Severe cracking will occur in short times during high temperatures (>150/sup 0/C) and in times of the order of years at elevated temperatures (100/sup 0/C to 140/sup 0/C). A theoretical treatment of stress distribution responsible for creep and for cracking by J.E. Reaugh of Science Applications, Inc. is contained in the Appendix. 29 refs., 32 figs.

  10. Correlation of electrical reactor cable failure with materials degradation

    International Nuclear Information System (INIS)

    Stuetzer, O.M.

    1986-03-01

    Complete circuit failure (shortout) of electrical cables typically used in nuclear power plant containments is investigated. Failure modes are correlated with the mechanical deterioration of the elastomeric cable materials. It is found that for normal reactor operation, electrical cables are reliable and safe over very long periods. During high temperature excursions, however, cables pulled across corners under high stress may short out due to conductor creep. Severe cracking will occur in short times during high temperatures (>150 0 C) and in times of the order of years at elevated temperatures (100 0 C to 140 0 C). A theoretical treatment of stress distribution responsible for creep and for cracking by J.E. Reaugh of Science Applications, Inc. is contained in the Appendix. 29 refs., 32 figs

  11. Consistent creep and rupture properties for creep-fatigue evaluation

    International Nuclear Information System (INIS)

    Schultz, C.C.

    1978-01-01

    The currently accepted practice of using inconsistent representations of creep and rupture behaviors in the prediction of creep-fatigue life is shown to introduce a factor of safety beyond that specified in current ASME Code design rules for 304 stainless steel Class 1 nuclear components. Accurate predictions of creep-fatigue life for uniaxial tests on a given heat of material are obtained by using creep and rupture properties for that same heat of material. The use of a consistent representation of creep and rupture properties for a mininum strength heat is also shown to provide adequate predictions. The viability of using consistent properties (either actual or those of a minimum heat) to predict creep-fatigue life thus identifies significant design uses for the results of characterization tests and improved creep and rupture correlations

  12. Consistent creep and rupture properties for creep-fatigue evaluation

    International Nuclear Information System (INIS)

    Schultz, C.C.

    1979-01-01

    The currently accepted practice of using inconsistent representations of creep and rupture behaviors in the prediction of creep-fatigue life is shown to introduce a factor of safety beyond that specified in current ASME Code design rules for 304 stainless steel Class 1 nuclear components. Accurate predictions of creep-fatigue life for uniaxial tests on a given heat of material are obtained by using creep and rupture properties for that same heat of material. The use of a consistent representation of creep and rupture properties for a minimum strength heat is also shown to provide reasonable predictions. The viability of using consistent properties (either actual or those of a minimum strength heat) to predict creep-fatigue life thus identifies significant design uses for the results of characterization tests and improved creep and rupture correlations. 12 refs

  13. Development of a constitutive model for the plastic deformation and creep of copper and its use in the estimate of the creep life of the copper canister

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2006-12-01

    A previously developed model for the plastic deformation and creep of copper (included as an Appendix to the present report) has been used as the basis for a discussion on the possibility of brittle creep fracture of the copper canister during long term storage of nuclear waste. Reported creep tests on oxygen free (OF) copper have demonstrated that copper can have an extremely low creep ductility. However with the addition of about 50 ppm phosphorus to the copper it appears as if the creep brittleness problem is avoided and that type of copper (OFP) has consequently been chosen as the canister material. It is shown in the report that the experiments performed on OFP copper does not exclude the possibility of creep brittleness of OFP copper in the very long term. The plasticity and creep model has been used to estimate creep life under conditions of intergranular creep cracking according to a model formulated by Cocks and Ashby. The estimated life times widely exceed the design life of the canister. However the observations of creep brittleness in OF copper indicate that the Cocks-Ashby model probably does not apply to the OF copper. Thus additional calculations have been done with the plasticity and creep model in order to estimate stress as a function of time for the probably most severe loading case of the canister with regard to creep failure, an earth quake shear. Despite the fact that the stress in the canister will remain at the 100 MPa level for thousands of years after an earth quake the low temperature, about 50 deg C or less, will make the solid state diffusion process assumed to control the brittle cracking process, too slow to lead to any significant brittle creep cracking in the canister

  14. Evaluation of the material in creep-exposed critical 12Cr-1Mo-V ferritic steel components of the pressure section of power plant steam boilers

    International Nuclear Information System (INIS)

    Dobrzanski, J.

    2002-01-01

    Metallographic examinations were carried out on ferritic 12Cr-1Mo-V steel with tempered martensite structure after 68,000 - 145,000 hr service in creep conditions. Structural evolution related to the form of martensite, carbide precipitation and internal damage due to long-term action of thermally activated processes was discussed. A generalised scheme of structural evolution and progress of the internal damage was developed in correlation with the life exhaustion ratio. Principles of the classification for ferritic 12Cr-1Mo-V steel after long-term-service were proposed together with the method for evolution of the state of the material for the industrial practice. (author)

  15. Prediction of creep-fatigue life by use of creep rupture ductility

    International Nuclear Information System (INIS)

    Yamaguchi, Koji; Suzuki, Naoyuki; Ijima, Kiyoshi; Kanazawa, Kenji

    1985-01-01

    It was clarified that tension strain hold reduced creep-fatigue life of many engineering materials in different degrees depending on material, temperature and test duration. However the reduction in the life due to holding for various durations could be correlated to the fraction of intergranular facets on fracture surfaces which was considered to be an index of the damage introduced during strain hold. This fraction of intergranular facets by creep-fatigue failure exhibited a direct relation to the creep rupture ductility of the material tested at the same temperature and for the same creep-fatigue life-time. From these results an empirical equation has been derived as follow; (Δ sub(epsilonsub(i)))/Dsub(c).(N sub(h sup(α))) = C, where Δ sub(epsilonsub(i)) is inelastic strain range, Dsub(c) is the creep rupture ductility for the same duration as creep-fatigue life time, Nsub(h) is the creep-fatigue life under tension strain hold conditions, and α and C are constants depending on the material and testing temperature. From the equation the life prediction is possible for a given inelastic strain range Δ sub(epsilonsub(i)) if the constants α and C, and Dsub(c) are known. The value of α was found to be 0.62 and 0.74 for various austenitic stainless steels and NCF800 at 600 0 C and 700 0 C, respectively, and 0.69 for 1 1/4Cr-1/2Mo steel at 600 0 C. The value of C was found to be 0.50 and 0.59 for various austenitic stainless steels and NCF800 at 600 0 C and 700 0 C, respectively, and 0.49 for 1 1/4Cr-1/2Mo steel at 600 0 C. The creep rupture ductility Dsub(c) is available in the NRIM Creep Data Sheets up to 10 5 h for multi-heats of many kinds of heat resistant alloys. (author)

  16. Creep of timber joints

    NARCIS (Netherlands)

    Van de Kuilen, J.W.G.

    2008-01-01

    A creep analysis has been performed on nailed, toothed-plates and split-ring joints in a varying uncontrolled climate. The load levels varied between 30% and 50% of the average ultimate short term strength of these joints, tested in accordance with ISO 6891. The climate in which the tests were

  17. Analysis of indentation creep

    Science.gov (United States)

    Don S. Stone; Joseph E. Jakes; Jonathan Puthoff; Abdelmageed A. Elmustafa

    2010-01-01

    Finite element analysis is used to simulate cone indentation creep in materials across a wide range of hardness, strain rate sensitivity, and work-hardening exponent. Modeling reveals that the commonly held assumption of the hardness strain rate sensitivity (mΗ) equaling the flow stress strain rate sensitivity (mσ...

  18. Creep buckling of shell structures

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki; Hagihara, Seiya

    2015-01-01

    The present article contains a review of the literatures on the creep buckling of shell structures published from late 1950's to recent years. In this article, the creep buckling studies on circular cylindrical shells, spherical shells, partial cylindrical shells and other shells are reviewed in addition to creep buckling criteria. Creep buckling is categorized into two types. One is the creep buckling due to quasi-static instability, in which the critical time for creep buckling is determined by tracing a creep deformation versus time curve. The other is the creep buckling due to kinetic instability, in which the critical time can be determined by examining the shape of total potential energy in the vicinity of a quasi-static equilibrium state. Bifurcation buckling and snap-through buckling during creep deformation belong to this type of creep buckling. A few detailed descriptions are given to the bifurcation and snap-through type of creep buckling based on the present authors' works. (author)

  19. Life assessment of Mod.9Cr-1Mo steel. Quantitative evaluation of microstructural damage in creep interrupted specimens and in creep-fatigue specimens

    International Nuclear Information System (INIS)

    Maruyama, Kouichi; Kato, Syoichi; Nagae, Yuji

    1999-02-01

    Boiler and steam turbine components in power generating plants are used under creep and creep-fatigue conditions. It is important to measure both creep and creep-fatigue damage of the components in order to assess the residual life of the components. Modified 9Cr-1Mo steel, a candidate material for steam generator in FBR, has a tempered martensitic lath structure. It was proposed in the second report that lath width in the lath structure is closely related to creep strain, and using this relation one can assess residual creep life of a structural component made of the steel. The objectives of this study are to investigate the change of the lath structure during creep.fatigue deformation, and to estimate creep strain by measuring area of cell composing the lath structure. The area of cell can be a better measure of creep deformation than the lath width. The lath structure is covered during creep-fatigue deformation. The lath structure becomes equiaxed cell structure under creep-fatigue more quickly compared with the lath structure recovered during creep. The lath structure recovered under creep-fatigue has a stationary value of the lath width determined by maximum stress at Nf/2. (Nf: number of cycles) If the recovery process of the lath structure can be investigated under creep-fatigue, the lath width can be a measure of the life assessment under creep-fatigue. Area of cell composing the lath structure increases with creep deformation and reaches a stationary value S s determined by creep stress. The rate of increase in the area is faster at a higher stress and temperature. A normalized change in the area of cell, ΔS/ΔS s , was introduced as a measure of the recovery process of martensitic lath structure. ΔS is the change in area of cell from the initial value S 0 , ΔS s is the difference between S s and S 0 . ΔS/ΔS s is uniquely related to creep strain independent of creep conditions. However, the scatter of data in ΔS/ΔS s -strain relation is wider than

  20. Strength and life under creeping

    International Nuclear Information System (INIS)

    Pospishil, B.

    1982-01-01

    Certain examples of the application of the Lepin modified creep model, which are of interest from technical viewpoint, are presented. Mathematical solution of the dependence of strength limit at elevated temperatures on creep characteristics is obtained. Tensile test at elevated temperatures is a particular case of creep or relaxation and both strength limit and conventional yield strength at elevated temperatures are completely determined by parameters of state equations during creep. The equation of fracture summing during creep is confirmed not only by the experiment data when stresses change sporadically, but also by good reflection of durability curve using the system of equations. The system presented on the basis of parameters of the equations obtained on any part of durability curve, permits to forecast the following parameters of creep: strain, strain rate, life time, strain in the process of fracture. Tensile test at elevated temperature is advisable as an addition when determining creep curves (time-strain curves) [ru

  1. Irradiation creep models - an overview

    International Nuclear Information System (INIS)

    Matthews, J.R.; Finnis, M.W.

    1988-01-01

    The modelling of irradiation creep is now highly developed but many of the basic processes underlying the models are poorly understood. A brief introduction is given to the theory of cascade interactions, point defect clustering and dislocation climb. The range of simple irradiation creep models is reviewed including: preferred nucleation of interstitial loops; preferred absorption of point defects by dislocations favourably orientated to an applied stress; various climb-enhanced glide and recovery mechanisms, and creep driven by internal stresses produced by irradiation growth. A range of special topics is discussed including: cascade effects; creep transients; structural and induced anisotropy; and the effect of impurities. The interplay between swelling and growth with thermal and irradiation creep is emphasized. A discussion is given on how irradiation creep theory should best be developed to assist the interpretation of irradiation creep observations and the requirements of reactor designers. (orig.)

  2. Creep fatigue damage under multiaxial conditions

    International Nuclear Information System (INIS)

    Lobitz, D.W.; Nickell, R.E.

    1977-01-01

    When structural components are subjected to severe cyclic loading conditions with intermittent periods of sustained loading at elevated temperature, the designer must guard against a failure mode caused by the interaction of time-dependent and time-independent deformation. This phenomena is referred to as creep-fatigue interaction. The most elementary form of interaction theory (called linear damage summation) is now embodied in the ASME Boiler and Pressure Vessel Code. In recent years, a competitor for the linear damage summation theory has emerged, called strainrange partitioning. This procedure is based upon the visualization of the cyclic strain in a uniaxial creep-fatigue test as a hysteresis loop, with the inelastic strains in the loop counter-balanced in one of two ways. The two theories are compared and contrasted in terms of ease of use, possible inconsistencies, and component life prediction. Future work to further test the damage theories is recommended

  3. Creep behaviour and creep mechanisms of normal and healing ligaments

    Science.gov (United States)

    Thornton, Gail Marilyn

    Patients with knee ligament injuries often undergo ligament reconstructions to restore joint stability and, potentially, abate osteoarthritis. Careful literature review suggests that in 10% to 40% of these patients the graft tissue "stretches out". Some graft elongation is likely due to creep (increased elongation of tissue under repeated or sustained load). Quantifying creep behaviour and identifying creep mechanisms in both normal and healing ligaments is important for finding clinically relevant means to prevent creep. Ligament creep was accurately predicted using a novel yet simple structural model that incorporated both collagen fibre recruitment and fibre creep. Using the inverse stress relaxation function to model fibre creep in conjunction with fibre recruitment produced a superior prediction of ligament creep than that obtained from the inverse stress relaxation function alone. This implied mechanistic role of fibre recruitment during creep was supported using a new approach to quantify crimp patterns at stresses in the toe region (increasing stiffness) and linear region (constant stiffness) of the stress-strain curve. Ligament creep was relatively insensitive to increases in stress in the toe region; however, creep strain increased significantly when tested at the linear region stress. Concomitantly, fibre recruitment was evident at the toe region stresses; however, recruitment was limited at the linear region stress. Elevating the water content of normal ligament using phosphate buffered saline increased the creep response. Therefore, both water content and fibre recruitment are important mechanistic factors involved in creep of normal ligaments. Ligament scars had inferior creep behaviour compared to normal ligaments even after 14 weeks. In addition to inferior collagen properties affecting fibre recruitment and increased water content, increased glycosaminoglycan content and flaws in scar tissue were implicated as potential mechanisms of scar creep

  4. Multi-axial Creep and the LICON Methodology for Accelerated Creep Testing

    International Nuclear Information System (INIS)

    Bowyer, William H.

    2006-05-01

    The copper-Iron canister for disposal of nuclear waste in the Swedish Programme has a design life exceeding 100,000 years. Whilst the operating temperature (100 deg C max.) and operating stress (50 MPa max.) are modest, the very long design life does require that the likely creep performance of the canister should be investigated. Many studies have been carried out by SKB but these have all involved very short duration tests at relatively high stresses. The process of predicting canister creep life by extrapolation of data from such tests has been challenged for two main reasons. The first is that the deformation and failure mechanisms in the tests employed are different from the mechanism expected under service conditions and the second is that the extrapolation is extreme. It has been recognised that there is usually scope for some increase in test temperatures and stresses which will accelerate the development of creep damage without compromising the use of extrapolation for life prediction. Cane demonstrated that in steels designed for high temperature and pressure applications, conditions of multi-axial stressing could lead to increases or decreases in the rate of damage accumulation without changing the damage mechanism. This provided a third method for accelerating creep testing which has been implemented as the LICON method. This report aims to explain the background to the LICON method and its application to the case of the copper canister. It seems likely that the method could be used to improve our knowledge of the creep resistance of the copper canister. Multiplication factors that may be achieved by the technique could be increased by attention to specimen design but an extensive and targeted programme of data collection on creep of copper would still be needed to implement the method to best advantage

  5. Multi-axial Creep and the LICON Methodology for Accelerated Creep Testing

    Energy Technology Data Exchange (ETDEWEB)

    Bowyer, William H. [Meadow End Farm, Farnham (United Kingdom)

    2006-05-15

    The copper-Iron canister for disposal of nuclear waste in the Swedish Programme has a design life exceeding 100,000 years. Whilst the operating temperature (100 deg C max.) and operating stress (50 MPa max.) are modest, the very long design life does require that the likely creep performance of the canister should be investigated. Many studies have been carried out by SKB but these have all involved very short duration tests at relatively high stresses. The process of predicting canister creep life by extrapolation of data from such tests has been challenged for two main reasons. The first is that the deformation and failure mechanisms in the tests employed are different from the mechanism expected under service conditions and the second is that the extrapolation is extreme. It has been recognised that there is usually scope for some increase in test temperatures and stresses which will accelerate the development of creep damage without compromising the use of extrapolation for life prediction. Cane demonstrated that in steels designed for high temperature and pressure applications, conditions of multi-axial stressing could lead to increases or decreases in the rate of damage accumulation without changing the damage mechanism. This provided a third method for accelerating creep testing which has been implemented as the LICON method. This report aims to explain the background to the LICON method and its application to the case of the copper canister. It seems likely that the method could be used to improve our knowledge of the creep resistance of the copper canister. Multiplication factors that may be achieved by the technique could be increased by attention to specimen design but an extensive and targeted programme of data collection on creep of copper would still be needed to implement the method to best advantage.

  6. Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O., E-mail: vasconv@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2016-07-01

    During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

  7. Evaluation of piping reliability and failure data for use in risk-based inspections of nuclear power plants

    International Nuclear Information System (INIS)

    Vasconcelos, V. de; Soares, W.A.; Costa, A.C.L. da; Rabello, E.G.; Marques, R.O.

    2016-01-01

    During operation of industrial facilities, components and systems can deteriorate over time, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) involves inspection planning based on information about risks, through assessing of probability and consequence of failures. In-service inspections are used in nuclear power plants, in order to ensure reliable and safe operation. Traditional deterministic inspection approaches investigate generic degradation mechanisms on all systems. However, operating experience indicates that degradation occurs where there are favorable conditions for developing a specific mechanism. Inspections should be prioritized at these places. Risk-Informed In-service Inspections (RI-ISI) are types of RBI that use Probabilistic Safety Assessment results, increasing reliability and plant safety, and reducing radiation exposure. These assessments use both available generic reliability and failure data, as well as plant specific information. This paper proposes a method for evaluating piping reliability and failure data important for RI-ISI programs, as well as the techniques involved. (author)

  8. Experimental study and modelling of high temperature creep flow and damage behaviour of 9Cr1Mo-NbV steel weldments; Etude experimentale et modelisation, du comportement, de l'endommagement et de la rupture en fluage a haute temperature de joint soudes en acier 9Cr1Mo-NbV

    Energy Technology Data Exchange (ETDEWEB)

    Gaffard, V

    2004-12-15

    Chromium martensitic stainless steels are under development since the 70's with the prospect of using them as structural components in thermal and nuclear power plants. The modified 9Cr1Mo-NbV steel is already used, especially in England and Japan, as a material for structural components in thermal power plants where welding is a commonly used joining technique. New generations of chromium martensitic stainless steels with improved mechanical properties for high pressure and temperature use are currently under development. However, observations of several in-service premature failures of welded components in 9Cr1Mo-NbV steel, outline a strong need for understanding the high temperature creep flow and damage behaviour of 9Cr1Mo-NbV steels and weldments. The present study aimed at experimentally determining and then modelling the high temperature creep flow and damage behaviour of both 9Cr1Mo-NbV steels and weldments (typically in the temperature range from 450 C to 650 C). The base metal was first studied as the reference material. It was especially evidenced that tempered chromium martensitic steels exhibit a change in both creep flow and damage behaviour for long term creep exposure. As a consequence, the classically performed extrapolation of 1,000 hours creep data to 100,000 hours creep lifetime predictions might be very hazardous. Based on experimental observations, a new model, integrating and coupling multiple creep flow and damage mechanisms, was developed in the framework of the mechanics of porous media. It was then successfully used to represent creep flow and damage behaviour of the base metal from high to low stress levels even for complex multiaxial loading conditions. Although the high temperature creep properties of the base metal are quite good, the occurrence of premature failure in weldments in high temperature creep conditions largely focused the attention of the scientific community. The lower creep strength of the weld component was also

  9. Resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants

    International Nuclear Information System (INIS)

    Johnson, R.E.

    1990-06-01

    This report describes the US Nuclear Regulatory Commission's (NRC's) Generic Safety Issue 29, ''Bolting Degradation or Failure in Nuclear Power Plants,'' including the bases for establishing the issue and its historical highlights. The report also describes the activities of the Atomic Industrial Forum (AIF) relevant to this issue, including its cooperation with the Materials Properties Council (MPC) to organize a task group to help resolve the issue. The Electric Power Research Institute, supported by the AIF/MPC task group, prepared and issued a two-volume document that provides, in part, the technical basis for resolving Generic Safety Issue 29. This report presents the NRC's review and evaluation of the two-volume document and NRC's conclusion that this document, in conjunction with other information from both industry and NRC, provides the bases for resolving this issue

  10. Continuous observation of cavity growth and coalescence by creep-fatigue tests in SEM

    International Nuclear Information System (INIS)

    Arai, Masayuki; Ogata, Takashi; Nitta, Akito

    1995-01-01

    Structural components operating at high temperatures in power plants are subjected to interaction of thermal fatigue and creep which results in creep-fatigue damage. In evaluating the life of those components, it is important to understand microscopic damage evolution under creep-fatigue conditions. In this study, static creep and creep-fatigue tests with tensile holdtime were conducted on SUS304 stainless steel by using a high-temperature fatigue machine combined with a scanning electron microscope (SEM), and cavity growth and coalescence behaviors on surface grain boundaries were observed continuously by the SEM. Quantitative analysis of creep cavity growth based on the observation was made for comparison with theoretical growth models. As a result, it was found that grain boundary cavities nucleate at random and grow preferentially on grain boundaries in a direction almost normal to the stress axis. Under the creep condition, the cavities grow monotonously on grain boundaries while they remain the elliptical shape. On the other hand, under the creep-fatigue condition the cavities grow with an effect of local strain distribution around the grain boundary due to cyclic loading and the micro cracks of one grain-boundary length were formed by coalescence of the cavities. Also, cavity nucleation and growth rates for creep-fatigue were more rapid than those for static creep and the constrained cavity growth model coincided well with the experimental data for creep. (author)

  11. The Sensitivity Analysis of Axial Pressure Tube Creep Profile for Dryout Power in PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Euiseung; Kim, Youngae [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Stern Laboratory performed the CHF tests with only one axial pressure tube creep profile per 3.3%, 5.1% peak crept channel and made CHF correlation including creep factor from the CHF test results. Wolsong nuclear power plants also have utilized the same CHF correlation derived by CNL. Pressure tube diameter creep rate is function of fast neutron, coolant temperature, and coolant pressure in a channel. It means that various axial pressure tube creep profiles exist in PHWR due to the history of operating conditions. Usually, CHF correlation is used during ROP(Regional Overpower Protection) Trip Setpoint Analysis or Safety Analysis in PHWR. The sensitivity analysis for CHF effects using various creep profiles is needed. This paper summarizes the comparison results of dryout power between CHF test creep profile and estimated creep profiles of Wolsong units. The effect of axial pressure tube creep profile for dryout power in fuel channel is evaluated by using Stern Lab. CHF test creep profile and 380 channel creep profiles of Wolsong. The dryout powers at 3.3% and 5.1% test conditions are slightly smaller when using 380 Wolsong channels creep profiles. These also show that the simulated dryout powers maintain consistency regardless of flow conditions.

  12. Recent Methodologies for Creep Deformation Analysis and Its Life Prediction

    International Nuclear Information System (INIS)

    Kim, Woo-Gon; Park, Jae-Young; Iung

    2016-01-01

    To design the high-temperature creeping materials, various creep data are needed for codification, as follows: i) stress vs. creep rupture time for base metals and weldments (average and minimum), ii) stress vs. time to 1% total strain (average), iii) stress vs. time to onset of tertiary creep (minimum), and iv) constitutive eqns. for conducting time- and temperature- dependent stress-strain (average), and v) isochronous stress-strain curves (average). Also, elevated temperature components such as those used in modern power generation plant are designed using allowable stress under creep conditions. The allowable stress is usually estimated on the basis of up to 10"5 h creep rupture strength at the operating temperature. The master curve of the “sinh” function was found to have a wider acceptance with good flexibility in the low stress ranges beyond the experimental data. The proposed multi-C method in the LM parameter revealed better life prediction than a single-C method. These improved methodologies can be utilized to accurately predict the long-term creep life or strength of Gen-IV nuclear materials which are designed for life span of 60 years

  13. Some factors influencing the creep behaviour of alloy 800

    International Nuclear Information System (INIS)

    Asbury, F.E.; Willoughby, G.

    1975-01-01

    Studies have been made of the stability of the creep behaviour of two commercial casts of Incoloy 800, one high carbon and the other low carbon. The effects of pre-ageing, of prolonged creep up to 10 4 hours duration, and of grain size were investigated. Three factors were found to excercise a major influence on creep behaviour. Firstly, when the high carbon alloy was heat treated at 1150degC super-saturation effects, ascribed principally to carbon, gave some initial strengthening which would not, however, persist for the duration of service life in nuclear power plant applications above 600degC. Secondly, a gamma-dash type phase precipitated readily at 550 to 600degC, giving a marked increase in creep strength. Nucleation was sluggish at higher temperatures but once established, this form of strengthening could persist up to at least 650degC. Creep under non-isothermal conditions at 600 to 700degC would be complex on account of the behaviour of this phase. The hardening associated with its precipitation was greater in the low carbon alloy. Finally it was demonstrated that, in spite of gamma-dash precipitation, fine grained low carbon material was weak in creep at low stresses and temperatures. This was ascribed to the occurrence of grain boundary diffusion creep. It appears that this source of weakening would persist in service, and severely restrict the maximum temperature of usage for fined grained high tensile material. (author)

  14. Creep-fatique interactions in 316 stainless steel under torsional loading

    International Nuclear Information System (INIS)

    Wei, K.; Dyson, B.F.

    1982-01-01

    Some fatigue, fatigue with creep dwells and creep tests have been performed in torsion using 316 stainless steel at 600 0 C. As expected from push-pull testing, the introduction of a creep dwell reduced fatigue endurances and changed the fracture from classical transgranular to intergranular. Optical microscopical examination revealed a large number of intergranular cracks concentrated along shear planes, but quantitative assessment identified the importance of creep tensile stresses in crack development. In contrast, little intergranular damage was found after torsion creep, which is consistent with its exhibited buckling mode of failure. It is concluded that reverse plastic strain is the cause of intergranular crack formation in the material and is therefore the primary mechanism of creep-fatigue interaction. (author)

  15. Creep and creep-rupture behavior of Alloy 718

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Booker, M.K.; Ding, J.L.

    1991-01-01

    Data obtained from creep and creep-rupture tests conducted on 18 heats of Alloy 718 were used to formulate models for predicting high temperature time dependent behavior of this alloy. Creep tests were conducted on specimens taken from a number of commercial product forms including plate, bar, and forgoing material that had been procured and heat treated in accordance with ASTM specifications B-670 or B-637. Data were obtained over the temperature range of 427 to 760 degree C ad at test times to about 87,000 h. Comparisons are given between experimental data and the analytical models. The analytical models for creep-rupture included one based on lot-centering regression analysis and two based on the Minimum Commitment Method. A ''master'' curve approach was used to develop and equation for estimating creep deformation up to the onset of tertiary creep. 11 refs., 13 figs

  16. Description of Concrete Creep under Time-Varying Stress Using Parallel Creep Curve

    OpenAIRE

    Park, Yeong-Seong; Lee, Yong-Hak; Lee, Youngwhan

    2016-01-01

    An incremental format of creep model was presented to take account of the development of concrete creep due to loading at different ages. The formulation was attained by introducing a horizontal parallel assumption of creep curves and combining it with the vertical parallel creep curve of the rate of creep method to remedy the disadvantage of the rate of creep method that significantly underestimates the amount of creep strain, regardless of its simple format. Two creep curves were combined b...

  17. Improved methods of creep-fatigue life assessment of components

    Energy Technology Data Exchange (ETDEWEB)

    Scholz, Alfred; Berger, Christina [Inst. fuer Werkstoffkunde (IfW), Technische Univ. Darmstadt (Germany)

    2009-07-01

    The improvement of life assessment methods contributes to a reduction of efforts at design and an effective long term operation of high temperature components, reduces technical risk and increases high economical advantages. Creep-fatigue at multi-stage loading, covering cold start, warm start and hot start cycles in typical loading sequences e.g. for medium loaded power plants, was investigated here. At hold times creep and stress relaxation, respectively, lead to an acceleration of crack initiation. Creep fatigue life time can be calculated by a modified damage accumulation rule, which considers the fatigue fraction rule for fatigue damage and the life fraction rule for creep damage. Mean stress effects, internal stress and interaction effects of creep and fatigue are considered. Along with the generation of advanced creep data, fatigue data and creep fatigue data as well scatter band analyses are necessary in order to generate design curves and lower bound properties inclusive. Besides, in order to improve lifing methods the enhancement of modelling activities for deformation and life time are important. For verification purposes, complex experiments at variable creep conditions as well as at creep fatigue interaction under multi-stage loading are of interest. Generally, the development of methods to transfer uniaxial material properties to multiaxial loading situations is a current challenge. For specific design purposes, a constitutive material model is introduced which is implemented as an user subroutine for Finite Element applications due to start-up and shut-down phases of components. Identification of material parameters have been performed by Neural Networks. (orig.)

  18. Simulation of irradiation creep

    International Nuclear Information System (INIS)

    Reiley, T.C.; Jung, P.

    1977-01-01

    The results to date in the area of radiation enhanced deformation using beams of light ions to simulate fast neutron displacement damage are reviewed. A comparison is made between these results and those of in-reactor experiments. Particular attention is given to the displacement rate calculations for light ions and the electronic energy losses and their effect on the displacement cross section. Differences in the displacement processes for light ions and neutrons which may effect the irradiation creep process are discussed. The experimental constraints and potential problem areas associated with these experiments are compared to the advantages of simulation. Support experiments on the effect of thickness on thermal creep are presented. A brief description of the experiments in progress is presented for the following laboratories: HEDL, NRL, ORNL, PNL, U. of Lowell/MIT in the United States, AERE Harwell in the United Kingdom, CEN Saclay in France, GRK Karlsruhe and KFA Julich in West Germany

  19. Quantitative analysis of microstructure deformation in creep fenomena of ferritic SA-213 T22 and austenitic SA-213 TP304H material

    Science.gov (United States)

    Mulyana, Cukup; Taufik, Ahmad; Gunawan, Agus Yodi; Siregar, Rustam Efendi

    2013-09-01

    The failure of critical component of fossil fired power plant that operated in creep range (high stress, high temperature and in the long term) depends on its microstructure characteristics. Ferritic low carbon steel (2.25Cr-1Mo) and Austenitic stainless alloy (18Cr-8Ni) are used as a boiler tube in the secondary superheater outlet header to deliver steam before entering the turbin. The tube failure is occurred in a form of rupture, resulting trip that disrupts the continuity of the electrical generation. The research in quantification of the microstructure deformation has been done in predicting the remaining life of the tube through interrupted accelerated creep test. For Austenitic Stainless Alloy (18Cr-8Ni), creep test was done in 550°C with the stress 424.5 MPa and for Ferritic Low Carbon Steel (2.25Cr-1Mo) in 570°C with the stress 189 MPa. The interrupted accelerated creep test was done by stopping the observation in condition 60%, 70%, 80% and 90% of remaining life, the creep test fracture was done before. Then the micro hardness test, photo micro, SEM and EDS were obtained from those samples. Refer to ASTM E122, microstructure parameters were calculated. The results indicated that there are a consistency of decreasing their grain diameters, increasing their grain size numbers, micro hardness, and the length of crack or void number per unit area with the decreasing of remaining life. While morphology of grain (stated in parameter α=LV/LH) relatively constant for austenitic. However, for ferritic the change of morphology revealed significantly. Fracture mode propagation of ferritic material is growth with voids transgranular and intergranular crack, and for austenitic material the fracture growth with intergranular creep fracture void and wedge crack. In this research, it was proposed a formulation of mathematical model for creep behavior corresponding their curve fitting resulted for the primary, secondary and tertiary in accelerated creep test. In

  20. Small ring testing of a creep resistant material

    International Nuclear Information System (INIS)

    Hyde, C.J.; Hyde, T.H.; Sun, W.; Nardone, S.; De Bruycker, E.

    2013-01-01

    Many components in conventional and nuclear power plant, aero-engines, chemical plant etc., operate at temperatures which are high enough for creep to occur. These include steam pipes, pipe branches, gas and steam turbine blades, etc. The manufacture of such components may also require welds to be part of them. In most cases, only nominal operating conditions (i.e. pressure, temperatures, system load, etc.) are known and hence precise life predictions for these components are not possible. Also, the proportion of life consumed will vary from position to position within a component. Hence, non-destructive techniques are adopted to assist in making decisions on whether to repair, continue operating or replace certain components. One such approach is to test a small sample removed from the component to make small creep test specimens which can be tested to give information on the remaining creep life of the component. When such a small sample cannot be removed from the operating component, e.g. in the case of small components, the component can be taken out of operation in order to make small creep test specimens, the results from which can then be used to assist with making decisions regarding similar or future components. This paper presents a small creep test specimen which can be used for the testing of particularly strong and creep resistant materials, such as nickel-based superalloys

  1. Creep in sodium

    International Nuclear Information System (INIS)

    Charnock, W.; Cordwell, J.E.

    1978-03-01

    Available information on the creep of austenitic, ferritic and Alloy-800 type steels in liquid sodium is critically reviewed. Creep properties of stainless steels can be affected by element transfer and corrosion. At reactor structural component temperatures environmental effects are likely to be less important than changes due to thermal ageing. At high clad temperatures (700 0 C) decarburisation may cause the loss of strength and ductility in unstabilised steels while cavity formation may cause embrittlement in stabilised steels. The properties of Alloy 800 are, in some experiments, found to deteriorate while in others they are enhanced. This may be a consequence of the metallurgical complexity of the material or arise from the nature of the various techniques employed. Low alloy ferritic steels tend to decarburise in sodium at temperatures greater than 500 0 C and this leads to loss of strength and an increase in ductility. High alloy ferritics are immune to this effect and appear to be able to tolerate a degree of carburisation. Although intergranular cracking may be enhanced in liquid sodium the mechanical consequences are not significant and evidence for the existence of an embrittlement effect not associated with element transfer or corrosion is weak. Stress and strain may enhance element transfer at crack tips. However in real cracks the gettering or supply action of the crack faces conditions the chemistry of the cracks in sodium and protects the crack tip from element transfer. Thus creep crack extension rates should be independent of changes in bulk coolant chemistry. (author)

  2. High temperature creep of vanadium

    International Nuclear Information System (INIS)

    Juhasz, A.; Kovacs, I.

    1978-01-01

    The creep behaviour of polycrystalline vanadium of 99.7% purity has been investigated in the temperature range 790-880 0 C in a high temperature microscope. It was found that the creep properties depend strongly on the history of the sample. To take this fact into account some additional properties such as the dependence of the yield stress and the microhardness on the pre-annealing treatment have also been studied. Samples used in creep measurements were selected on the basis of their microhardness. The activation energy of creep depends on the microhardness and on the creep temperature. In samples annealed at 1250 0 C for one hour (HV=160 kgf mm -2 ) the rate of creep is controlled by vacancy diffusion in the temperature range 820-880 0 C with an activation energy of 78+-8 kcal mol -1 . (Auth.)

  3. Creep of fibrous composite materials

    DEFF Research Database (Denmark)

    Lilholt, Hans

    1985-01-01

    Models are presented for the creep behaviour of fibrous composite materials with aligned fibres. The models comprise both cases where the fibres remain rigid in a creeping matrix and cases where the fibres are creeping in a creeping matrix. The treatment allows for several contributions...... to the creep strength of composites. The advantage of combined analyses of several data sets is emphasized and illustrated for some experimental data. The analyses show that it is possible to derive creep equations for the (in situ) properties of the fibres. The experiments treated include model systems...... such as Ni + W-fibres, high temperature materials such as Ni + Ni3Al + Cr3C2-fibres, and medium temperature materials such as Al + SiC-fibres. For the first two systems reasonable consistency is found for the models and the experiments, while for the third system too many unquantified parameters exist...

  4. Numerical algorithms in secondary creep

    International Nuclear Information System (INIS)

    Feijoo, R.A.; Taroco, E.

    1980-01-01

    The problem of stationary creep is presented as well as its variational formulation, when weak constraints are established, capable of assuring one single solution. A second, so-called elasto-creep problem, is further analysed, together with its variational formulation. It is shown that its stationary solution coincides with that of the stationary creep and the advantages of this formulation with respect to the former one is emphasized. Some numerical applications showing the efficiency of the method propesed are finally presented [pt

  5. Creep analysis of orthotropic shells

    International Nuclear Information System (INIS)

    Mehra, V.K.; Ghosh, A.

    1975-01-01

    A method of creep analysis of orthotropic cylindrical shells subjected to axisymmetric loads has been developed. A general study of creep behaviour of cylindrical shells subjected to a uniform internal pressure has been conducted for a wide range of values of anisotropy coefficients and creep law exponent. Analysis includes determination of stress re-distribution, strain rates, stationary state stresses. Application of reference stress technique has been extended to analysis of shells. (author)

  6. The importance of the strain rate and creep on the stress corrosion cracking mechanisms and models

    International Nuclear Information System (INIS)

    Aly, Omar F.; Mattar Neto, Miguel; Schvartzman, Monica M.A.M.

    2011-01-01

    Stress corrosion cracking is a nuclear, power, petrochemical, and other industries equipment and components (like pressure vessels, nozzles, tubes, accessories) life degradation mode, involving fragile fracture. The stress corrosion cracking failures can produce serious accidents, and incidents which can put on risk the safety, reliability, and efficiency of many plants. These failures are of very complex prediction. The stress corrosion cracking mechanisms are based on three kinds of factors: microstructural, mechanical and environmental. Concerning the mechanical factors, various authors prefer to consider the crack tip strain rate rather than stress, as a decisive factor which contributes to the process: this parameter is directly influenced by the creep strain rate of the material. Based on two KAPL-Knolls Atomic Power Laboratory experimental studies in SSRT (slow strain rate test) and CL (constant load) test, for prediction of primary water stress corrosion cracking in nickel based alloys, it has done a data compilation of the film rupture mechanism parameters, for modeling PWSCC of Alloy 600 and discussed the importance of the strain rate and the creep on the stress corrosion cracking mechanisms and models. As derived from this study, a simple theoretical model is proposed, and it is showed that the crack growth rate estimated with Brazilian tests results with Alloy 600 in SSRT, are according with the KAPL ones and other published literature. (author)

  7. Thermal ratcheting and creep damage

    International Nuclear Information System (INIS)

    Clement, G.; Cousseran, P.; Roche, R.L.

    1983-01-01

    Several proposals have been made to assist adesigners with thermal ratcheting in the creep range, the more known has been made by O'DONNELL and POROWSKY. Unfortunately these methods are not validated by experiments, and they take only inelastic distortion into consideration as creep effects. The aim of the work presented here is to correct these deficiencies - in providing an experimental basis to ratcheting analysis rules in the creep range, - in considering the effect of cyclic straining (like cyclic thermal stresses) on the time to rupture by creep. Experimental tests have been performed on austenitic stainless steel at 650 0 C for the first item. Results of these tests and results available in the open literature have been used to built a practical rule of ratcheting analysis. This rule giving a conservative value of the creep distortion, is based on the concept of effective primary stress which is an amplification of the primary stress really applied. Concerning the second point (time to rupture), it was necessary to obtain real creep rupture and not instability. According to the proposal of Pr LECKIE, tests were performed on specimens made out of copper, and of aluminium alloys at temperatures between 150 0 C and 300 0 C. With such materials creep rupture is obtained without necking. Experimental tests show that cyclic straining reduces the time to creep rupture under load controlled stress. Caution must be given to the designer: cyclic thermal stress can lead to premature creep rupture. (orig./GL)

  8. Creep of high temperature composites

    International Nuclear Information System (INIS)

    Sadananda, K.; Feng, C.R.

    1993-01-01

    High temperature creep deformation of composites is examined. Creep of composites depends on the interplay of many factors. One of the basic issues in the design of the creep resistant composites is the ability to predict their creep behavior from the knowledge of the creep behavior of the individual components. In this report, the existing theoretical models based on continuum mechanics principles are reviewed. These models are evaluated using extensive experimental data on molydisilicide-silicon carbide composites obtained by the authors. The analysis shows that the rule of mixture based on isostrain and isostress provides two limiting bounds wherein all other theoretical predictions fall. For molydisilicide composites, the creep is predominantly governed by the creep of the majority phase, i.e. the matrix with fibers deforming elastically. The role of back stresses both on creep rates and activation energies are shown to be minimum. Kinetics of creep in MoSi 2 is shown to be controlled by the process of dislocation glide with climb involving the diffusion of Mo atoms

  9. Jojoba could stop the desert creep

    Energy Technology Data Exchange (ETDEWEB)

    1982-03-25

    The Sahara desert is estimated to be expanding at a rate of 5km a year. The Sudanese government is experimenting with jojoba in six different regions as the bush has the potential to stop this ''desert creep''. The plant, a native to Mexico, is long known for its resistance to drought and for the versatile liquid wax that can be extracted from its seeds. It is estimated that one hectare of mature plants could produce 3000 kg of oil, currently selling at $50 per litre, and so earn valuable foreign currency.

  10. Creep Crack Initiation and Growth Behavior for Ni-Base Superalloys

    Science.gov (United States)

    Nagumo, Yoshiko; Yokobori, A. Toshimitsu, Jr.; Sugiura, Ryuji; Ozeki, Go; Matsuzaki, Takashi

    The structural components which are used in high temperature gas turbines have various shapes which may cause the notch effect. Moreover, the site of stress concentration might have the heterogeneous microstructural distribution. Therefore, it is necessary to clarify the creep fracture mechanism for these materials in order to predict the life of creep fracture with high degree of accuracy. In this study, the creep crack growth tests were performed using in-situ observational testing machine with microscope to observe the creep damage formation and creep crack growth behavior. The materials used are polycrystalline Ni-base superalloy IN100 and directionally solidified Ni-base superalloy CM247LC which were developed for jet engine turbine blades and gas turbine blades in electric power plants, respectively. The microstructural observation of the test specimens was also conducted using FE-SEM/EBSD. Additionally, the analyses of two-dimensional elastic-plastic creep finite element using designed methods were conducted to understand the effect of microstructural distribution on creep damage formation. The experimental and analytical results showed that it is important to determine the creep crack initiation and early crack growth to predict the life of creep fracture and it is indicated that the highly accurate prediction of creep fracture life could be realized by measuring notch opening displacement proposed as the RNOD characteristic.

  11. Machinery failure analysis and troubleshooting practical machinery management for process plants

    CERN Document Server

    Bloch, Heinz P

    2012-01-01

    Solve the machinery failure problems costing you time and money with this classic, comprehensive guide to analysis and troubleshooting  Provides detailed, complete and accurate information on anticipating risk of component failure and avoiding equipment downtime Includes numerous photographs of failed parts to ensure you are familiar with the visual evidence you need to recognize Covers proven approaches to failure definition and offers failure identification and analysis methods that can be applied to virtually all problem situations Demonstr

  12. Tensile cracks in creeping solids

    International Nuclear Information System (INIS)

    Riedel, H.; Rice, J.R.

    1979-02-01

    The loading parameter determining the stress and strain fields near a crack tip, and thereby the growth of the crack, under creep conditions is discussed. Relevant loading parameters considered are the stress intensity factor K/sub I/, the path-independent integral C*, and the net section stress sigma/sub net/. The material behavior is modelled as elastic-nonlinear viscous where the nonlinear term describes power law creep. At the time t = 0 load is applied to the cracked specimen, and in the first instant the stress distribution is elastic. Subsequently, creep deformation relaxes the initial stress concentration at the crack tip, and creep strains develop rapidly near the crack tip. These processes may be analytically described by self-similar solutions for short times t. Small scale yielding may be defined. In creep problems, this means that elastic strains dominate almost everywhere except in a small creep zone which grows around the crack tip. If crack growth ensues while the creep zone is still small compared with the crack length and the specimen size, the stress intensity factor governs crack growth behavior. If the calculated creep zone becomes larger than the specimen size, the stresses become finally time-independent and the elastic strain rates can be neglected. In this case, the stress field is the same as in the fully-plastic limit of power law hardening plasticity. The loading parameter which determines the near tip fields uniquely is then the path-independent integral C*.K/sub I/ and C* characterize opposite limiting cases. The case applied in a given situation is decided by comparing the creep zone size with the specimen size and the crack length. Besides several methods of estimating the creep zone size, a convenient expression for a characteristic time is derived, which characterizes the transition from small scale yielding to extensive creep of the whole specimen

  13. Experimental investigation of creep behavior of clastic rock in Xiangjiaba Hydropower Project

    Directory of Open Access Journals (Sweden)

    Yu Zhang

    2015-01-01

    Full Text Available There are many fracture zones crossing the dam foundation of the Xiangjiaba Hydropower Project in southwestern China. Clastic rock is the main media of the fracture zone and has poor physical and mechanical properties. In order to investigate the creep behavior of clastic rock, triaxial creep tests were conducted using a rock servo-controlling rheological testing machine. The results show that the creep behavior of clastic rock is significant at a high level of deviatoric stress, and less time-dependent deformation occurs at high confining pressure. Based on the creep test results, the relationship between axial strain and time under different confining pressures was investigated, and the relationship between axial strain rate and deviatoric stress was also discussed. The strain rate increases rapidly, and the rock sample fails eventually under high deviatoric stress. Moreover, the creep failure mechanism under different confining pressures was analyzed. The main failure mechanism of clastic rock is plastic shear, accompanied by a significant compression and ductile dilatancy. On the other hand, with the determined parameters, the Burgers creep model was used to fit the creep curves. The results indicate that the Burgers model can exactly describe the creep behavior of clastic rock in the Xiangjiaba Hydropower Project.

  14. Optimum tungsten content in high strength 9 to 12% chromium containing creep resistant steels

    International Nuclear Information System (INIS)

    Hasegawa, Y.; Muraki, T.; Mimura, H.

    2000-01-01

    Tungsten containing ferritic creep resistant steels are the candidate materials for ultra-super-critical fossil power plant because of their high creep rupture strength. But the strengthening mechanisms by tungsten addition have not yet been completely studied. In this report, creep rupture time and creep strain rate measurement decided the optimum tungsten content in 9 to 12% chromium ferritic steels. The precipitation behavior of Laves phase and the precise discussion of creep strain rate analyses explain the contribution of Laves phase at the lath boundary and the contribution of tungsten in solid solution. P92 contains the optimum amount of tungsten and chromium, 1.8 mass% and 9 mass% respectively judging from the creep rupture strength point of view. (orig.)

  15. Variability of tissue mineral density can determine physiological creep of human vertebral cancellous bone.

    Science.gov (United States)

    Kim, Do-Gyoon; Shertok, Daniel; Ching Tee, Boon; Yeni, Yener N

    2011-06-03

    Creep is a time-dependent viscoelastic deformation observed under a constant prolonged load. It has been indicated that progressive vertebral deformation due to creep may increase the risk of vertebral fracture in the long-term. The objective of this study was to examine the relationships of creep with trabecular architecture and tissue mineral density (TMD) parameters in human vertebral cancellous bone at a physiological static strain level. Architecture and TMD parameters of cancellous bone were analyzed using microcomputerized tomography (micro-CT) in specimens cored out of human vertebrae. Then, creep and residual strains of the specimens were measured after a two-hour physiological compressive constant static loading and unloading cycle. Creep developed (3877 ± 2158 με) resulting in substantial levels of non-recoverable post-creep residual strain (1797 ± 1391 με). A strong positive linear correlation was found between creep and residual strain (r = 0.94, p creep rate. The TMD variability (GL(COV)) was the strongest correlate of creep rate (r = 0.79, p < 0.001). This result suggests that TMD variability may be a useful parameter for estimating the long-term deformation of a whole vertebral body. The results further suggest that the changes in TMD variability resulting from bone remodeling are of importance and may provide an insight into the understanding of the mechanisms underlying progressive failure of vertebral bodies and development of a clinical fracture. Copyright © 2011 Elsevier Ltd. All rights reserved.

  16. Creep Measurement Video Extensometer

    Science.gov (United States)

    Jaster, Mark; Vickerman, Mary; Padula, Santo, II; Juhas, John

    2011-01-01

    Understanding material behavior under load is critical to the efficient and accurate design of advanced aircraft and spacecraft. Technologies such as the one disclosed here allow accurate creep measurements to be taken automatically, reducing error. The goal was to develop a non-contact, automated system capable of capturing images that could subsequently be processed to obtain the strain characteristics of these materials during deformation, while maintaining adequate resolution to capture the true deformation response of the material. The measurement system comprises a high-resolution digital camera, computer, and software that work collectively to interpret the image.

  17. Comparative study on the high-temperature tensile and creep properties of Alloy 617 base and weld metals

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Hong, Sung Deok; Kim, Yong Wan; Kim, Seon Jin; Park, Jae Young; Ekaputra, I. M. W.

    2013-01-01

    This paper presents a comparative investigation on the high-temperature tensile and creep properties of Alloy 617 base metal (BM) and weld metal (WM) fabricated by a gas tungsten arc weld process. The WM had higher yield strength and lower ultimate tensile strength than the BM does; however, its elongation was significantly lower than that of the BM. The creep curve of the BM and WM was somewhat different from that of typical heat-resistance steel, and did not show a textbook creep. The WM exhibited a longer creep rupture life, lower creep rate, and lower rupture ductility than the BM. However, as the creep rupture time reached approximately 36,800 h, the creep life of the WM was expected to be almost similar to that of the BM; and after 36,800 h, its creep life was expected to be worse than the BM. Loner creep tests is needed to investigate the long-term creep life of the WM. The creep failure mode of the BM and WM was obviously an intergranular cracking of the cavity formation and growth mechanisms, although it was more evident in the WM. The BM had a more ductile fracture surface than the WM

  18. Vortex pinning and creep experiments

    International Nuclear Information System (INIS)

    Kes, P.H.

    1991-01-01

    A brief review of basic flux-pinning and flux-creep ingredients and a selection of experimental results on high-temperature-superconductivity compounds is presented. Emphasis is put on recent results and on those properties which are central to the emerging understanding of the flux-pinning and flux-creep mechanisms of these fascinating materials

  19. Nanogranular origin of concrete creep.

    Science.gov (United States)

    Vandamme, Matthieu; Ulm, Franz-Josef

    2009-06-30

    Concrete, the solid that forms at room temperature from mixing Portland cement with water, sand, and aggregates, suffers from time-dependent deformation under load. This creep occurs at a rate that deteriorates the durability and truncates the lifespan of concrete structures. However, despite decades of research, the origin of concrete creep remains unknown. Here, we measure the in situ creep behavior of calcium-silicate-hydrates (C-S-H), the nano-meter sized particles that form the fundamental building block of Portland cement concrete. We show that C-S-H exhibits a logarithmic creep that depends only on the packing of 3 structurally distinct but compositionally similar C-S-H forms: low density, high density, ultra-high density. We demonstrate that the creep rate ( approximately 1/t) is likely due to the rearrangement of nanoscale particles around limit packing densities following the free-volume dynamics theory of granular physics. These findings could lead to a new basis for nanoengineering concrete materials and structures with minimal creep rates monitored by packing density distributions of nanoscale particles, and predicted by nanoscale creep measurements in some minute time, which are as exact as macroscopic creep tests carried out over years.

  20. Creep buckling analysis of shells

    International Nuclear Information System (INIS)

    Stone, C.M.; Nickell, R.E.

    1977-01-01

    The current study was conducted in an effort to determine the degree of conservatism or lack of conservatism in current ASME design rules concerning time-dependent (creep) buckling. In the course of this investigation, certain observations were made concerning the numerical solution of creep buckling problems. It was demonstrated that a nonlinear finite element code could be used to solve the time-dependent buckling problem. A direct method of solution was presented which proved to be computationally efficient and provided answers which agreed very well with available analytical solutions. It was observed that the calculated buckling times could vary widely for small errors in computed displacements. The presence of high creep strain rates contributed to the prediction of early buckling times when calculated during the primary creep stage. The predicted time estimates were found to increase with time until the secondary stage was reached and the estimates approached the critical times predicted without primary creep. It can be concluded, therefore, that for most nuclear piping components, whose primary creep stage is small compared to the secondary stage, the effect of primary creep is negligible and can be omitted from the calculations. In an evaluation of the past and current ASME design rules for time-dependent, load controlled buckling, it was concluded that current use of design load safety factors is not equivalent to a safety factor of ten on service life for low creep exponents

  1. Microstructural degradation mechanisms during creep in strength enhanced high Cr ferritic steels and their evaluation by hardness measurement

    International Nuclear Information System (INIS)

    Ghassemi Armaki, Hassan; Chen, Ruiping; Kano, Satoshi; Maruyama, Kouichi; Hasegawa, Yasushi; Igarashi, Masaaki

    2011-01-01

    Graphical abstract: Effect of static recovery on the acceleration of subgrain coarsening during creep plastic deformation. Display Omitted Highlights: → Short-term 'H' and long-term 'L' creep regions have different creep characteristics. → Strain-induced recovery of subgrains proceeds in the both creep regions 'H' and 'L'. → In region ''L', two additional degradation mechanisms accelerate creep failure. → Thermal coarsening of precipitates and subgrains appear during long-term creep ''L'. → In region 'L', strain-induced coarsening of precipitates accelerates creep failure. - Abstract: There are two creep regions with different creep characteristics: short-term creep region 'H', where precipitates and subgrains are thermally stable, and long-term creep region 'L', where thermal coarsening of precipitates and subgrains appear. In region 'H', the normalized subgrain size (λ-λ 0 )/(λ * -λ 0 ) has a linear relation with creep strain and its slope is 10ε -1 . But, region L is the time range in which the static recovery and the strain-induced recovery progress simultaneously. In this region, the static recovery accelerates the strain-induced recovery, and subgrain size is larger than that line which neglects the contribution of the static recovery. In region 'L', the Δλ/Δλ * -strain present a linear relation with a slope 35ε -1 . There is a linear relation between hardness and subgrain size. Hardness drop, H 0 - H, as a function of Larson-Miller parameter can be a good measure method for assessment of hardness drop and consequently degradation of microstructure. Hardness drop shows an identical slope in creep region 'H', whereas hardness drop due to thermal aging and creep in region 'L' show together a similar slope. In region 'H', degradation of microstructure is mainly due to recovery of subgrains controlled by creep plastic deformation, and precipitates do not have a major role. However, in creep region 'L', there are three degradation mechanisms

  2. Application of the failure modes and effects analysis technique to the emergency cooling system of an experimental nuclear power plant

    International Nuclear Information System (INIS)

    Conceicao Junior, Osmar; Silva, Antonio Teixeira e

    2009-01-01

    This study consists on the application of the failure modes and effects analysis (FMEA), a hazard identification and a risk assessment technique, to the emergency cooling system (ECS), of an experimental nuclear power plant. The choice of this technique was due to its detailed analysis of each component of the system, enabling the identification of all possible ways of failure and its related consequences (in order of importance), allowing the designer to improve the system, maximizing its security and reliability. Through the application of this methodology, it could be observed that the ECS is an intrinsically safe system, in spite of the modifications proposed. (author)

  3. Statistical investigations of the failure behaviour of components in the AVR-experimental nuclear power plant. Vol. 1

    International Nuclear Information System (INIS)

    Hennings, W.

    1989-08-01

    From operational reports of the years 1970 to 1984, failure rates of valves in gas circuits of the AVR experimental power plant were determined. Also, potential influences of environmental and operational conditions were investigated. The resulting failure rates are for manual valves app. 0,1.10 -6 /h, for pneumatic valves between 3 and 9.10 -6 /h, for solenoid valves between 1,5 and 4.10 -6 /h and for control valves between 12 and 41.10 -6 /h. (orig.) [de

  4. Influence of shock waves as a result of assumed vessel failure on parts of the plant relevant to safety

    International Nuclear Information System (INIS)

    Danisch, R.; Graubner, U.

    1981-01-01

    The shock wave induced rupture is of subordinate importance for the laying out of the parts of the plant relevant to safety. It is covered by the precautions for maximum potential earthquakes, aircraft crashes and chemical explosions. The failure of vessels in the power house (WAZUe, SPWB) as the result of a maximum potential earthquake is extremely improbable. If a combination of the stresses resulting from maximum potential earthquakes with the hypothetical stresses resulting from vessel failure is undertaken, it can be seen that the total stresses are only increased by a minimal amount, due to the quadratic averaging of less than 3%. (orig./DG) [de

  5. Creep Deformation and Fracture Processes in OF and OFP Copper

    International Nuclear Information System (INIS)

    Bowyer, William H.

    2004-10-01

    The literature on creep processes in many materials, including copper, has been thoroughly reviewed and complemented by Ashby and co-workers. They have provided physical models which describe the deformation and fracture processes with good qualitative and quantitative agreement with experimental data for many cases. A description of the deformation and fracture models is provided and the relevant equations are included in the appendices. Published data from the canister development programme has been compared with the predictions from the models. The purpose was to improve our understanding of (1) a reported benefit to creep performance which arises from additions of 50 ppm phosphorus to oxygen free (OF) copper, and (2) an observed transition from brittle to ductile failure in OF copper. The models adequately describe the general variations in the observed creep behaviour of the experimental materials. Steady state creep rates for OF copper are observed to be up to one order of magnitude higher than the model predicts for pure copper across a wide range of temperatures and stresses in the power law and power law breakdown regimes. For OF copper with 50ppm of phosphorus added (OFP copper), observed steady state creep rates in the power law breakdown regime are up to one order of magnitude lower than the model predicts for pure copper. Creep lives in the experimental OFP material are also higher than creep lives for OF material under similar conditions. The lower creep deformation rates and the longer creep lives of OFP material are attributed the known effects of phosphorus on recovery in copper. The model predicts that the same mechanism will improve creep lives under repository conditions. It is suggested that the factor of improvement under repository conditions will be less than the factor which is observed in the power law breakdown regime. Predicted creep lives, based on measured steady state creep rates and stress exponents ('n' values) are in good agreement

  6. Piping failures in United States nuclear power plants 1961-1995

    International Nuclear Information System (INIS)

    Bush, S.H.; Do, M.J.; Slavich, A.L.; Chockie, A.D.

    1996-01-01

    Over 1500 reported piping failures were identified and summarized based on an extensive review of tens of thousands of event reports that have been submitted to the US regulatory agencies over the last 35 years. The data base contains only piping failures; failures in vessels, pumps, valves and steam generators or any cracks that were not through-wall are not included. It was observed that there has been a marked decrease in the number of failures after 1983 for almost all sizes of pipes. This is likely due to the changes in the reporting requirements at that time and the corrective actions taken by utilities to minimize fatigue failures of small lines and IGSCC in BWRs. One failure mechanism that continues to occur is erosion-corrosion, which accounts for most of the ruptures reported and probably is responsible for the absence of downward trends in ruptures. Fatigue-vibration is also a significant contributor to piping failures. However, most of such events occur in lines approx. one inch or less in diameter. Together, erosion-corrosion and fatigue-vibration account for over 43 per cent of the failures. The overwhelming majority of failures have been leaks, over half the failures occurred in pipes with a diameter of one inch or less. Included in the report is a listing of the number of welds in various systems in LWRs

  7. Thermal ratcheting and creep damage

    International Nuclear Information System (INIS)

    Clement, G.; Cousseran, P.; Roche, R.L.

    1983-08-01

    Creep is a cause of deformation; it may also result in rupture in time. Although LMFBR structures are not heavily loaded, they are subjected to large thermal transients. Can structure lifetime be shortened by such transients. Several proposals have been made to assist adesigners with thermal ratcheting in the creep range. Unfortunately these methods are not validated by experiments, and they take only inelastic distorsion into consideration as creep effects. The aim of the work presented here is to correct these deficiencies in providing an experimental basis to ratcheting analysis rules in the creep range, and in considering the effect of cyclic straining (like cyclic thermal stresses) on the time to rupture by creep. Experimental tests have been performed on austenitic stainless steel at 650 0 C for the first item. Results of these tests and results available in the open literature have been used to built a practical rule of ratcheting analysis. This rule giving a conservative value of the creep distortion, is based on the concept of effective primary stress which is an amplification of the primary stress really applied. Concerning the second point (time to rupture), it was necessary to obtain real creep rupture and not instability. According to the proposal of Pr LECKIE, tests were performed on specimen made out of copper, and of aluminium alloys at temperatures between 150 0 C and 300 0 C. With such materials creep rupture is obtained without necking. Experimental tests show that cyclic straining reduces the time to creep rupture under load controlled stress. Caution must be given to the designer: cyclic thermal stress can lead to premature creep rupture

  8. Reassembling Surveillance Creep

    DEFF Research Database (Denmark)

    Bøge, Ask Risom; Lauritsen, Peter

    2017-01-01

    We live in societies in which surveillance technologies are constantly introduced, are transformed, and spread to new practices for new purposes. How and why does this happen? In other words, why does surveillance “creep”? This question has received little attention either in theoretical developm......We live in societies in which surveillance technologies are constantly introduced, are transformed, and spread to new practices for new purposes. How and why does this happen? In other words, why does surveillance “creep”? This question has received little attention either in theoretical...... development or in empirical analyses. Accordingly, this article contributes to this special issue on the usefulness of Actor-Network Theory (ANT) by suggesting that ANT can advance our understanding of ‘surveillance creep’. Based on ANT’s model of translation and a historical study of the Danish DNA database......, we argue that surveillance creep involves reassembling the relations in surveillance networks between heterogeneous actors such as the watchers, the watched, laws, and technologies. Second, surveillance creeps only when these heterogeneous actors are adequately interested and aligned. However...

  9. Standard test method for creep-fatigue testing

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers the determination of mechanical properties pertaining to creep-fatigue deformation or crack formation in nominally homogeneous materials, or both by the use of test specimens subjected to uniaxial forces under isothermal conditions. It concerns fatigue testing at strain rates or with cycles involving sufficiently long hold times to be responsible for the cyclic deformation response and cycles to crack formation to be affected by creep (and oxidation). It is intended as a test method for fatigue testing performed in support of such activities as materials research and development, mechanical design, process and quality control, product performance, and failure analysis. The cyclic conditions responsible for creep-fatigue deformation and cracking vary with material and with temperature for a given material. 1.2 The use of this test method is limited to specimens and does not cover testing of full-scale components, structures, or consumer products. 1.3 This test method is primarily ...

  10. Creep damage of 12% CrMoV weldments

    International Nuclear Information System (INIS)

    Kussmaul, K.; Maile, K.; Theofel, H.

    1989-01-01

    Creep tests were performed to determine the creep behaviour of similar welded joints of 12% CrMoV-steels which had been made using various heat inputs. The specimens were taken transverse to the seam. The transition from the coarse-grained to the fine-grained area of the heat affected zone (HAZ) proved to be the location of failure after longer rupture times. All tested specimens lie in the +-20% scatterband of the material standard DIN 17175. Creep rupture was initiated by the nucleation and growth of cavities. The appearance of the damage zone near the fracture face depends on testing conditions and heat input. The nucleation of cavities can be detected at an early stage of lifetime

  11. Creep Burst Testing of a Woven Inflatable Module

    Science.gov (United States)

    Selig, Molly M.; Valle, Gerard D.; James, George H.; Oliveras, Ovidio M.; Jones, Thomas C.; Doggett, William R.

    2015-01-01

    A woven Vectran inflatable module 88 inches in diameter and 10 feet long was tested at the NASA Johnson Space Center until failure from creep. The module was pressurized pneumatically to an internal pressure of 145 psig, and was held at pressure until burst. The external environment remained at standard atmospheric temperature and pressure. The module burst occurred after 49 minutes at the target pressure. The test article pressure and temperature were monitored, and video footage of the burst was captured at 60 FPS. Photogrammetry was used to obtain strain measurements of some of the webbing. Accelerometers on the test article measured the dynamic response. This paper discusses the test article, test setup, predictions, observations, photogrammetry technique and strain results, structural dynamics methods and quick-look results, and a comparison of the module level creep behavior to the strap level creep behavior.

  12. Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel and NDE assessment

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Kim, Jong Bum; Kim, Seok Hun; Yoo, Bong

    2001-03-01

    Creep fatigue crack growth contributes to the failure of FRB reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety and setup the in-service inspection strategy. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS crack, NDE short crack and shallow long crack) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. The total crack growth of the largest PTS crack is very small after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life. The crack depth growth of the shallow long crack is faster than that of the NDE short crack. In the ISI of the large-scale FBR reactor vessel, the ultrasonic inspection is beneficial to detect the shallow circumferential cracks.

  13. Above- and Belowground Trophic Interactions on Creeping Thistle (Cirsium arvense) in High- and Low-Diversity Plant Communities: Potential for Biotic Resistance?

    NARCIS (Netherlands)

    Bezemer, T.M.; Graça, O.; Rousseau, P.; Putten, van der W.H.

    2004-01-01

    The capacity of local communities to control introduced plants is called biotic resistance. Biotic resistance has been almost exclusively tested for plant competition and above-ground herbivores and pathogens, while neglecting root herbivores and soil pathogens. Here, we present biotic resistance by

  14. Above- and Belowground Trophic Interactions on Creeping Thistle (Cirsium arvense) in High- and Low-Diversity Plant Communities: Potential for Biotic Resistance?

    NARCIS (Netherlands)

    Bezemer, T.M.; Graça, O.; Rousseau, P.; Van der Putten, W.H.

    2004-01-01

    The capacity of local communities to control introduced plants is called biotic resistance. Biotic resistance has been almost exclusively tested for plant competition and aboveground herbivores and pathogens, while neglecting root herbivores and soil pathogens. Here, we present biotic resistance by

  15. Creep Behavior and Durability of Cracked CMC

    Science.gov (United States)

    Bhatt, R. T.; Fox, Dennis; Smith, Craig

    2015-01-01

    To understand failure mechanisms and durability of cracked Ceramic matrix composites (CMCs), Melt Infiltration (MI) SiCSiC composites with Sylramic-iBN fibers and full Chemical vapour infiltration SiCSiC composites with Sylramic-ion bombarded BN (iBN) and Hi-Nicalon -S fibers were pre-cracked between 150 to 200 megapascal and then creep and Sustained Peak Low Cycle Fatigue (SPLCF) tested at 13150 C at stress levels from 35 to 103 megapascal for up to 200 hours under furnace and burner rig conditions. In addition creep testing was also conducted on pre-cracked full Chemical vapour infiltration SiCSiC composites at 14500 C between 35 and 103 megapascal for up to 200 hours under furnace conditions. If the specimens survived the 200 hour durability tests, then they were tensile tested at room temperature to determine their residual tensile properties. The failed specimens were examined by Scanning electron microscope (SEM) to determine the failure modes and mechanisms. The influence of crack healing matrix, fiber types, crack density, testing modes and interface oxidation on durability of cracked Ceramic matrix composites (CMCs) will be discussed.

  16. Mechanics of Failure Mechanisms in Structures

    CERN Document Server

    Carlson, R L; Craig, J I

    2012-01-01

    This book focuses on the mechanisms and underlying mechanics of failure in various classes of materials such as metallic, ceramic, polymeric, composite and bio-material.  Topics include tensile and compressive fracture, crack initiation and growth, fatigue and creep rupture in metallic materials, matrix cracking and delamination and environmental degradation in polymeric composites, failure of bio-materials such as prosthetic heart valves and prosthetic hip joints, failure of ceramics and ceramic matrix composites, failure of metallic matrix composites, static and dynamic buckling failure, dynamic excitations and creep buckling failure in structural systems. Chapters are devoted to failure mechanisms that are characteristic of each of the materials.  The work also provides the basic elements of fracture mechanics and studies in detail several niche topics such as the effects of toughness gradients, variable amplitude loading effects in fatigue, small fatigue cracks, and creep induced brittleness. Furthe...

  17. Standard test method for creep-fatigue crack growth testing

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method covers the determination of creep-fatigue crack growth properties of nominally homogeneous materials by use of pre-cracked compact type, C(T), test specimens subjected to uniaxial cyclic forces. It concerns fatigue cycling with sufficiently long loading/unloading rates or hold-times, or both, to cause creep deformation at the crack tip and the creep deformation be responsible for enhanced crack growth per loading cycle. It is intended as a guide for creep-fatigue testing performed in support of such activities as materials research and development, mechanical design, process and quality control, product performance, and failure analysis. Therefore, this method requires testing of at least two specimens that yield overlapping crack growth rate data. The cyclic conditions responsible for creep-fatigue deformation and enhanced crack growth vary with material and with temperature for a given material. The effects of environment such as time-dependent oxidation in enhancing the crack growth ra...

  18. An experimental and theoretical investigation of creep buckling

    International Nuclear Information System (INIS)

    Ohya, H.

    1977-01-01

    Creep buckling is one of the failure modes which must be taken into consideration for the design of structures exposed to elevated temperatures. And, rules are provided in ASME Boiler and Pressure Vessel Code Case 1592 to prevent the creep buckling. However, methods of analysis are not provided in Code Case, and selecting the methods of analysis is left to owners and manufacturers. The purpose of the present paper is to investigate creep buckling phenomena and the methods of analysis. Creep buckling experiments were performed on aluminum alloy 2024-T4 cylindrical shells having radius to thickness ratios of 16, 25, 50 and 80, in single, double and triple step axial compression at 250 0 C. It was observed that buckling occurred at one of the edges and the buckling mode depended on the radius to thickness ratio and also on the applied stress level. Thicker cylinders buckled in axisymmetric mode. Thinner ones under higher applied stress levels buckled in the asymmetric mode, whereas under lower applied stress levels they buckled in the axisymmetric mode. Creep buckling times were obtained from end shortening record of the cylinders. Experimental results were compared with theoretical values obtained by two methods. (Auth.)

  19. Negative creep in nickel base superalloys

    DEFF Research Database (Denmark)

    Dahl, Kristian Vinter; Hald, John

    2004-01-01

    Negative creep describes the time dependent contraction of a material as opposed to the elongation seen for a material experiencing normal creep behavior. Negative creep occurs because of solid state transformations that results in lattice contractions. For most applications negative creep will h...

  20. Mechanisms of transient radiation-induced creep

    International Nuclear Information System (INIS)

    Pyatiletov, Yu.S.

    1981-01-01

    Radiation-induced creep at the transient stage is investigated for metals. The situation, when several possible creep mechanisms operate simultaneously is studied. Among them revealed are those which give the main contribution and determine thereby the creep behaviour. The time dependence of creep rate and its relation to the smelling rate is obtained. The results satisfactorily agree with the available experimental data [ru

  1. Modeling Creep Processes in Aging Polymers

    Science.gov (United States)

    Olali, N. V.; Voitovich, L. V.; Zazimko, N. N.; Malezhik, M. P.

    2016-03-01

    The photoelastic method is generalized to creep in hereditary aging materials. Optical-creep curves and mechanical-creep or optical-relaxation curves are used to interpret fringe patterns. For materials with constant Poisson's ratio, it is sufficient to use mechanical- or optical-creep curves for this purpose

  2. The plant-specific impact of different pressurization rates in the probabilistic estimation of containment failure modes

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Yang, Joon Eon; Ha, Jae Joo

    2003-01-01

    The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities

  3. A multivariate statistical methodology for detection of degradation and failure trends using nuclear power plant operational data

    International Nuclear Information System (INIS)

    Samanta, P.K.; Teichmann, T.

    1990-01-01

    In this paper, a multivariate statistical method is presented and demonstrated as a means for analyzing nuclear power plant transients (or events) and safety system performance for detection of malfunctions and degradations within the course of the event based on operational data. The study provides the methodology and illustrative examples based on data gathered from simulation of nuclear power plant transients (due to lack of easily accessible operational data). Such an approach, once fully developed, can be used to detect failure trends and patterns and so can lead to prevention of conditions with serious safety implications

  4. Lattice continuum and diffusional creep.

    Science.gov (United States)

    Mesarovic, Sinisa Dj

    2016-04-01

    Diffusional creep is characterized by growth/disappearance of lattice planes at the crystal boundaries that serve as sources/sinks of vacancies, and by diffusion of vacancies. The lattice continuum theory developed here represents a natural and intuitive framework for the analysis of diffusion in crystals and lattice growth/loss at the boundaries. The formulation includes the definition of the Lagrangian reference configuration for the newly created lattice, the transport theorem and the definition of the creep rate tensor for a polycrystal as a piecewise uniform, discontinuous field. The values associated with each crystalline grain are related to the normal diffusional flux at grain boundaries. The governing equations for Nabarro-Herring creep are derived with coupled diffusion and elasticity with compositional eigenstrain. Both, bulk diffusional dissipation and boundary dissipation accompanying vacancy nucleation and absorption, are considered, but the latter is found to be negligible. For periodic arrangements of grains, diffusion formally decouples from elasticity but at the cost of a complicated boundary condition. The equilibrium of deviatorically stressed polycrystals is impossible without inclusion of interface energies. The secondary creep rate estimates correspond to the standard Nabarro-Herring model, and the volumetric creep is small. The initial (primary) creep rate is estimated to be much larger than the secondary creep rate.

  5. Multiaxial creep-fatigue rules

    International Nuclear Information System (INIS)

    Spindler, M.W.; Hales, R.; Ainsworth, R.A.

    1997-01-01

    Within the UK, a comprehensive procedure, called R5, is used to assess the high temperature response of structures. One part of R5 deals with creep-fatigue initiation, and in this paper we describe developments in this part of R5 to cover multiaxial stress states. To assess creep-fatigue, damage is written as the linear sum of fatigue and creep components. Fatigue is assessed using Miner's law with the total endurance split into initiation and growth cycles. Initiation is assessed by entering the curve of initiation cycles vs strain range using a Tresca equivalent strain range. Growth is assessed by entering the curve of growth cycles vs strain range using a Rankine equivalent strain range. The number of allowable cycles is obtained by summing the initiation and growth cycles. In this way the problem of defining an equivalent strain range applicable over a range of endurance is avoided. Creep damage is calculated using ductility exhaustion methods. In this paper we address two aspects; first, the nature of stress relaxation and, hence, accumulated creep strain in multiaxial stress fields; secondly, the effect of multiaxial stress on creep ductility. The effect of multiaxial stress state on creep ductility has been examined using experimental data and mechanistic models. Good agreement is demonstrated between an empirical description of test data and a cavity growth model, provided a simple nucleation criterion is included. A simple scaling factor is applied to uniaxial creep ductility, defined as a function of stress state. The factor is independent of the cavity growth mechanisms and yields a value of equivalent strain which can be conveniently used in determining creep damage by ductility exhaustion. (author). 14 refs, 4 figs

  6. Multiaxial creep of fine grained 0.5Cr-0.5Mo-0.25V and coarse grained 1Cr-0.5Mo steels

    International Nuclear Information System (INIS)

    Browne, R.J.; Flewitt, P.E.J.; Lonsdale, D.

    1991-01-01

    To explore the multiaxial creep response of materials used for electrical power generating plant, two steels, a fine grained 0.5Cr-0.5Mo-0.25V steel in a normalised and tempered condition with high creep ductility and a coarse grained 1Cr-0.5Mo steel in a quenched and tempered condition with low uniaxial creep ductility, have been selected. A range of multiaxial stress testing techniques which span the stress states that would allow identification of any technique dependent variables has been used. The deformation and failure of the normalised and tempered 0.5Cr-0.5Mo-0.25V steel for a range of multiaxial test techniques and, therefore, stress states may be described by an equivalent stress criterion. The results from the multiaxial tests carried out on the fully bainitic 1Cr-0.5Mo steel show that the multiaxial stress rupture criterion (MSRC) varies with stress state; at high triaxiality (notch), it is controlled by the maximum principal stress, whereas at low triaxiality (shear) it is dependent on both maximum principal stress and equivalent stress. Furthermore, a simple description of stress state based on maximum principal and equivalent stress does not define this uniquely, since the MSRC derived from uniaxial and torsion testing does not describe the failure of notch, tube, or double shear tests. (author)

  7. Development of inelastic design method for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Takahashi, Yukio; Take, Kohji; Kaguchi, Hitoshi; Fukuda, Yoshio; Uno, Tetsuro.

    1991-01-01

    Effective utilization of inelastic analysis in structural design assessment is expected to play an important role for avoiding too conservative design of liquid metal reactor plants. Studies have been conducted by the authors to develop a guideline for application of detailed inelastic analysis in design assessment. Both fundamental material characteristics tests and structural failure tests were conducted. Fundamental investigations were made on inelastic analysis method and creep-fatigue life prediction method based on the results of material characteristics tests. It was demonstrated through structural failure tests that the design method constructed based on these fundamental investigations can predict failure lives in structures subjected to cyclic thermal loadings with sufficient accuracy. (author)

  8. Creep in electronic ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Routbort, J. L.; Goretta, K. C.; Arellano-Lopez, A. R.

    2000-04-27

    High-temperature creep measurements combined with microstructural investigations can be used to elucidate deformation mechanisms that can be related to the diffusion kinetics and defect chemistry of the minority species. This paper will review the theoretical basis for this correlation and illustrate it with examples from some important electronic ceramics having a perovskite structure. Recent results on BaTiO{sub 3}, (La{sub 1{minus}x}Sr){sub 1{minus}y}MnO{sub 3+{delta}}, YBa{sub 2}Cu{sub 3}O{sub x}, Bi{sub 2}Sr{sub 2}CaCu{sub 2}O{sub x}, (Bi,Pb){sub 2}Sr{sub 2}Ca{sub 2}Cu{sub 3}O{sub x} and Sr(Fe,Co){sub 1.5}O{sub x} will be presented.

  9. Creep strength and microstructural evolution of 9-12% Cr heat resistant steels during creep exposure at 600 C and 650 C

    Energy Technology Data Exchange (ETDEWEB)

    Mendez Martin, Francisca [Graz Univ. of Technology (Austria). Inst. for Materials Science and Welding; Panait, Clara Gabriela [MINES ParisTech, UMR CNRS, Evry (France). Centre des Materiaux; V et M France CEV, Aulnoye-Aymeries (France); Bendick, Walter [Salzgitter Mannesmann Forschung GmbH (SZMF), Duisburg (DE)] (and others)

    2010-07-01

    9-12% Cr heat resistant steels are used for applications at high temperatures and pressures in steam power plants. 12% Cr steels show higher creep strength and higher corrosion resistance compared to 9% Cr steels for short term creep exposure. However, the higher creep strength of 12 %Cr steels drops increasingly after 10,000-20,000 h of creep. This is probably due to a microstructural instability such as the precipitation of new phases (e.g. Laves phases and Z-phases), the growth of the precipitates and the recovery of the matrix. 9% Cr and 12% Cr tempered martensitic steels that have been creep tested for times up to 50,000 h at 600 C and 650 C were investigated using Transmission Electron Microscopy (TEM) on extractive replicas and thin foils together with Backscatter Scanning Electron Microscopy (BSE-SEM) to better understand the different creep behaviour of the two different steels. A significant precipitation of Laves phase and low amounts of Z-phase was observed in the 9% Cr steels after long-term creep exposure. The size distribution of Laves phases was measured by image analysis of SEM-BSE images. In the 12% Cr steel two new phases were identified, Laves phase and Z-phase after almost 30,000 h of creep test. The quantification of the different precipitated phases was studied. (orig.)

  10. Influence of flowing sodium on creep deformation and rupture behaviour of 316L(N) austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Ravi, S., E-mail: sravi@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Laha, K.; Mathew, M.D. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Vijayaraghavan, S.; Shanmugavel, M.; Rajan, K.K. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2012-08-15

    The influence of flowing sodium on creep deformation and rupture behaviour of AISI 316L(N) austenitic stainless steel has been investigated at 873 K over a stress range of 235-305 MPa. The results were compared with those obtained from testing in air environment. The steady state creep rates of the material were not influenced appreciably by the testing environments. The time to onset of tertiary stage of creep deformation was delayed in sodium environment. The creep-rupture lives of the material increased in sodium environment, which became more pronounced at lower applied stresses. The increase in rupture life of the material in flowing sodium was accompanied by an increase in rupture ductility. The creep damage on specimen surface as well as inside the specimen was less in specimen tested in sodium. SEM fractographic investigation revealed predominantly transgranular dimple failure for the specimen tested in sodium, whereas predominantly intergranular creep failure was observed in the air tested specimens. Almost no oxidation was observed in the specimens creep tested in the sodium environment. Absence of oxidation and less creep damage cavitation extended the secondary state in liquid sodium tests and lead to increase in creep rupture life and ductility of the material as compared to in air.

  11. Numerical description of creep of highly creep resistant alloys

    International Nuclear Information System (INIS)

    Preussler, T.

    1991-01-01

    Fatigue tests have been performed with a series of highly creep resistant materials for gas turbines and related applications for gaining better creep data up to long-term behaviour. The investigations were performed with selected individual materials in the area of the main applications down to strains and stresses relevant to design, and have attained trial durations of 25000 to 60000 h. In continuing former research, creep equations for a selection of characterizing individual materials have been improved and partly newly developed on the basis of a differentiated evaluation. Concerning the single materials, there are: one melt each of the materials IN-738 LC, IN-939, IN-100, FSX-414 and Inconel 617. The applied differentiated evaluation is based on the elastoplastical behaviour from the hot-drawing test, the creep behaviour from the non interrupted or the interrupted fatigue test, and the contraction behaviour from the annealing test. The creep equations developed describe the high temperature deformation behaviour taking into account primary, secondary and partly the tertiary creep dependent of temperature, stress and time. These equations are valid for the whole application area of the respective material. (orig./MM) [de

  12. Effect of inclusion content on the creep rupture properties of type 17Cr-8Ni-2Mo weld metals

    International Nuclear Information System (INIS)

    Senior, B.A.

    1988-01-01

    It has been known for some time that austenitic weldments exhibit low and variable creep rupture properties, but many of the factors controlling these properties are not well understood. In this investigation, two welds (Type 316) with similar compositions and fabricated using the same welding parameters, but with different electrode coatings, have been examined after creep testing. The results indicate that the inclusion and silicon content of type 316 welds can strongly influence their creep rupture properties, a high inclusion density being associated with poor creep rupture properties, and a low silicon content with a higher creep rate. This has been explained with reference to the micromechanism of creep failure operating in these welds. (author)

  13. Evaluation of nuclear power plant component failure probability and core damage probability using simplified PSA model

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2000-01-01

    It is anticipated that the change of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components may cause the change of component failure probability and result in the change of core damage probability. It is also anticipated that the change is different depending on the initiating event frequency or the component types. This study assessed the change of core damage probability using simplified PSA model capable of calculating core damage probability in a short time period, which is developed by the US NRC to process accident sequence precursors, when various component's failure probability is changed between 0 and 1, or Japanese or American initiating event frequency data are used. As a result of the analysis, (1) It was clarified that frequency of surveillance test, preventive maintenance or parts replacement of motor driven pumps (high pressure injection pumps, residual heat removal pumps, auxiliary feedwater pumps) should be carefully changed, since the core damage probability's change is large, when the base failure probability changes toward increasing direction. (2) Core damage probability change is insensitive to surveillance test frequency change, since the core damage probability change is small, when motor operated valves and turbine driven auxiliary feed water pump failure probability changes around one figure. (3) Core damage probability change is small, when Japanese failure probability data are applied to emergency diesel generator, even if failure probability changes one figure from the base value. On the other hand, when American failure probability data is applied, core damage probability increase is large, even if failure probability changes toward increasing direction. Therefore, when Japanese failure probability data is applied, core damage probability change is insensitive to surveillance tests frequency change etc. (author)

  14. Service life prediction. Development of models for predicting the service life of power plant components subject to thermomechanical creep fatigue; Lebensdauervorhersage. Entwicklung von Modellen zur Lebensdauervorhersage von Kraftwerksbauteilen unter thermisch-mechanischer Kriechermuedungsbeanspruchung. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Cui, L.; Scholz, A. [Technische Univ. Darmstadt (Germany). Institut fuer Werkstoffkunde; Hartrott, P. von; Schlesinger, M. [Fraunhofer-Institut fuer Werkstoffmechanik (IWM), Freiburg im Breisgau (Germany)

    2009-07-01

    Extensive use is made of massive components of heat resistant and highly heat resistant materials in installations of the power and heating industry. These components are exposed to varying thermomechanical stress as a result of ramping-up and down processes. In this research project two computer-assisted methods of predicting service life until crack initiation were extended to include cases of thermomechanical multi-axis stress conducive to creep fatigue and of superposition of high-cycle stress on power plant components. Investigations were limited to rotor steel of type X12CrMoWVNbN10-1-1. Complex thermomechanical multi-axis experiments were performed on round, notched and cruciform test specimens of close-to-life dimensions in order to demonstrate by experiment the validity of these models. The results of these calculations showed an acceptable degree of agreement between experiment and simulation for both models. Calculations on earlier TMF experiments performed at IfW on hollow specimens of 1%CrMoNiV showed good predictability for both the SARA and the ThoMat programme. Calculations on experiments performed at MPA Stuttgart on model bodies consisting of the same 1%CrMoNiV showed a predictability of acceptable variability considering the complexity of the stresses involved. A further outcome of this project is that the use of SARA appears universally suitable for the construction of new plants and in the service area, while the use of ThoMat appears suited for detail optimisation in the development process.

  15. Plasticity and creep of metals

    CERN Document Server

    Rusinko, Andrew

    2011-01-01

    Here is a systematic presentation of the postulates, theorems and principles of mathematical theories of plasticity and creep in metals, and their applications. Special attention is paid to analysis of the advantages and shortcomings of the classical theories.

  16. Principal physical mechanisms of material creep resistance and rupture at elevated temperatures

    International Nuclear Information System (INIS)

    Krishtal, M.A.

    1977-01-01

    Mechanisms of creep and long-term failure of refractory materials at different temperatures and stress levels are considered. At high temperatures and low stresses the diffusion (vacancial) mechanism is observed. Temperatures being low and stresses sufficiently high, dislocation mechanism involving avalanche dislocation break-off is manifested. Intermediate conditions provide other mechanisms, i.e. dislocation glide, dislocation climbing, grain-boundary and sub-grain-boundary mechanisms. Quantitative relationships between creep rate and some structural and kinetic parameters are discussed. Account of the creep mechanism is necessary when selecting methods for strengthening of alloys

  17. Digital signal processing of data from borehole creep closure

    International Nuclear Information System (INIS)

    Chakrabarti, S.; Patrick, W.C.; Duplancic, N.

    1987-01-01

    Digital signal processing, a technique commonly used in the fields of electrical engineering and communication technology, has been successfully used to analyze creep closure data obtained from a 0.91 m diameter by 5.13 deep borehole in bedded salt. By filtering the ''noise'' component of the closure data from a test borehole, important data trends were made more evident and average creep closure rates were able to be calculated. This process provided accurate estimates of closure rates that are used in the design of lined boreholes in which heat-generating transuranic nuclear wastes are emplaced at the Waste Isolation Pilot Plant

  18. Creep crack growth in phosphorus alloyed oxygen free copper

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Rui; Seitisleam, Facredin (Swerea KIMAB (Sweden)); Sandstroem, Rolf; Jin, Lai-Zhe (Materials Science and Engineering, Royal Inst. of Technology (Sweden))

    2011-01-15

    Using standard compact tension (CT) specimens taken from a pierce and draw cylinder, creep crack growth (CCG) has been studied in phosphorus-alloyed oxygen-free copper (Cu-OFP) parent metal at 22, 75, 175, and 215 deg C. Pre- and post-test metallography are performed. At higher temperatures the rupture time of CCG is shorter by a factor up of 65 than that of uniaxial at same stress/reference stress. At 175 and 215 deg C, crack does grow by creep about 10 mm before final instantaneous failure. In contrast, there is hardly any visible crack growth at 22 and 75 deg C. The tests were interrupted after 5000 to 13000 hours. For ruptured tests at 175 and 215 deg C, strongly elongated and deformed grains are observed adjacent to crack. Extensive and intergranular creep cavities and microcracks are found several mm around crack. For interrupted tests at 22 and 75 deg C, strongly elongated and deformed grains, creep cavities, as well as microcracks are observed close to crack tip. Surface cracks from both sides have initiated and grown about 45 deg to the load direction towards inside. For the interrupted tests, hardness adjacent to crack tip has more than doubled because of work hardening, or heavy deformation. This is consistent with large crack tip opening. The true strain at the crack tip is estimated to 10 and 4 for the tests at 22 and 75 deg C, respectively. The stress state behind the crack tip has been modelled with FEM. Stress relaxation after loading has also been taken into account. A model for the creep damage based on the creep strain rate has been formulated that can describe the uniaxial creep rupture data without fitting parameters. Based on the formulation for the creep damage, a model for the crack propagation has been set up. When the creep damage has reached the value unity in front of the crack tip, the crack is assumed to propagate. Taking multiaxial effects into account the observed life times of the CT specimens can be well described. The multiaxial

  19. Creep crack growth in phosphorus alloyed oxygen free copper

    International Nuclear Information System (INIS)

    Wu, Rui; Seitisleam, Facredin; Sandstroem, Rolf; Jin, Lai-Zhe

    2011-01-01

    Using standard compact tension (CT) specimens taken from a pierce and draw cylinder, creep crack growth (CCG) has been studied in phosphorus-alloyed oxygen-free copper (Cu-OFP) parent metal at 22, 75, 175, and 215 deg C. Pre- and post-test metallography are performed. At higher temperatures the rupture time of CCG is shorter by a factor up of 65 than that of uniaxial at same stress/reference stress. At 175 and 215 deg C, crack does grow by creep about 10 mm before final instantaneous failure. In contrast, there is hardly any visible crack growth at 22 and 75 deg C. The tests were interrupted after 5000 to 13000 hours. For ruptured tests at 175 and 215 deg C, strongly elongated and deformed grains are observed adjacent to crack. Extensive and intergranular creep cavities and microcracks are found several mm around crack. For interrupted tests at 22 and 75 deg C, strongly elongated and deformed grains, creep cavities, as well as microcracks are observed close to crack tip. Surface cracks from both sides have initiated and grown about 45 deg to the load direction towards inside. For the interrupted tests, hardness adjacent to crack tip has more than doubled because of work hardening, or heavy deformation. This is consistent with large crack tip opening. The true strain at the crack tip is estimated to 10 and 4 for the tests at 22 and 75 deg C, respectively. The stress state behind the crack tip has been modelled with FEM. Stress relaxation after loading has also been taken into account. A model for the creep damage based on the creep strain rate has been formulated that can describe the uniaxial creep rupture data without fitting parameters. Based on the formulation for the creep damage, a model for the crack propagation has been set up. When the creep damage has reached the value unity in front of the crack tip, the crack is assumed to propagate. Taking multiaxial effects into account the observed life times of the CT specimens can be well described. The multiaxial

  20. Failure diagnostics of rotational machines and machine groups in the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Nagy, I.; Kiss, G.

    1997-01-01

    Machine failure diagnostic system based on vibration analysis is described. The design of the measurements, the measurement procedure, and the evaluation of the results are presented. The common diagnostic technology for groups of rotation machinery has several advantages. The rules used for the evaluation of three-directional vibration measurements are shown, and the identification criteria of some specific failures are demonstrated. The steps of expert decision making and the diagnosis procedure are discussed through practical examples. (R.P.)

  1. Assessment of Creep Deformation, Damage, and Rupture Life of 304HCu Austenitic Stainless Steel Under Multiaxial State of Stress

    Science.gov (United States)

    Sahoo, K. C.; Goyal, Sunil; Parameswaran, P.; Ravi, S.; Laha, K.

    2018-03-01

    The role of the multiaxial state of stress on creep deformation and rupture behavior of 304HCu austenitic stainless steel was assessed by performing creep rupture tests on both smooth and notched specimens of the steel. The multiaxial state of stress was introduced by incorporating circumferential U-notches of different root radii ranging from 0.25 to 5.00 mm on the smooth specimens of the steel. Creep tests were carried out at 973 K over the stress range of 140 to 220 MPa. In the presence of notch, the creep rupture strength of the steel was found to increase with the associated decrease in rupture ductility. Over the investigated stress range and notch sharpness, the strengthening was found to increase drastically with notch sharpness and tended toward saturation. The fractographic studies revealed the mixed mode of failure consisting of transgranular dimples and intergranular creep cavitation for shallow notches, whereas the failure was predominantly intergranular for relatively sharper notches. Detailed finite element analysis of stress distribution across the notch throat plane on creep exposure was carried out to assess the creep failure of the material in the presence of notch. The reduction in von-Mises stress across the notch throat plane, which was greater for sharper notches, increased the creep rupture strength of the material. The variation in fracture behavior of the material in the presence of notch was elucidated based on the von-Mises, maximum principal, and hydrostatic stresses. Electron backscatter diffraction analysis of creep strain distribution across the notch revealed localized creep straining at the notch root for sharper notches. A master curve for predicting creep rupture life under the multiaxial state of stress was generated considering the representative stress having contributions from both the von-Mises and principal stress components of the stress field in the notch throat plane. Rupture ductility was also predicted based on the

  2. Investigation into Cause of High Temperature Failure of Boiler Superheater Tube

    Science.gov (United States)

    Ghosh, D.; Ray, S.; Roy, H.; Shukla, A. K.

    2015-04-01

    The failure of the boiler tubes occur due to various reasons like creep, fatigue, corrosion and erosion. This paper highlights a case study of typical premature failure of a final superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement, chemical analysis, oxide scale thickness measurement, microstructural examination are conducted as part of the investigations. Apart from these investigations, sulfur print, Energy Dispersive spectroscopy (EDS) and X ray diffraction analysis (XRD) are also conducted to ascertain the probable cause of failure of final super heater tube. Finally it has been concluded that the premature failure of the super heater tube can be attributed to the combination of localized high tube metal temperature and loss of metal from the outer surface due to high temperature corrosion. The corrective actions have also been suggested to avoid this type of failure in near future.

  3. Prediction of long-term creep curves

    International Nuclear Information System (INIS)

    Oikawa, Hiroshi; Maruyama, Kouichi

    1992-01-01

    This paper aims at discussing how to predict long-term irradiation enhanced creep properties from short-term tests. The predictive method based on the θ concept was examined by using creep data of ferritic steels. The method was successful in predicting creep curves including the tertiary creep stage as well as rupture lifetimes. Some material constants involved in the method are insensitive to the irradiation environment, and their values obtained in thermal creep are applicable to irradiation enhanced creep. The creep mechanisms of most engineering materials definitely change at the athermal yield stress in the non-creep regime. One should be aware that short-term tests must be carried out at stresses lower than the athermal yield stress in order to predict the creep behavior of structural components correctly. (orig.)

  4. Ferritic Alloys with Extreme Creep Resistance via Coherent Hierarchical Precipitates.

    Science.gov (United States)

    Song, Gian; Sun, Zhiqian; Li, Lin; Xu, Xiandong; Rawlings, Michael; Liebscher, Christian H; Clausen, Bjørn; Poplawsky, Jonathan; Leonard, Donovan N; Huang, Shenyan; Teng, Zhenke; Liu, Chain T; Asta, Mark D; Gao, Yanfei; Dunand, David C; Ghosh, Gautam; Chen, Mingwei; Fine, Morris E; Liaw, Peter K

    2015-11-09

    There have been numerous efforts to develop creep-resistant materials strengthened by incoherent particles at high temperatures and stresses in response to future energy needs for steam turbines in thermal-power plants. However, the microstructural instability of the incoherent-particle-strengthened ferritic steels limits their application to temperatures below 900 K. Here, we report a novel ferritic alloy with the excellent creep resistance enhanced by coherent hierarchical precipitates, using the integrated experimental (transmission-electron microscopy/scanning-transmission-electron microscopy, in-situ neutron diffraction, and atom-probe tomography) and theoretical (crystal-plasticity finite-element modeling) approaches. This alloy is strengthened by nano-scaled L21-Ni2TiAl (Heusler phase)-based precipitates, which themselves contain coherent nano-scaled B2 zones. These coherent hierarchical precipitates are uniformly distributed within the Fe matrix. Our hierarchical structure material exhibits the superior creep resistance at 973 K in terms of the minimal creep rate, which is four orders of magnitude lower than that of conventional ferritic steels. These results provide a new alloy-design strategy using the novel concept of hierarchical precipitates and the fundamental science for developing creep-resistant ferritic alloys. The present research will broaden the applications of ferritic alloys to higher temperatures.

  5. A metallographic examination of structural degradation during creep-fatigue

    International Nuclear Information System (INIS)

    Hales, R.

    1979-07-01

    A series of specimens of T316 stainless steel, which had been tested under creep-fatigue conditions, has been examined by optical and scanning electron microscopy. The development of cavities which are associated with grain-boundary carbide precipitates has been recorded. These cavities increase in size and number with increasing hold time at peak tensile strain and cause the propagating fatigue crack to follow an intergranular path. At a strain range of +- 0.25% the dominant damage mechanism is due to creep damage when the tensile hold time is greater than one minute. The fatigue crack which causes final failure is nucleated at a stress raiser and it is possible that in a smooth specimen failure may occur without the nucleation of a fatigue crack at all but rather by ductile shearing. (author)

  6. PROBABILITY OF FAILURE OF THE TRUDOCK CRANE SYSTEM AT THE WASTE ISOLATION PILOT PLANT (WIPP)

    International Nuclear Information System (INIS)

    Greenfield, M.A.; Sargent, T.J.

    2000-01-01

    This probabilistic analysis of WIPP TRUDOCK crane failure is based on two sources of failure data. The source for operator errors is the report by Swain and Guttman, NUREG/CR-1278-F, August 1983. The source for crane cable hook breaks was initially made by WIPP/WID-96- 2196, Rev. O by using relatively old (1970s) U.S. Navy data (NUREG-0612). However, a helpful analysis by R.K. Deremer of PLG guided the authors to values that were more realistic and more conservative, with the recommendation that the crane cable/hook failure rate should be 2.5 x 10-6 per demand. This value was adopted and used. Based on these choices a mean failure rate of 9.70 x 10-3(1/yr) was calculated. However, a mean rate by itself does not reveal the level of confidence to be associated with this number. Guidance to making confidence calculations came from the report by Swain and Guttman, who stated that failure data could be described by lognormal distributions. This is in agreement with the widely use d reports (by DOE and others) NPRD-95 and NPRD-91, on failure data. The calculations of confidence levels showed that the mean failure rate of 9.70x 10-3(1/yr) corresponded to a percentile value of approximately 71; i.e. there is a 71% likelihood that the failure rate is less than 9.70x 10-3(1/yr). One also calculated that there is a 95% likelihood that the failure rate is less than 29.6x 10-3(1/yr). Or, as stated previously, there is a 71% likelihood that not more than one dropped load will occur in 103 years. Also, there is a 95% likelihood that not more than one dropped load will occur in approximately 34 years. It is the responsibility of DOE to select the confidence level at which it desires to operate

  7. Online interferometric study of viscoelastic rupture and necking deformation of as-spun (iPP) fibres due to creep process.

    Science.gov (United States)

    Sokkar, Taha; El-Farahaty, Kermal; Azzam, Amira

    2015-01-01

    Creep deformation under constant load leads to rupture when the polymer chains can no longer separate and accommodate the load. This fracture phenomenon is investigated interferometrically. The creep behaviour of as-spun isotactic Polypropylene (iPP) fibres is studied at different stresses, different initial lengths and different radii. The creep rate, which defines the velocity of the creep deformation and the dimensional stability of the material, is studied. The failure time and stress of iPP due to creep process is determined. The necking deformation was in situ detected during creep process. The mean refractive indices (n(P) andn⊥) profiles of iPP fibres were determined at different positions along the fibre axis before and after necking. The relation between the creep behaviour and different optical and structural parameters is investigated. Microinterferograms are given for illustration. © 2015 The Authors Journal of Microscopy © 2015 Royal Microscopical Society.

  8. Application range affected by software failures in safety relevant instrumentation and control systems of nuclear power plants

    International Nuclear Information System (INIS)

    Jopen, Manuela; Mbonjo, Herve; Sommer, Dagmar; Ulrich, Birte

    2017-03-01

    This report presents results that have been developed within a BMUB-funded research project (Promotion Code 3614R01304). The overall objective of this project was to broaden the knowledge base of GRS regarding software failures and their impact in software-based instrumentation and control (I and C) systems. To this end, relevant definitions and terms in standards and publications (DIN, IEEE standards, IAEA standards, NUREG publications) as well as in the German safety requirements for nuclear power plants were analyzed first. In particular, it was found that the term ''software fault'' is defined differently and partly contradictory in the considered literature sources. For this reason, a definition of software fault was developed on the basis of the software life cycle of software-based I and C systems within the framework of this project, which takes into account the various aspects relevant to software faults and their related effects. It turns out that software failures result from latent faults in a software-based control system, which can lead to a non-compliant behavior of a software-based I and C system. Hereby a distinction should be made between programming faults and specification faults. In a further step, operational experience with software failures in software-based I and C systems in nuclear facilities and in nonnuclear sector was investigated. The identified events were analyzed with regard to their cause and impacts and the analysis results were summarized. Based on the developed definition of software failure and on the COMPSIS-classification scheme for events related to software based I and C systems, the COCS-classification scheme was developed to classify events from operating experience with software failures, in which the events are classified according to the criteria ''cause'', ''affected system'', ''impact'' and ''CCF potential''. This classification scheme was applied to evaluate the events identified in the framework of this project

  9. Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

    1996-01-01

    A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

  10. Creep and shrinkage analysis for concrete spent fuel dry storage module

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, D. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: zhangd@aecl.ca

    2009-07-01

    CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

  11. Creep and shrinkage analysis for concrete spent fuel dry storage module

    International Nuclear Information System (INIS)

    Zhang, D.

    2009-01-01

    CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

  12. Systematic analysis and prevention of human originated common cause failures in relation to maintenance activities at Finnish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Laakso, K. [VTT Industrial Systems, Espoo (Finland)

    2006-12-15

    The focus in human reliability analysis of nuclear power plants has traditionally been on human performance in disturbance conditions. On the other hand, human maintenance failures and design deficiencies, remained latent in the system, have an impact on the severity of a disturbance, e.g. by disabling safety-related equipment on demand. Especially common cause failures (CCFs) of safety related systems can affect the core damage risk to a significant extent. The topic has been addressed in Finnish studies, where experiences of latent human errors have been searched and analysed systematically from the maintenance history stored in the the power plant information systems of the Loviisa and Olkiluoto NPPs. Both the single and multiple errors (CCFs) were classified in detail and documented as error and event reports. The human CCFs involved human, organisational and technical factors. The review of the analysed single and multiple errors showed that instrumentation and control and electrical equipment are more prone to human error caused failure events than the other maintenance objects. The review of the analysed experience showed that most errors stem from the refuelling and maintenance outage periods. More than half of the multiple errors from the outages remained latent to the power operating periods. The review of the analysed multiple errors showed that difficulties with small plant modifications and planning of maintenance and operability were significant sources of common cause failures. The most dependent human errors originating from small modifications could be reduced by a more tailored planning and coverage of their start-up testing programs. Improvements could also be achieved by identifying better in work planning from the operating experiences those complex or intrusive repair and preventive maintenance work tasks and actions which are prone to errors. Such uncertain cases in important equipment require a more tailored work planning of the installation

  13. Systematic analysis and prevention of human originated common cause failures in relation to maintenance activities at Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Laakso, K.

    2006-12-01

    The focus in human reliability analysis of nuclear power plants has traditionally been on human performance in disturbance conditions. On the other hand, human maintenance failures and design deficiencies, remained latent in the system, have an impact on the severity of a disturbance, e.g. by disabling safety-related equipment on demand. Especially common cause failures (CCFs) of safety related systems can affect the core damage risk to a significant extent. The topic has been addressed in Finnish studies, where experiences of latent human errors have been searched and analysed systematically from the maintenance history stored in the the power plant information systems of the Loviisa and Olkiluoto NPPs. Both the single and multiple errors (CCFs) were classified in detail and documented as error and event reports. The human CCFs involved human, organisational and technical factors. The review of the analysed single and multiple errors showed that instrumentation and control and electrical equipment are more prone to human error caused failure events than the other maintenance objects. The review of the analysed experience showed that most errors stem from the refuelling and maintenance outage periods. More than half of the multiple errors from the outages remained latent to the power operating periods. The review of the analysed multiple errors showed that difficulties with small plant modifications and planning of maintenance and operability were significant sources of common cause failures. The most dependent human errors originating from small modifications could be reduced by a more tailored planning and coverage of their start-up testing programs. Improvements could also be achieved by identifying better in work planning from the operating experiences those complex or intrusive repair and preventive maintenance work tasks and actions which are prone to errors. Such uncertain cases in important equipment require a more tailored work planning of the installation

  14. Mechanical Behavior of Low Porosity Carbonate Rock: From Brittle Creep to Ductile Creep.

    Science.gov (United States)

    Nicolas, A.; Fortin, J.; Gueguen, Y.

    2014-12-01

    Mechanical compaction and associated porosity reduction play an important role in the diagenesis of porous rocks. They may also affect reservoir rocks during hydrocarbon production, as the pore pressure field is modified. This inelastic compaction can lead to subsidence, cause casing failure, trigger earthquake, or change the fluid transport properties. In addition, inelastic deformation can be time - dependent. In particular, brittle creep phenomena have been deeply investigated since the 90s, especially in sandstones. However knowledge of carbonates behavior is still insufficient. In this study, we focus on the mechanical behavior of a 14.7% porosity white Tavel (France) carbonate rock (>98% calcite). The samples were deformed in a triaxial cell at effective confining pressures ranging from 0 MPa to 85 MPa at room temperature and 70°C. Experiments were carried under dry and water saturated conditions in order to explore the role played by the pore fluids. Two types of experiments have been carried out: (1) a first series in order to investigate the rupture envelopes, and (2) a second series with creep experiments. During the experiments, elastic wave velocities (P and S) were measured to infer crack density evolution. Permeability was also measured during creep experiments. Our results show two different mechanical behaviors: (1) brittle behavior is observed at low confining pressures, whereas (2) ductile behavior is observed at higher confining pressures. During creep experiments, these two behaviors have a different signature in term of elastic wave velocities and permeability changes, due to two different mechanisms: development of micro-cracks at low confining pressures and competition between cracks and microplasticity at high confining pressure. The attached figure is a summary of 20 triaxial experiments performed on Tavel limestone under different conditions. Stress states C',C* and C*' and brittle strength are shown in the P-Q space: (a) 20°C and dry

  15. Impact of valve failures on the safety and reliability of light water nuclear power plants

    International Nuclear Information System (INIS)

    Riddington, J.W.; Reyer, R.J.

    1980-01-01

    A study of the causes of, and solutions for, recurrent valve failures has been performed. The frequency and root causes of valve problems were identified from licensee event reports and meetings with utility, NSSS, and valve manufacturer personnel. Three generic problems (stem leakage, seat leakage, and inadequate specification) and four valve specific problems were identified. The four valve specific problems and their principal causes are: (1) BWR pilot operated safety relief valves (pilot valve leakage); (2) spring loaded safety relief valves (water solid and two-phase flow behavior); (3) PWR feedwater regulating valves (trim degradation and packing failures); and (4) air operated solenoid valves (jamming due to foreign matter in service air). The first two valve specific problems are the subject of current industry programs. Programs intended to address stem leakage, seat leakage, timely exchange of valve failure information, testing of valves, and adequate specification, selection, and maintenance of valves will be outlined

  16. The estimate of ecological risk for ground ecosystems in case of nuclear power plant failures

    International Nuclear Information System (INIS)

    Kremlenkov, D.Y.; Kremlenkov, M.Y.

    2003-01-01

    Full text: The stochastic nature of radiation damage generates a need of forecasting information about possible consequences for environment and people. In this article it is given the estimate of probable damage to forest-and agricultural ecosystems from radionuclide emergency pollution in case of nuclear plant failures (for early emergency period). This estimate is based on radio-ecological risk conception which provide with the application of radioactive substances distribution models in atmosphere, as were calculation of absorbent radiation dose in critical ecosystem groups-calculation of probable area of lost ecosystems has been done by using the program written in Pascal. The quantitative estimate of environmental loss has been conducted for diverse classes of atmospheric stability. The value of ecological dose range (ELD) to coniferous forest is 30 Gy, deciduous forest - 300 Gy, agricultural crop - 60 Gy. The value of minimum ecological dose range (MELD) for all ecosystems is 10 Gy. In dose spread from MELD to ELD the ecological damage is proportional to absorbed dose. The ecological damage to ground ecosystems caused by cesium-137 and strontium-90 emergency pollution is primarily depended on the scale of radionuclide emergency pollution as well as weather conditions and radio-stability of critical vegetal ecosystem groups. On the assumption of a dose spread from MELD to ELD, ecological risk defined in probable ecosystem's destruction area is estimated: for cesium-137 pollution about 2 % of coniferous forest and from 4 to 9 % of deciduous forest; for strontium-90 pollution from 2 to 4 % of agricultural crop. As the scale of cesium-137 emergency pollution rise from 10 4 to 10 5 Cu the probable damage determined in ecosystem's destruction area increase 12-19 times to coniferous forest ecosystem and 15-36 times to deciduous forest according to weather conditions. The probable damage to coniferous and deciduous forest rise 11-17 times in proportion as the scale

  17. Application of the failure modes and effects analysis technique to theemergency cooling system of an experimental nuclear power plant

    International Nuclear Information System (INIS)

    Conceicao Junior, Osmar

    2009-01-01

    This study consists on the application of the Failure Modes and EffectsAnalysis (FMEA), a hazard identification and a risk assessment technique, tothe Emergency Cooling System (ECS) of an experimental nuclear power plant,which is responsible for mitigating the consequences of an eventual loss ofcoolant accident on the Pressurized Water Reactor (PWR). Such analysisintends to identify possible weaknesses on the design of the system andpropose some improvements in order to maximize its reliability. To achievethis goal a detailed study of the system was carried on (through itstechnical documentation), the correspondent reliability block diagram wasobtained, the FMEA analysis was executed and, finally, some suggestions werepresented. (author)

  18. Assessment of long-term creep strength of grade 91 steel

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, Kazuhiro; Sawada, Kota; Kushima, Hideaki [National Inst. for Materials Science, Tsukuba, Ibaraki (Japan)

    2010-07-01

    In 2004 and 2005 long-term creep rupture strength of ASME Grade 91 type steels of plate, pipe, forging and tube materials was evaluated in Japan by means of region splitting analysis method in consideration of 50% of 0.2% offset yield stress. According to the evaluated 100,000h creep rupture strength of 94MPa for plate, pipe and forging steels and 92MPa for tube steel at 600 C, allowable tensile stress of the steels regulated in the Interpretation for the Technical Standard for Thermal Power Plant was slightly reduced. New creep rupture data of the steels obtained in the long-term indicate further reduction of long-term creep rupture strength. Not only creep rupture strength, but also creep deformation property of the ASME Grade 91 steel was investigated and need of reevaluation of long-term creep strength of Grade 91 steel was indicated. A refinement of region splitting analysis method for creep rupture like prediction was discussed. (orig.)

  19. Investigation into the effects of operating conditions and design parameters on the creep life of high pressure turbine blades in a stationary gas turbine engine

    OpenAIRE

    Eshati, Samir; Abu, Abdullahi; Laskaridis, Panagiotis; Haslam, Anthony

    2011-01-01

    A physics–based model is used to investigate the relationship between operating conditions and design parameters on the creep life of a stationary gas turbine high pressure turbine (HPT) blade. A performance model is used to size the blade and to determine its stresses. The effects of radial temperature distortion, turbine inlet temperature, ambient temperature and compressor degradation on creep life are then examined. The results show variations in creep life and failure locat...

  20. Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

    Energy Technology Data Exchange (ETDEWEB)

    Dr. F. W. Brust; Dr. G. M. Wilkowski; Dr. P. Krishnaswamy; Mr. Keith Wichman

    2010-01-27

    methods often need to be addressed as plants age. Shutdown inspection intervals can only be designed using creep and creep-fatigue crack growth techniques. (5) The use of crack growth procedures can aid in examining the seriousness of creep damage in structural components. How cracks grow can be used to assess margins on components and lead to further safe operation. After examining the pros and cons of all these methods, the R5 code was chosen as the most up-to-date and validated high temperature creep and creep fatigue code currently used in the world at present. R5 is considered the leader because the code: (1) has well established and validated rules, (2) has a team of experts continually improving and updating it, (3) has software that can be used by designers, (4) extensive validation in many parts with available data from BE resources as well as input from Imperial college's database, and (5) was specifically developed for use in nuclear plants. R5 was specifically developed for use in gas cooled nuclear reactors which operate in the UK and much of the experience is based on materials and temperatures which are experienced in these reactors. If the next generation advanced reactors to be built in the US used these same materials within the same temperature ranges as these reactors, then R5 may be appropriate for consideration of direct implementation within ASME code NH or Section XI. However, until more verification and validation of these creep/fatigue crack growth rules for the specific materials and temperatures to be used in the GEN IV reactors is complete, ASME should consider delaying this implementation. With this in mind, it is this authors opinion that R5 methods are the best available for code use today. The focus of this work was to examine the literature for creep and creep-fatigue crack growth procedures that are well established in codes in other countries and choose a procedure to consider implementation into ASME NH. It is very important to

  1. Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

    International Nuclear Information System (INIS)

    Brust, F.W.; Wilkowski, G.M.; Krishnaswamy, P.; Wichman, Keith

    2010-01-01

    methods often need to be addressed as plants age. Shutdown inspection intervals can only be designed using creep and creep-fatigue crack growth techniques. (5) The use of crack growth procedures can aid in examining the seriousness of creep damage in structural components. How cracks grow can be used to assess margins on components and lead to further safe operation. After examining the pros and cons of all these methods, the R5 code was chosen as the most up-to-date and validated high temperature creep and creep fatigue code currently used in the world at present. R5 is considered the leader because the code: (1) has well established and validated rules, (2) has a team of experts continually improving and updating it, (3) has software that can be used by designers, (4) extensive validation in many parts with available data from BE resources as well as input from Imperial college's database, and (5) was specifically developed for use in nuclear plants. R5 was specifically developed for use in gas cooled nuclear reactors which operate in the UK and much of the experience is based on materials and temperatures which are experienced in these reactors. If the next generation advanced reactors to be built in the US used these same materials within the same temperature ranges as these reactors, then R5 may be appropriate for consideration of direct implementation within ASME code NH or Section XI. However, until more verification and validation of these creep/fatigue crack growth rules for the specific materials and temperatures to be used in the GEN IV reactors is complete, ASME should consider delaying this implementation. With this in mind, it is this authors opinion that R5 methods are the best available for code use today. The focus of this work was to examine the literature for creep and creep-fatigue crack growth procedures that are well established in codes in other countries and choose a procedure to consider implementation into ASME NH. It is very important to

  2. Datalogger for the creep laboratory

    International Nuclear Information System (INIS)

    Sambasivan, S.I.; Karthikeyan, T.V.; Chowdhary, D.M.; Anantharaman, P.N.

    1989-01-01

    The creep laboratory, MDL/ICGAR is a facility to study the creep properties of materials which are of interest to the fast reactor programme. The creep test is conducted over a few days to several months and years depending on the test variables employed. In these tests the creep strain and creep rate as a function of time are studied while the load and temperature are kept constant. The datalogger automates the process of recording the strain information as a function of time and also monitors the temperature throughout the test. The system handles 126 temperature channels and 42 strain channels from 27 machines. The temperature inputs are from the thermocouples and for cold junction compensation RTD's are used. An extensometer with a linear variable differential transformer (LVDT) or Super Linear Variable Capacitor (SLVC) form the set up to measure strain. The data logger consists of a front end analog input sub-system (AISS), a 8085 based Data Acquisition System (DAS) communicating to a microcomputer with CP/M operating system. The system responds to the user through the console and outputs of a dot matrix printer. The system, running a real time executive, also allows for on line enabling or disabling of a channel, printing of data, examining the current status and value, setting and getting time etc. (author)

  3. Large earthquakes and creeping faults

    Science.gov (United States)

    Harris, Ruth A.

    2017-01-01

    Faults are ubiquitous throughout the Earth's crust. The majority are silent for decades to centuries, until they suddenly rupture and produce earthquakes. With a focus on shallow continental active-tectonic regions, this paper reviews a subset of faults that have a different behavior. These unusual faults slowly creep for long periods of time and produce many small earthquakes. The presence of fault creep and the related microseismicity helps illuminate faults that might not otherwise be located in fine detail, but there is also the question of how creeping faults contribute to seismic hazard. It appears that well-recorded creeping fault earthquakes of up to magnitude 6.6 that have occurred in shallow continental regions produce similar fault-surface rupture areas and similar peak ground shaking as their locked fault counterparts of the same earthquake magnitude. The behavior of much larger earthquakes on shallow creeping continental faults is less well known, because there is a dearth of comprehensive observations. Computational simulations provide an opportunity to fill the gaps in our understanding, particularly of the dynamic processes that occur during large earthquake rupture and arrest.

  4. Creep Behaviour of Modified Mar-247 Superalloy

    Directory of Open Access Journals (Sweden)

    Cieśla M.

    2016-06-01

    Full Text Available The paper presents the results of analysis of creep behaviour in short term creep tests of cast MAR-247 nickel-based superalloy samples made using various modification techniques and heat treatment. The accelerated creep tests were performed under temperature of 982 °C and the axial stresses of σ = 150 MPa (variant I and 200 MPa (variant II. The creep behaviour was analysed based on: creep durability (creep rupture life, steady-state creep rate and morphological parameters of macro- and microstructure. It was observed that the grain size determines the creep durability in case of test conditions used in variant I, durability of coarse-grained samples was significantly higher.

  5. Creep-fatigue life property of FBR high-temperature structural materials under tension-torsion loading and life evaluation method

    International Nuclear Information System (INIS)

    Ogata, Takashi; Nitta, Akito

    1994-01-01

    Creep-fatigue damage in high temperature structural components in a FBR progress under multiaxial stress condition depending on their operating conditions and configuration. Therefore, multiaxial stress effects on creep-fatigue damage evolution must be clarified to make precise creep-fatigue damage evaluation of these components. In this study, creep-fatigue tests in FBR high temperature materials such as SUS304, 316FR stainless steels and a modified 9Cr steel were conducted under biaxial stress subjecting tension-compression and torsion loading, in order to examine biaxial stress effects on failure mechanism and life property, and to discuss creep-fatigue life evaluation methods under biaxial stress. Main results obtained in this study are summarized as follows: 1. The main cracks under cyclic torsion loading propagated by shear mode in three materials. But intergranular failure was occurred in SUS304 and 316FR, and transgranular failure was observed in Mod.9Cr steel. 2. Nonlinear damage accumulation model proposed based on uniaxial creep-fatigue test results was extended to apply for creep-fatigue damage evaluation under biaxial stress state by considering the biaxial stress effects on fatigue and creep damage evolution. 3. It was confirmed that creep-fatigue life under biaxial stress could be predicted by the extended evaluation method with higher accuracy than existing methods. (author)

  6. Development status und future possibilities for martensitic creep resistant steels

    Energy Technology Data Exchange (ETDEWEB)

    Hald, J. [Technical Univ. Denmark, Lyngby (Denmark). Dept. of Mechanical Engineering

    2010-07-01

    In the last four decades new stronger modified 9%Cr martensitic creep resistant steels have been introduced in power plants, which has enabled increases in maximum achievable steam conditions from the previous 250 bar and 540-560 C up to the values of 300 bar and 600-620 C currently being introduced all over the world. In order to further increase the steam parameters of steel based power plants up to a target value of 650 C/325 bar it is necessary to double the creep strength of the martensitic steels. At the same time the resistance against steam oxidation must be improved by an increase of the chromium content in the steels from 9% to 12%. However, so far all attempts to make stronger 12%Cr steels have led to breakdowns in long-term creep strength. Significant progress has been achieved in the understanding of microstructure stability of the martensitic 9-12%Cr steels: Observed microstructure instabilities in 11-12%Cr steels are explained by Z-phase precipitation, which dissolves fine MN nitrides. Improved understanding of effects of B and N on long-term creep properties has formed the basis of a series of new stronger 9%Cr test alloys with improved creep strength. In parallel 9%Cr test steels with low C content show very promising behavior in long-term tests. However, the 9%Cr steels must be surface coated to protect against steam oxidation at high temperature applications above 620%C. A possibility to use fine Z-phases for strengthening of the martensitic steels has been identified, and this opens a new pathway for development of stable strong 12%Cr steels. There are still good prospects for the realization of a 325 bar / 650 C steam power plant all based on steel. (orig.)

  7. Fluid Creep and Over-resuscitation.

    Science.gov (United States)

    Saffle, Jeffrey R

    2016-10-01

    Fluid creep is the term applied to a burn resuscitation, which requires more fluid than predicted by standard formulas. Fluid creep is common today and is linked to several serious edema-related complications. Increased fluid requirements may accompany the appropriate resuscitation of massive injuries but dangerous fluid creep is also caused by overly permissive fluid infusion and the lack of colloid supplementation. Several strategies for recognizing and treating fluid creep are presented. Copyright © 2016 Elsevier Inc. All rights reserved.

  8. Failure Analysis of PRDS Pipe in a Thermal Power Plant Boiler

    Science.gov (United States)

    Ghosh, Debashis; Ray, Subrata; Mandal, Jiten; Mandal, Nilrudra; Shukla, Awdhesh Kumar

    2018-04-01

    The pressure reducer desuperheater (PRDS) pipeline is used for reducing the pressure and desuperheating of the steam in different auxiliary pipeline. When the PRDS pipeline is failed, the reliability of the boiler is affected. This paper investigates the probable cause/causes of failure of the PRDS tapping line. In that context, visual inspection, outside diameter and wall thickness measurement, chemical analysis, metallographic examination and hardness measurement are conducted as part of the investigative studies. Apart from these tests, mechanical testing and fractographic analysis are also conducted as supplements. Finally, it has been concluded that the PRDS pipeline has mainly failed due to graphitization due to prolonged exposure of the pipe at higher temperature. The improper material used is mainly responsible for premature failure of the pipe.

  9. Influence of variations in creep curve on creep behavior of a high-temperature structure

    International Nuclear Information System (INIS)

    Hada, Kazuhiko

    1986-01-01

    It is one of the key issues for a high-temperature structural design guideline to evaluate the influence of variations in creep curve on the creep behavior of a high-temperature structure. In the present paper, a comparative evaluation was made to clarify such influence. Additional consideration was given to the influence of the relationship between creep rupture life and minimum creep rate, i.e., the Monkman-Grant's relationship, on the creep damage evaluation. The consideration suggested that the Monkman-Grant's relationship be taken into account in evaluating the creep damage behavior, especially the creep damage variations. However, it was clarified that the application of the creep damage evaluation rule of ASME B and P.V. Code Case N-47 to the ''standard case'' which was predicted from the average creep property would predict the creep damage on the safe side. (orig./GL)

  10. High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Manzel, R.; Everett, M. R.; Graham, L. W.

    1971-05-15

    The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 1021 n cm-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.

  11. Creep damage behaviour of modified 9Cr-1Mo steel weld joints

    International Nuclear Information System (INIS)

    Sakthivel, T.; Laha, K.; Vasudevan, M.; Panneer Selvi, S.

    2016-01-01

    Creep deformation and rupture behaviour of modified 9Cr-1Mo steel weld joints fabricated by single-pass activated TIG (A-TIG) and shielded metal arc welding (SMAW) processes have been investigated at 923 K over a stress range of 50 to 110 MPa after post weld heat treatment (PWHT). The weld joints exhibited significantly lower creep rupture lives than the base metal at lower applied stresses. Creep rupture location of the weld joints were found to occur in the ICHAZ. An extensive localized creep deformation, coarsening of M 23 C 6 precipitates in the ICHAZ with creep exposure led to the premature type IV failure of the joints. The coarsening of M 23 C 6 precipitates was extensive in the mid-section of the ICHAZ than the sub-surface of the joints, and was more predominant in the SMAW joint. While A-TIG weld joint exhibited reduced creep cavitation and coarsening of M 23 C 6 precipitates due to lower deformation constraints by adjacent regions in the ICHAZ. Hence, A-TIG weld joint exhibited higher creep rupture life than the SMAW joint. (author)

  12. Cumulative fatigue and creep-fatigue damage at 3500C on recrystallized zircaloy 4

    International Nuclear Information System (INIS)

    Brun, G.; Pelchat, J.; Floze, J.C.; Galimberti, M.

    1985-06-01

    An experimental programme undertaken by C.E.A., E.D.F. and FRAGEMA with the aim of characterizing the fatigue and creep fatigue behaviour of zircaloy-4 following annealing treatments (recrystallized, stress-delived) is in progress. The results given below concern only recrystallized material. Cyclic properties, low-cycle fatigue curves and creep behaviour laws under stresses have been established. Sequential tests of pure fatigue and creep-fatigue were performed. The cumulative life fractions at fracture depend on the sequence of leading, stress history and number of cycles of prestressing. The MINER's rule appears to be conservative with regard to a low-high loading sequence whereas it is not for the reverse high-low loading sequences. Fatigue and creep damage are not interchangeable. Pre-creep improves the fatigue resistance. Pre-fatigue improves the creep strength as long as the beneficial effect of cyclic hardening overcomes the damaging effect of surface cracking. The introduction of a tension hold time into the fatigue cycle slightly increases cyclic hardening and reduces the number of cycles to failure. For hold times of less than one hour, the sum of fatigue and creep life fractions is closed to one

  13. Effect of grain boundary complexions on the deformation behavior of Ni bicrystal during bending creep.

    Science.gov (United States)

    Reddy, K Vijay; Pal, Snehanshu

    2018-03-07

    The dependence of creep deformation behavior of nickel bicrystal specimens on grain boundary (GB) complexion was investigated by performing a simulated bending creep test using molecular dynamics methods. Strain burst phenomena were observed during the low temperature [500 K, i.e., creep process. Atomic strain and dislocation analyses showed that the time of occurrence of strain burst depends on how easily GB migration happens in bicrystal specimens. Specimens with kite monolayer segregation GB complexion were found to be stable at low temperature (500 K), whereas specimens with split-kite GB complexion were stable at a comparatively higher temperature (900 K). In case of further elevated creep temperatures, e.g., 1100 K and 1300 K, split-kite GB complexion becomes unstable and leads to early failure of the specimen at those temperatures. Additionally, it was observed that split-kite bilayer segregation and normal kite GB complexions exhibit localized increases in elastic modulus during bending creep process, occurring at temperatures of 1100 K and 1300 K, respectively, due to the formation of interpenetrating icosahedral clusters. Graphical abstract Representative creep curves during bending creep deformation of various grain boundary complexions at 900 K.

  14. Creep Deformation and Rupture Behavior of Single- and Dual-Pass 316LN Stainless-Steel-Activated TIG Weld Joints

    Science.gov (United States)

    Vijayanand, V. D.; Vasudevan, M.; Ganesan, V.; Parameswaran, P.; Laha, K.; Bhaduri, A. K.

    2016-06-01

    Creep deformation and rupture behavior of single-pass and dual-pass 316LN stainless steel (SS) weld joints fabricated by an autogenous activated tungsten inert gas welding process have been assessed by performing metallography, hardness, and conventional and impression creep tests. The fusion zone of the single-pass joint consisted of columnar zones adjacent to base metals with a central equiaxed zone, which have been modified extensively by the thermal cycle of the second pass in the dual-pass joint. The equiaxed zone in the single-pass joint, as well as in the second pass of the dual-pass joint, displayed the lowest hardness in the joints. In the dual-pass joint, the equiaxed zone of the first pass had hardness comparable to the columnar zone. The hardness variations in the joints influenced the creep deformation. The equiaxed and columnar zone in the first pass of the dual-pass joint was more creep resistant than that of the second pass. Both joints possessed lower creep rupture life than the base metal. However, the creep rupture life of the dual-pass joint was about twofolds more than that of the single-pass joint. Creep failure in the single-pass joint occurred in the central equiaxed fusion zone, whereas creep cavitation that originated in the second pass was blocked at the weld pass interface. The additional interface and strength variation between two passes in the dual-pass joint provides more restraint to creep deformation and crack propagation in the fusion zone, resulting in an increase in the creep rupture life of the dual-pass joint over the single-pass joint. Furthermore, the differences in content, morphology, and distribution of delta ferrite in the fusion zone of the joints favors more creep cavitation resistance in the dual-pass joint over the single-pass joint with the enhancement of creep rupture life.

  15. Application of aerospace failure-reporting systems to power plants. Final report

    International Nuclear Information System (INIS)

    Koukol, J.F.; Lapin, E.E.; Leverton, W.F.; Pickering, W.H.

    1980-06-01

    Failure reporting and analysis is a principal element of the overall quality assurance scheme that helped achieve, and now sustains, a high level of reliability in our national aerospace effort. The aerospace endeavor has many points of congruence with other highly technological activities. These are marked by great economic investment, an extended interval between concept and final implementation, the involvement of many independent entities with the government exercising a dominating influence, a considerable exposure to public view and review by public bodies, a notoriety accompanying untoward events, and extreme consequences attending failure. This report is written in the expectation that the lessons learned in arriving at the present state in aerospace can be adopted by others. It is the object of the report to illuminate the essential features of the aerospace failure reporting system. Two schemes are described. One typifies that which is currently employed by the Jet Propulsion Laboratory (JPL) operated by the California Institute of Technology for the NASA/JPL Voyager project and is based on procedures developed over several decades of deep space exploration. The other is typical of that employed by the Space Divison of the Air Force for military space programs

  16. Design of creep machine and creep specimen chamber for carrying out creep tests in flowing liquid sodium

    Energy Technology Data Exchange (ETDEWEB)

    Ravi, S., E-mail: sravi@igcar.gov.in; Laha, K.; Sakthy, S.; Mathew, M.D.; Jayakumar, T.

    2014-02-15

    Highlights: • Design of a lever type creep machine for carrying out creep test in flowing sodium. • Leveling of lever during creep was achieved by automated movement of fulcrum. • Design of creep chamber for providing constant sodium flow rate across creep specimen. • Minimum use of bellow in chamber for sodium containment and mechanical isolation. • Mini-lever mechanism to counter balance load reduction on specimen due to bellow stiffness. - Abstract: A creep testing system has been designed, fabricated, installed and validated for carrying out creep tests in flowing liquid sodium. The testing system consists of two sections namely creep testing machine and an environmental chamber. The testing system has the ability of (i) applying tensile load to the test specimen through a lever, (ii) monitoring continuously the creep elongation and (iii) allowing sodium to flow around the creep specimen at constant velocity. The annular space between the creep specimen and the environmental chamber has been suitably designed to maintain constant sodium flow velocity. Primary and secondary bellows are employed in the environmental chamber to (i) mechanically isolate the creep specimen, (ii) prevent the flowing sodium in contact with air and (iii) maintain an argon gas cover to the leaking sodium if any from primary bellow, with a provision to an alarm get activated by a spark plug. The lever-horizontality during creep test has been maintained by automatically lifting up the fulcrum instead of lowering down the pull rod as conventionally used. A mini lever mechanism has been incorporated in the load train to counter balance the load reduction on specimen from the changing stiffness of the bellows. The validation of the testing system has been established by carrying out creep tests on 316L(N) stainless steel at 873 K over a wide stress range and comparing the results with those obtained in air by employing the developed and conventional creep testing machines.

  17. Finite element simulation for creep crack growth

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki; Sasaki, Toru; Nakagaki, Michihiko; Brust, F.W.

    1992-01-01

    A finite element method was applied to a generation phase simulation of creep crack growth. Experimental data on creep crack growth in a 1Cr-1Mo-1/4V steel compact tension specimen were numerically simulated using a node-release technique and the variations of various fracture mechanics parameters such as CTOA, J, C * and T * during creep crack growth were calculated. The path-dependencies of the integral parameters J, C * and T * were also obtained to examine whether or not they could characterize the stress field near the tip of a crack propagating under creep condition. The following conclusions were obtained from the present analysis. (1) The J integral shows strong path-dependency during creep crack growth, so that it is does not characterize creep crack growth. (2) The C * integral shows path-dependency to some extent during creep crack growth even in the case of Norton type steady state creep law. Strictly speaking, we cannot use it as a fracture mechanics parameter characterizing creep crack growth. It is, however, useful from the practical viewpoint because it correlates well the rate of creep crack growth. (3) The T * integral shows good path-independency during creep crack growth. Therefore, it is a candidate for a fracture mechanics parameter characterizing creep crack growth. (author)

  18. Creep resistance and material degradation of a candidate Ni–Mo–Cr corrosion resistant alloy

    Energy Technology Data Exchange (ETDEWEB)

    Shrestha, Sachin L., E-mail: sachin@ansto.gov.au [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation (ANSTO), Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia); Bhattacharyya, Dhriti [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation (ANSTO), Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia); Yuan, Guangzhou; Li, Zhijun J. [Center of Thorium Molten Salts Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences (China); Budzakoska-Testone, Elizabeth; De Los Reyes, Massey; Drew, Michael; Edwards, Lyndon [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation (ANSTO), Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia)

    2016-09-30

    This study investigated the creep deformation properties of GH3535, a Ni–Mo–Cr corrosion resistant structural alloy being considered for use in future Gen IV molten salt nuclear reactors (MSR) operating at around 700 °C. Creep testing of the alloy was conducted at 650–750 °C under applied stresses between 85–380 MPa. From the creep rupture results the long term creep strain and rupture life of the alloy were estimated by applying the Dorn Shepard and Larson Miller time-temperature parameters and the alloy's allowable ASME design stresses at the MSR's operating temperature were evaluated. The material's microstructural degradation at creep rupture was characterised using scanning electron microscopy (SEM), electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). The microstructural study revealed that the material failure was due to wedge cracking at triple grain boundary points and cavitation at coarse secondary grain boundary precipitates, nucleated and grown during high temperature exposure, leading to intergranular crack propagation. EBSD local misorientation maps clearly show that the root cause of cavitation and crack propagation was due to large strain localisation at the grain boundaries and triple points instigated by grain boundary sliding during creep deformation. This caused the grain boundary decohesion and subsequent material failure.

  19. Low cycle fatigue and creep fatigue behavior of alloy 617 at high temperature

    International Nuclear Information System (INIS)

    Cabet, Celine; Carroll, Laura; Wright, Richard

    2013-01-01

    Alloy 617 is the leading candidate material for an intermediate heat exchanger (IHX) application of the very high temperature nuclear reactor (VHTR), expected to have an outlet temperature as high as 950 C. Acceptance of Alloy 617 in Section III of the ASME Code for nuclear construction requires a detailed understanding of the creep-fatigue behavior. Initial creep-fatigue work on Alloy 617 suggests a more dominant role of environment with increasing temperature and/or hold times evidenced through changes in creep-fatigue crack growth mechanisms and failure life. Continuous cycle fatigue and creep-fatigue testing of Alloy 617 was conducted at 950 C and 0.3% and 0.6% total strain in air to simulate damage modes expected in a VHTR application. Continuous cycle fatigue specimens exhibited transgranular cracking. Intergranular cracking was observed in the creep-fatigue specimens and the addition of a hold time at peak tensile strain degraded the cycle life. This suggests that creep-fatigue interaction occurs and that the environment may be partially responsible for accelerating failure. (authors)

  20. Experimental creep behaviour determination of cladding tube materials under multi-axial loadings

    International Nuclear Information System (INIS)

    Grosjean, Catherine; Poquillon, Dominique; Salabura, Jean-Claude; Cloue, Jean-Marc

    2009-01-01

    Cladding tubes are structural parts of nuclear plants, submitted to complex thermomechanical loadings. Thus, it is necessary to know and predict their behaviour to preserve their integrity and to enhance their lifetime. Therefore, a new experimental device has been developed to control the load path under multi-axial load conditions. The apparatus is designed to determine the thermomechanical behaviour of zirconium alloys used for cladding tubes. First results are presented. Creep tests with different biaxial loadings were performed. Results are analysed in terms of thermal expansion and of creep strain. The anisotropy of the material is revealed and iso-creep strain curves are given.

  1. Creep behaviour of thin walled composite tubes

    International Nuclear Information System (INIS)

    Thiebaud, F.; Muzic, B.; Perreux, D.; Varchon, D.; Oytana, C.; Lebras, J.

    1993-01-01

    Fiber reinforced composites are more and more employed in high performance structure for nuclear power plant, mainly as water piping tubes. The increase of the use of composites is due to the advantages that they give : high stiffness, large ultimate strength, corrosion resistance. This last advantage is sought for the pieces in contact with water, and it's one of the reason why the composite can be preferred to metal. However the mechanical behaviour of composite is actually poorly known. The high anisotropy is the main difficulty to obtain a realistic model of behaviour. This problem implies that the safety factor used in the design of structure is often too large. In this article a general overview of the mechanical behaviour of tube made in glass epoxy material is proposed. We discuss especially the creep behaviour under biaxial loadings. The form of the proposed model presently allows predicting a nonlinearity of the behaviour and provides a good correlation with the experimental data of several tests not described in this paper. It accounts for the change of the Poisson ratio during creep and cyclic tests. However the complete identification requires long time testings and consequently the model must be corrected to take into account the damage which occurs in these cases

  2. Creep-fatigue life assessment of cruciform weldments using the linear matching method

    International Nuclear Information System (INIS)

    Gorash, Yevgen; Chen, Haofeng

    2013-01-01

    This paper presents a creep-fatigue life assessment of a cruciform weldment made of the steel AISI type 316N(L) and subjected to reversed bending and cyclic dwells at 550 °C using the Linear Matching Method (LMM) and considering different weld zones. The design limits are estimated by the shakedown analysis using the LMM and elastic-perfectly-plastic material model. The creep-fatigue analysis is implemented using the following material models: 1) Ramberg–Osgood model for plastic strains under saturated cyclic conditions; 2) power-law model in “time hardening” form for creep strains during primary creep stage. The number of cycles to failure N ⋆ under creep-fatigue interaction is defined by: a) relation for cycles to fatigue failure N ∗ dependent on numerical total strain range Δε tot for the fatigue damage ω f ; b) long-term strength relation for the time to creep rupture t ∗ dependent on numerical average stress σ ¯ during dwell Δt for the creep damage ω cr ; c) non-linear creep-fatigue interaction diagram for the total damage. Numerically estimated N ⋆ for different Δt and Δε tot shows good quantitative agreement with experiments. A parametric study of different dwell times Δt is used to formulate the functions for N ⋆ and residual life L ⋆ dependent on Δt and normalised bending moment M -tilde , and the corresponding contour plot intended for design applications is created. -- Highlights: ► Ramberg–Osgood model is used for plastic strains under saturated cyclic conditions. ► Power-law model in time-hardening form is used for creep strains during dwells. ► Life assessment procedure is based on time fraction rule to evaluate creep damage. ► Function for cycles to failure is dependent on dwell period and normalised moment. ► Function for FSRF dependent on dwell period takes into account the effect of creep

  3. A code for simulation of human failure events in nuclear power plants: SIMPROC

    International Nuclear Information System (INIS)

    Gil, Jesus; Fernandez, Ivan; Murcia, Santiago; Gomez, Javier; Marrao, Hugo; Queral, Cesar; Exposito, Antonio; Rodriguez, Gabriel; Ibanez, Luisa; Hortal, Javier; Izquierdo, Jose M.; Sanchez, Miguel; Melendez, Enrique

    2011-01-01

    Over the past years, many Nuclear Power Plant organizations have performed Probabilistic Safety Assessments to identify and understand key plant vulnerabilities. As part of enhancing the PSA quality, the Human Reliability Analysis is essential to make a realistic evaluation of safety and about the potential facility's weaknesses. Moreover, it has to be noted that HRA continues to be a large source of uncertainty in the PSAs. Within their current joint collaborative activities, Indizen, Universidad Politecnica de Madrid and Consejo de Seguridad Nuclear have developed the so-called SIMulator of PROCedures (SIMPROC), a tool aiming at simulate events related with human actions and able to interact with a plant simulation model. The tool helps the analyst to quantify the importance of human actions in the final plant state. Among others, the main goal of SIMPROC is to check the Emergency Operating Procedures being used by operating crew in order to lead the plant to a safe shutdown plant state. Currently SIMPROC is coupled with the SCAIS software package, but the tool is flexible enough to be linked to other plant simulation codes. SIMPROC-SCAIS applications are shown in the present article to illustrate the tool performance. The applications were developed in the framework of the Nuclear Energy Agency project on Safety Margin Assessment and Applications (SM2A). First an introductory example was performed to obtain the damage domain boundary of a selected sequence from a SBLOCA. Secondly, the damage domain area of a selected sequence from a loss of Component Cooling Water with a subsequent seal LOCA was calculated. SIMPROC simulates the corresponding human actions in both cases. The results achieved shown how the system can be adapted to a wide range of purposes such as Dynamic Event Tree delineation, Emergency Operating Procedures and damage domain search.

  4. Mechanical behavior of 9Cr-1Mo-1V steel due to creep fatigue deformation

    International Nuclear Information System (INIS)

    Kim, Sang Tae; Kim, Jae Kyoung; Lee, Hak Sun; Oh, Sang Hyun; Kwun, Sook In; Kim, Chung Seok

    2005-01-01

    Creep-fatigue tests with trapezoid load wave were performed on a 9Cr-1Mo-1V steel at high temperature(550 .deg. C). Trapezoid load wave is considering about hold time for creep effects. we could find out some information in the relationship between number of cycles to failure and hold time. The number of cycles to failure depended on hold time. The cyclic behavior of 9Cr-1Mo-1V steel was characterized by cyclic softening with increasing number of cycles in high temperature. Also we could observe some cavity in the specimens. The size of cavity was different from each hold time

  5. Study of tertiary creep instability in several elevated-temperature structural materials

    International Nuclear Information System (INIS)

    Booker, M.K.; Sikka, V.K.

    1978-01-01

    Data for a number of common elevated temperature structural materials have been analyzed to yield mathematical predictions for the time and strain to tertiary creep at various rupture lives and temperatures. Materials examined include types 304 and 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, alloy 718, Hastelloy alloy X, and ERNiCr--3 weld metal. Data were typically examined over a range of creep temperatures for rupture lives ranging from less than 100 to greater than 10,000 hours. Within a given material, trends in these quantities can be consistently described, but it is difficult to directly relate the onset of tertiary creep to failure-inducing instabilities. A series of discontinued tests for alloy 718 at 649 and 620 0 C showed that the material fails by intergranular cracking but that no significant intergranular cracking occurs until well after the onset of tertiary creep

  6. A review of creep analysis and design under multi-axial stress states

    International Nuclear Information System (INIS)

    Yao, H.-T.; Xuan Fuzhen; Wang Zhengdong; Tu Shantung

    2007-01-01

    The existence of multi-axial states of stress cannot be avoided in elevated temperature components. It is essential to understand the associated failure mechanisms and to predict the lifetime in practice. Although metal creep has been studied for about 100 years, many problems are still unsolved, in particular for those involving multi-axial stresses. In this work, a state-of-the-art review of creep analysis and engineering design is carried out, with particular emphasis on the effect of multi-axial stresses. The existing theories and creep design approaches are grouped into three categories, i.e., the classical plastic theory (CPT) based approach, the cavity growth mechanism (CGM) based approach and the continuum damage mechanics (CDM) based approach. Following above arrangements, the constitutive equations and design criteria are addressed. In the end, challenges on the precise description of the multi-axial creep behavior and then improving the strength criteria in engineering design are presented

  7. A life evaluation under creep-fatigue-environment interaction of Ni-base wrought alloys

    International Nuclear Information System (INIS)

    Hattori, Hiroshi; Kitagawa, Masaki; Ohtomo, Akira; Itoh, Mitsuyoshi

    1986-01-01

    In order to determine a failure criteria under cyclic loading and affective environment for HTGR systems, a series of strain controlled low-cycle fatigue tests were carried out at HTGR maximum gas temperatures in air, in vacuum and in HTGR helium environments on two nickel-base wrought alloys, namely Inconel 617 and Hastelloy XR. This paper first describes the creep-fatigue-environment properties of these alloys followed by a proposal of an evaluation method of creep-fatigue-environment interaction based on the experimental data to define the more reasonable design criteria, which is a modification of the linear damage summation rule. Second, the creep-fatigue properties of Hastelloy XR at 900 deg C and the result evaluated by this proposed method are shown. This criterion is successfully applied to the life prediction at 900 deg C. In addition, the creep-fatigue properties of Hastelloy XR-II are discussed. (author)

  8. Success and failure of nuclear power plant Quality Assurance when applied to research and development activities

    Energy Technology Data Exchange (ETDEWEB)

    Michels, R.D.; Stark, W.A.; Eberhart, C.F.

    1986-01-01

    Quality Assurance in some form or fashion has been practiced at research and development laboratories and facilities for quite some time. This quality assurance effort has largely been that of the scientist taking good notes as the experiment or observation is conducted to provide adequate documentation of the research and development effort. With the advent of the national laboratories participating in such projects as the Waste Isolation Project Plant (WIPP), Three Mile Island, and the National Waste Storage Program, a need for planned, systematic quality assurance for research and development activities has come into being. Based on experience evolved from nuclear power plant and facility construction, the government has chosen to utilize for these new projects the criteria developed for use in construction of power plants. While these criteria have worked reasonably well in the construction, maintenance, and operation of those nuclear plants, their application to research and development activities has not been an unqualified success. This paper compares and contrasts four criteria; surveillance, inspection, procurement, and calibration for application in the research and development situation, and nuclear power plants.

  9. Success and failure of nuclear power plant Quality Assurance when applied to research and development activities

    International Nuclear Information System (INIS)

    Michels, R.D.; Stark, W.A.; Eberhart, C.F.

    1986-01-01

    Quality Assurance in some form or fashion has been practiced at research and development laboratories and facilities for quite some time. This quality assurance effort has largely been that of the scientist taking good notes as the experiment or observation is conducted to provide adequate documentation of the research and development effort. With the advent of the national laboratories participating in such projects as the Waste Isolation Project Plant (WIPP), Three Mile Island, and the National Waste Storage Program, a need for planned, systematic quality assurance for research and development activities has come into being. Based on experience evolved from nuclear power plant and facility construction, the government has chosen to utilize for these new projects the criteria developed for use in construction of power plants. While these criteria have worked reasonably well in the construction, maintenance, and operation of those nuclear plants, their application to research and development activities has not been an unqualified success. This paper compares and contrasts four criteria; surveillance, inspection, procurement, and calibration for application in the research and development situation, and nuclear power plants

  10. Early Detection of Plant Equipment Failures: A Case Study in Just-in-Time Maintenance

    International Nuclear Information System (INIS)

    Parlos, Alexander G.; Kim, Kyusung; Bharadwaj, Raj M.

    2001-01-01

    The development and testing of a model-based fault detection system for electric motors is briefly presented. The fault detection system was developed using only motor nameplate information. The fault detection results presented utilize only motor voltage and current sensor information, minimizing the need for expensive or intrusive sensors. Dynamic recurrent neural networks are used to predict the input-output response of a three-phase induction motor while using an estimate of the motor speed signal. Multiresolution (or wavelet) signal-processing techniques are used in combination with more traditional methods to estimate fault features for use in winding insulation and motor mechanical and electromechanical failure detection

  11. Early Detection of Plant Equipment Failures: A Case Study in Just-in-Time Maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Parlos, Alexander G.; Kim, Kyusung; Bharadwaj, Raj M.

    2001-06-17

    The development and testing of a model-based fault detection system for electric motors is briefly presented. The fault detection system was developed using only motor nameplate information. The fault detection results presented utilize only motor voltage and current sensor information, minimizing the need for expensive or intrusive sensors. Dynamic recurrent neural networks are used to predict the input-output response of a three-phase induction motor while using an estimate of the motor speed signal. Multiresolution (or wavelet) signal-processing techniques are used in combination with more traditional methods to estimate fault features for use in winding insulation and motor mechanical and electromechanical failure detection.

  12. Outline of incidents and failures of Japanese nuclear power plants during April 1995 and March 1996

    International Nuclear Information System (INIS)

    1996-01-01

    The number of incidents and failures reported to MITI by electric utility companies, as stipulated in the ''Electric Utility Industry Law'' and the ''Law for Regulation of Nuclear Source Material, Nuclear Fuel Material, and Reactor'', amounted to 14 cases in FY 1995. The number of reports per unit was 0.3, showing the same values with last year. The above 14 cases consists of 9 cases of manual shutdown and 1 case of automatic shutdown during operation (including test and adjustment operation), and 4 cases were found during reactor shutdown. Figs, 1 tab

  13. Failure analysis of globe control valves with spring-diaphragm actuator for nuclear power plant applications

    International Nuclear Information System (INIS)

    Khan, M.R.; Chen, W.W.H.; Wang, T.Y.

    1997-01-01

    The results of the failure analysis of a globe control valve with spring-diaphragm actuator indicated that the diaphragm failed because the service loading is close to the strength of the diaphragm. The resulting impact force is significantly larger than the plug guide strength and that cause it to bulge out after the impact. To improve the valve performance, proper torque should be used to tighten the actuator diaphragm case fasteners. A stronger actuator diaphragm could be used to provide additional safety margin during operation. Stiffening the plug guide may avoid jamming the bushing

  14. Study of the Atucha I nuclear power plant's residual heat removal system unavailability through the fault tree analysis and common cause failures

    International Nuclear Information System (INIS)

    Terrado, C.A.

    1991-06-01

    The present essay offers a comprehensive research of the Atucha I nuclear power plant's residual heat removal system unavailability, including Fault Tree Analysis and Common Cause Failures (CCF) treatment. The study is developed within the Event Tree perspective that considers the loss of external electrical power of the initiating event. The event was constructed by the Safety Evaluations Division of the Ezeiza Atomic Center in Argentina. According to the Event Tree, the research includes system demand during plant operation with 132 KV and emergency generation (Diesel motor generators). The system unavailability assessment is approached in two different ways: a) Considering independent failures only. b) Taking into account the existence of Common Cause Events, and modeling dependent failures. The Fault Tree quantification is played using the AIEA PSAPACK Code. The assessment data base is compiled from plant specific records and generic data bases like TECDOC 478. After Fault Tree model logic development, some general procedures used in common cause failures treating are applied to pick up another set of solutions. The results of the study are: a) Four Fault Trees have been developed to model the abovementioned system: 132 KV and emergency generation, both including and excluding CCF. b) The following unavailability values were obtained: 132 KV independent failures only: 7 10 -4 . Emergency generation independent failures only: 1.53 10 -2 . 132 KV dependent and independent failures: 3.6 10 -3 . Emergency generation dependent and independent failures: 1.74 10 -2 . The major conclusions obtained from the precedent results are: a) When using 132 KV system configuration, minimal cut sets involving common cause failures represents 81%from total system unavailability. b) The dependent failures treatment is an important task to be considered in safety assessments in order to reach more realistic values. (Author) [es

  15. Damage Assessment of Heat Resistant Steels through Electron BackScatter Diffraction Strain Analysis under Creep and Creep-Fatigue Conditions

    Science.gov (United States)

    Fujiyama, Kazunari; Kimachi, Hirohisa; Tsuboi, Toshiki; Hagiwara, Hiroyuki; Ogino, Shotaro; Mizutani, Yoshiki

    EBSD(Electron BackScatter Diffraction) analyses were conducted for studying the quantitative microstructural metrics of creep and creep-fatigue damage for austenitic SUS304HTB boiler tube steel and ferritic Mod.9Cr piping steel. KAM(Kernel Average Misorientation) maps and GOS(Grain Orientation Spread) maps were obtained for these samples and the area averaged values KAMave and GOSave were obtained. While the increasing trends of these misorientation metrics were observed for SUS304HTB steel, the decreasing trends were observed for damaged Mod.9Cr steel with extensive recovery of subgrain structure. To establish more universal parameter representing the accumulation of damage to compensate these opposite trends, the EBSD strain parameters were introduced for converting the misorientation changes into the quantities representing accumulated permanent strains during creep and creep-fatigue damage process. As KAM values were dependent on the pixel size (inversely proportional to the observation magnification) and the permanent strain could be expressed as the shear strain which was the product of dislocation density, Burgers vector and dislocation movement distance, two KAM strain parameters MεKAMnet and MεδKAMave were introduced as the sum of product of the noise subtracted KAMnet and the absolute change from initial value δKAMave with dislocation movement distance divided by pixel size. MεδKAMave parameter showed better relationship both with creep strain in creep tests and accumulated creep strain range in creep-fatigue tests. This parameter can be used as the strain-based damage evaluation and detector of final failure.

  16. Creep-fatigue damage in austenitic stainless steels

    International Nuclear Information System (INIS)

    Rezgui, Brahim.

    1980-06-01

    This is a study of hold time effects on the low cycle fatigue (L.C.F.) properties of 316L austenitic stainless steel at 600 0 C in air. Results obtained for different plastic strain levels indicate that a tension hold time at peak strain lead to a reduction in fatigue life. The importance of this effect depend on the length of hold period, and also on the strain amplitude. No saturation had been observed. Metallographic and microstructural analysis of failed specimens indicates mechanisms by which failure is produced. For continuous cycling the fractures occurs by the initiation and the propagation of a trans-granular crack. Creep damage in the bulk of material is formed during periods of tensile stress relaxation; it causes a change in the failure mode which became intergranular. It is the interaction between this creep-damage and fatigue cracks which is partly responsable for the reduction in the fatigue life. Predictions based upon linear cumulative damage method indicate that virgin material properties may be irrelevant in creep-fatigue interactions [fr

  17. On cyclic yield strength in definition of limits for characterisation of fatigue and creep behaviour

    Science.gov (United States)

    Gorash, Yevgen; MacKenzie, Donald

    2017-06-01

    This study proposes cyclic yield strength as a potential characteristic of safe design for structures operating under fatigue and creep conditions. Cyclic yield strength is defined on a cyclic stress-strain curve, while monotonic yield strength is defined on a monotonic curve. Both values of strengths are identified using a two-step procedure of the experimental stress-strain curves fitting with application of Ramberg-Osgood and Chaboche material models. A typical S-N curve in stress-life approach for fatigue analysis has a distinctive minimum stress lower bound, the fatigue endurance limit. Comparison of cyclic strength and fatigue limit reveals that they are approximately equal. Thus, safe fatigue design is guaranteed in the purely elastic domain defined by the cyclic yielding. A typical long-term strength curve in time-to-failure approach for creep analysis has two inflections corresponding to the cyclic and monotonic strengths. These inflections separate three domains on the long-term strength curve, which are characterised by different creep fracture modes and creep deformation mechanisms. Therefore, safe creep design is guaranteed in the linear creep domain with brittle failure mode defined by the cyclic yielding. These assumptions are confirmed using three structural steels for normal and high-temperature applications. The advantage of using cyclic yield strength for characterisation of fatigue and creep strength is a relatively quick experimental identification. The total duration of cyclic tests for a cyclic stress-strain curve identification is much less than the typical durations of fatigue and creep rupture tests at the stress levels around the cyclic yield strength.

  18. Point defects and the creep of metals

    International Nuclear Information System (INIS)

    Nichols, F.A.

    1976-01-01

    Basic concepts felt to be important in diffusion-controlled creep of metals are reviewed and it is suggested that such creep is controlled by edge-dislocation climb under a rather wide range of conditions. The effect of a damage-producing flux on such creep processes is explored. It is shown that processes such as Herring-Nabarro creep are unaffected by irradiation. Evidence is presented for a climb-plus-glide mechanism of radiation creep for stresses above unirradiated yield or flow stresses. At lower stresses a preferential dislocation loop nucleation model is suggested

  19. Creep cavitation effects in polycrystalline alumina

    International Nuclear Information System (INIS)

    Porter, J.R.; Blumenthal, W.; Evans, A.G.

    1981-01-01

    Fine grained polycrystalline alumina has been deformed in creep at high temperatures, to examine the evolution of cavities at grain boundaries. Cavities with equilibrium and crack-like morphologies have been observed, distributed nonuniformly throughout the material. The role of these cavities during creep has been described. A transition from equilibrium to crack-like morphology has been observed and correlated with a model based on the influence of the surface to boundary diffusivity ratio and the local tensile stress. The contribution of cavitation to the creep rate and total creep strain has been analyzed and excluded as the principal cause of the observed non-linear creep rate

  20. Creep of Hi-Nicalon S Fiber Tows at Elevated Temperature in Air and in Steam

    Science.gov (United States)

    2013-03-01

    with steam port and alumina steam feeding tube with slot to divert steam away from directly impacting the fiber tow specimen . . . . 25 4.7 Hi-Nicalon...The activation energy for the creep of Hi-Nicalon fibers was determined to be 360 kJ/mol and the predominant creep mechanism was identified as...engines, gas turbines for electrical power/steam cogeneration , as well as nuclear power plant components. It is recognized that the structural

  1. Creep of ice: further studies

    International Nuclear Information System (INIS)

    Heard, H.C.; Durham, W.B.; Kirby, S.H.

    1987-01-01

    Detailed studies have been done of ice creep as related to the icy satellites, Ganymede and Callisto. Included were: (1) the flow of high-pressure water ices II, III, and V, and (2) frictional sliding of ice I sub h. Work was also begun on the study of the effects of impurities on the flow of ice. Test results are summarized

  2. Thermal creep of Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Murty, K.L.; Clevinger, G.S.; Papazoglou, T.P.

    1977-01-01

    Data on the hoop creep characteristics of Zircaloy tubing were collected at temperatures between 600 F and 800 F, and at stress levels ranging from 10 ksi to 25 ksi using internal pressurization tests. At low driving forces, exposures as long as 2000 hours were found insufficient to establish steady state creep. The experimental data at temperatures of 650 F to 800 F correlate well with an exponential stress dependence, and the activation energy for creep was found to be in excellent agreement with that for self-diffusion. The range of stresses and temperatures is too small to study the overall effect of these variables on the activation energy for creep. The experimental steady state creep-rates and those predicted from the creep equation used agree within a factor of 1.3. These correlations imply that the mechanism for hoop creep of Zircaloy-4 cladding is characterized by an activation energy of approximately 60 kcal/mole and an activation area of about 20b 3 . In addition, the exponential stress dependence implies that the activation area for creep is stress-independent. These results suggest that the climb of edge dislocations is the rate controlling mechanism for creep of Zircaloy-4. The transient creep regime was also analysed on the premise that primary creep is directly related to the rate of dispersal of dislocation entanglements by climb. (Auth.)

  3. Modification of creep and low cycle fatigue behaviour induced by welding

    Directory of Open Access Journals (Sweden)

    A. Carofalo

    2014-10-01

    Full Text Available In this work, the mechanical properties of Waspaloy superalloy have been evaluated in case of welded repaired material and compared to base material. Test program considered flat specimens on base and TIG welded material subjected to static, low-cycle fatigue and creep test at different temperatures. Results of uniaxial tensile tests showed that the presence of welded material in the gage length specimen does not have a relevant influence on yield strength and UTS. However, elongation at failure of TIG material was reduced with respect to the base material. Moreover, low-cycle fatigue properties have been determined carrying out tests at different temperature (room temperature RT and 538°C in both base and TIG welded material. Welded material showed an increase of the data scatter and lower fatigue strength, which was anyway not excessive in comparison with base material. During test, all the hysteresis cycles were recorded in order to evaluate the trend of elastic modulus and hysteresis area against the number of cycles. A clear correlation between hysteresis and fatigue life was found. Finally, creep test carried out on a limited number of specimens allowed establishing some changes about the creep rate and time to failure of base and welded material. TIG welded specimen showed a lower time to reach a fixed strain or failure when a low stress level is applied. In all cases, creep behaviour of welded material is characterized by the absence of the tertiary creep.

  4. Creep in Topopah Spring Member welded tuff. Yucca Mountain Site Characterization Project

    Energy Technology Data Exchange (ETDEWEB)

    Martin, R.J. III; Boyd, P.J.; Noel, J.S. [New England Research, Inc., White River Junction, VT (United States); Price, R.H. [Sandia National Labs., Albuquerque, NM (United States)

    1995-06-01

    A laboratory investigation has been carried out to determine the effects of elevated temperature and stress on the creep deformation of welded tuffs recovered from Busted Butte in the vicinity of Yucca Mountain, Nevada. Water saturated specimens of tuff from thermal/mechanical unit TSw2 were tested in creep at a confining pressure of 5.0 MPa, a pore pressure of 4.5 MPa, and temperatures of 25 and 250 C. At each stress level the load was held constant for a minimum of 2.5 {times} 10{sup 5} seconds and for as long as 1.8 {times} 10{sup 6} seconds. One specimen was tested at a single stress of 80 MPa and a temperature of 250 C. The sample failed after a short time. Subsequent experiments were initiated with an initial differential stress of 50 or 60 MPa; the stress was then increased in 10 MPa increments until failure. The data showed that creep deformation occurred in the form of time-dependent axial and radial strains, particularly beyond 90% of the unconfined, quasi-static fracture strength. There was little dilatancy associated with the deformation of the welded tuff at stresses below 90% of the fracture strength. Insufficient data have been collected in this preliminary study to determine the relationship between temperature, stress, creep deformation to failure, and total failure time at a fixed creep stress.

  5. Creep in Topopah Spring Member welded tuff. Yucca Mountain Site Characterization Project

    International Nuclear Information System (INIS)

    Martin, R.J. III; Boyd, P.J.; Noel, J.S.; Price, R.H.

    1995-06-01

    A laboratory investigation has been carried out to determine the effects of elevated temperature and stress on the creep deformation of welded tuffs recovered from Busted Butte in the vicinity of Yucca Mountain, Nevada. Water saturated specimens of tuff from thermal/mechanical unit TSw2 were tested in creep at a confining pressure of 5.0 MPa, a pore pressure of 4.5 MPa, and temperatures of 25 and 250 C. At each stress level the load was held constant for a minimum of 2.5 x 10 5 seconds and for as long as 1.8 x 10 6 seconds. One specimen was tested at a single stress of 80 MPa and a temperature of 250 C. The sample failed after a short time. Subsequent experiments were initiated with an initial differential stress of 50 or 60 MPa; the stress was then increased in 10 MPa increments until failure. The data showed that creep deformation occurred in the form of time-dependent axial and radial strains, particularly beyond 90% of the unconfined, quasi-static fracture strength. There was little dilatancy associated with the deformation of the welded tuff at stresses below 90% of the fracture strength. Insufficient data have been collected in this preliminary study to determine the relationship between temperature, stress, creep deformation to failure, and total failure time at a fixed creep stress

  6. Ratchetting in the creep range

    International Nuclear Information System (INIS)

    Ponter, A.R.S.; Cocks, A.C.F.; Clement, G.; Roche, R.; Corradi, L.; Franchi, A.

    1985-01-01

    This report attempts to present a ''State of the Art'' of this problem from three contracting and complementary points of view which reflect separate traditions within the discipline of structural analysis. Part I gives a brief summary of the essential elements of the three constitutive parts and a set of conclusions and recommendations are then formulated. Part II is an attempt by a group at CEA Saclay, France, to distil from available experimental data a set of rules expressed in terms of the stress classifications of the ASME codes, which will ensure the prevention of excessive creep ratchetting. The resulting stresses to an effective (or reference) stress and the creep assessment is then made in terms of the creep produced by the effective stress. They aim at analytical procedures for LMFBR components that operate in the creep region and are subject to considerable thermal transients. Part III by Ponter and Cocks of the University of Leicester is a theoretical study of the problem using bounding and other approximate techniques. The problem is studied in a sequence of increasingly complex problems commencing with an isothermal structure subjected to constant load and terminating in a structure subjected to arbitrary cyclic thermal loading. The results are expressed in terms of a reference stress derived from a plastic shakedown solution, and a reference history of temperature. These techniques are capable of providing assessment of the creep deformation of a structure when the plastic shakedown properties of the structures are known. The particular circumstances which occur in a LMFBR are emphasized. Part IV by Corradi and Franchi discusses the methods by which finite element solution may be calculated. These are surveyed with particular reference to the numerical problems involved and the relationship between computational procedure and the form of the constitutive equation. 162 refs

  7. Creep-fatigue interactions in an austenitic stainless steel

    International Nuclear Information System (INIS)

    Majumdar, S.; Maiya, P.S.

    1978-01-01

    A phenomenological model of the interaction between creep and fatigue in Type 304 stainless steel at elevated temperatures is presented. The model is based on a crack-growth equation and an equation governing cavity growth, expressed in terms of current plastic strain and plastic strain rate. Failure is assumed to occur when a proposed interaction equation is satisfied. Various parameters of the equations can be obtained by correlation with continuously cycling fatigue and monotonic creep-rupture test data, without the use of any hold-time fatigue tests. Effects of various wave shapes such as tensile, compressive, and symmetrical hold on the low-cycle fatigue life can be computed by integrating the damage-rate equations along the appropriate loading path. Microstructural evidence in support of the proposed model is also discussed

  8. Low stress creep of stainless steel

    International Nuclear Information System (INIS)

    Crossland, I.G.; Clay, B.D.; Baker, C.

    1976-06-01

    The creep of 20%Cr, 25%Ni, Nb stainless steel has been examined at temperatures from 675 to 775 0 C at sheer stressed below 13 MPa and grain sizes from 6 to 20μm. The results have indicated that the initial creep rates were linearly dependent upon stress but with a threshold stress below which no creep occurred, i.e. Bingham behaviour; in addition, the creep activation energy at small strains was substantially lower than the lattice self-diffusion value and the initial creep rates were approximately related to the grain size through an inverse cube relation. It has been concluded that at low strains (approaching the initial elastic deflection) the creep mechanism was probably that of grain boundary diffusion creep (Coble, 1963) and this is further supported by the close agreement between the observed and theoretically predicted creep rate values. Steady-state creep rates were not observed; initially the creep rates fell rapidly with strain after which a more gradual decrease occurred. Whilst the creep rate - stress relationship continued to be of a Bingham form, the progressive reduction in creep rate with strain was found to be mainly attributable to an increase in the effective viscosity, threshold stress effects being generally of secondary importance. A model has been proposed which explains the initial creep rates as being due to Cable creep with elastic accommodation at grain boundary particles. At higher strains grain boundary collapse caused by vacancy sinking is accommodated at precipitate particles by plastic deformation of the adjacent matrix material. (author)

  9. Computer simulation of black out followed by multiple failures in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1989-01-01

    The computer code RELAP 5/MOD 1 has been utilized to investigate the thermal-hydraulic behaviour of a standard 1300 MWe pressurized water reactor plant of the KWU design during a station blackout following a inadequate performance of the pressurizer and steam generator safety valves. During the simulation the reactor scram system the emergency coolant system of the primary loop and the emergency Feedwater system of the secondary loop are considered inactive. (author) [pt

  10. Potential seismic structural failure modes associated with the Zion Nuclear Plant. Seismic safety margins research program (Phase I). Project VI. Fragilities

    International Nuclear Information System (INIS)

    1979-10-01

    The Zion 1 and 2 Nuclear Power Plant consists of a number of structures. The most important of these from the viewpoint of safety are the containment buildings, the auxiliary building, the turbine building, and the crib house (or intake structure). The evaluation of the potential seismic failure modes and determination of the ultimate seismic capacity of the structures is a complex undertaking which will require a large number of detailed calculations. As the first step in this evaluation, a number of potential modes of structural failure have been determined and are discussed. The report is principally directed towards seismically induced failure of structures. To some extent, modes involving soil foundation failures are discussed in so far as they affect the buildings. However, failure modes involving soil liquefaction, surface faulting, tsunamis, etc., are considered outside the scope of this evaluation

  11. Safety assessment for electricity generation failure accident of gas cooled nuclear power plant using system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2013-04-15

    The power production failure happens in the loss of coolant of the nuclear power plants (NPPs). The air ingress is a serious accident in gas cooled NPPs. The quantification of the study performed by the system dynamics (SD) method which is processed by the feedback algorithms. The Vensim software package is used for the simulation, which is performed by the Monte-Carlo method. Two kinds of considerations as the economic and safety properties are important in NPPs. The result shows the stability of the operation when the power can be decided. The maximum value of risk is the 11.77 in 43rd and the minimum value is 0.0 in several years. So, the success of the circulation of coolant is simulated by the dynamical values. (orig.)

  12. Impression creep behaviour of Mod. 9Cr-1Mo steel weld joints

    International Nuclear Information System (INIS)

    Ridhin Raj, V.R.; Kottda, Ravi Sankar; Kamaraj, M.; Maduraimuthu, V.M.; Vasudevan, M.

    2016-01-01

    P91 steel (9Cr-1Mo) steel is extensively used in power plants for super heater coils, headers and steam piping. The aim of the present work is to study the creep behaviour of different zones of A-TIG weld joint using impression creep technique and compare it with that of the TIG weld joint. P91 steel weld joints were made by A-TIG welding without using any filler material and multi-pass TIG welding is done using ER90S-B9 filler rods. Welds were subjected to post-weld heat treatment (PWHT). Impression creep tests were carried out at 650 °C on the base metal, weld metal and HAZ regions. Optical Microscope and TEM were used to correlate microstructures with observed creep rates. The FGHAZ showed significantly higher impression creep rate compared to that of the base metal and weld metal. Fine grain size and relatively coarser M 23 C 6 carbide particles are responsible for higher creep rate. The impression creep rate of A-TIG weld metal and coarse grain HAZ was found to be lower than that of base metal. This is attributed to the higher grain size in weld metal and coarse HAZ attributed to the higher grain size in weld metal and to the higher peak temperature observed during A-TIG welding. (author)

  13. Failure analysis of leakage on titanium tubes within heat exchangers in a nuclear power plant. Part II: Mechanical degradation

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Y.; Yang, Z.G. [Department of Materials Science, Fudan University, Shanghai (China); Yuan, J.Z. [Third Qinshan Nuclear Power Co. Ltd., Haiyan, Zhejiang Province (China)

    2012-01-15

    Serious failure incidents like clogging, quick thinning, and leakage frequently occurred on lots of titanium tubes of heat exchangers in a nuclear power plant in China. In the Part I of the whole failure analysis study with totally two parts, factors mainly involving three kinds of electrochemical corrosions were investigated, including galvanic corrosion, crevice corrosion, and hydrogen-assisted corrosion. In the current Part II, through microscopically analyzing the ruptures on the leaked tubes by scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS), another four causes dominantly lying in the aspect of mechanical degradation were determined - clogging, erosion, mechanical damaging, and fretting. Among them, the erosion effect was the primary one, thus the stresses it exerted on the tube wall were also supplementarily evaluated by finite element method (FEM). Based on the analysis results, the different degradation extents and morphologies by erosion on the tubes when they were clogged by different substances such as seashell, rubber debris, and sediments were compared, and relevant mechanisms were discussed. Finally, countermeasures were put forward as well. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  14. Application of nonhomogeneous Poisson process to reliability analysis of repairable systems of a nuclear power plant with rates of occurrence of failures time-dependent

    International Nuclear Information System (INIS)

    Saldanha, Pedro L.C.; Simone, Elaine A. de; Melo, Paulo Fernando F.F. e

    1996-01-01

    Aging is used to mean the continuous process which physical characteristics of a system, a structure or an equipment changes with time or use. Their effects are increases in failure probabilities of a system, a structure or an equipment, and their are calculated using time-dependent failure rate models. The purpose of this paper is to present an application of the nonhomogeneous Poisson process as a model to study rates of occurrence of failures when they are time-dependent. To this application, an analysis of reliability of service water pumps of a typical nuclear power plant is made, as long as the pumps are effectively repaired components. (author)

  15. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  16. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Rui; Seitisleam, F; Sandstroem, R [Swedish Institute for Metals Research, Stockholm (Sweden)

    1999-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  17. Experimental and modeling results of creep fatigue life of Inconel 617 and Haynes 230 at 850 C

    International Nuclear Information System (INIS)

    Chen, Xiang; Sokolov, Mikhail A.; Sham, Sam; Erdman, Donald L. III; Busby, Jeremy T.; Mo, Kun; Stubbins, James

    2013-01-01

    Creep fatigue testing of Ni-based superalloy Inconel 617 and Haynes 230 were conducted in the air at 850 C. Tests were performed with fully reversed axial strain control at a total strain range of 0.5%, 1.0% or 1.5% and hold time at maximum tensile strain for 3, 10 or 30 min. In addition, two creep fatigue life prediction methods, i.e. linear damage summation and frequency-modified tensile hysteresis energy modeling, were evaluated and compared with experimental results. Under all creep fatigue tests, Haynes 230 performed better than Inconel 617. Compared to the low cycle fatigue life, the cycles to failure for both materials decreased under creep fatigue test conditions. Longer hold time at maximum tensile strain would cause a further reduction in both material creep fatigue life. The linear damage summation could predict the creep fatigue life of Inconel 617 for limited test conditions, but considerably underestimated the creep fatigue life of Haynes 230. In contrast, frequency-modified tensile hysteresis energy modeling showed promising creep fatigue life prediction results for both materials.

  18. Creep Behavior in Interlaminar Shear of a SiC/SiC Ceramic Composite with a Self-healing Matrix

    Science.gov (United States)

    Ruggles-Wrenn, M. B.; Pope, M. T.

    2014-02-01

    Creep behavior in interlaminar shear of a non-oxide ceramic composite with a multilayered matrix was investigated at 1,200 °C in laboratory air and in steam environment. The composite was produced via chemical vapor infiltration (CVI). The composite had an oxidation inhibited matrix, which consisted of alternating layers of silicon carbide and boron carbide and was reinforced with laminated Hi-Nicalon™ fibers woven in a five-harness-satin weave. Fiber preforms had pyrolytic carbon fiber coating with boron carbide overlay applied. The interlaminar shear properties were measured. The creep behavior was examined for interlaminar shear stresses in the 16-22 MPa range. Primary and secondary creep regimes were observed in all tests conducted in air and in steam. In air and in steam, creep run-out defined as 100 h at creep stress was achieved at 16 MPa. Larger creep strains were accumulated in steam. However, creep strain rates and creep lifetimes were only moderately affected by the presence of steam. The retained properties of all specimens that achieved run-out were characterized. Composite microstructure, as well as damage and failure mechanisms were investigated.

  19. Application of non-monotonic logic to failure diagnosis of nuclear power plant

    International Nuclear Information System (INIS)

    Takahashi, M.; Kitamura, M.; Sugiyama, K.

    1989-01-01

    A prototype diagnosis system for nuclear power plants was developed based on Truth Maintenance systems: TMS and Dempster-Shafer probability theory. The purpose of this paper is to establish basic technique for more intelligent, man-computer cooperative diagnosis system. The developed system is capable of carrying out the diagnostic inference under the imperfect observation condition with the help of the proposed belief revision procedure with TMS and the systematic uncertainty treatment with Dempster-Shafer theory. The usefulness and potentiality of the present non-monotonic logic were demonstrated through simulation experiments

  20. Creep properties of phosphorus alloyed oxygen free copper under multiaxial stress state

    International Nuclear Information System (INIS)

    Rui Wu; Sandstroem, Rolf; Seitisleam, Facredin

    2009-10-01

    Phosphorus alloyed oxygen free copper (Cu-OFP) canisters are planned to be used for spent nuclear fuel in Sweden. The copper canisters will be subjected to creep under multiaxial stress states in the repository. Creep tests have therefore been carried out at 75 deg C using double notch specimens with notch acuities of 0.5, 2, 5, and 18.8, respectively. The creep lifetime for notched specimens is considerably longer than that for the smooth one at a given net section stress, indicating that the investigated Cu-OFP is notch insensitive (notch strengthening). The notch strengthening factor in time is, for instance, greater than 70 at 180 MPa for the bluntest notch (notch acuity = 0.5). The creep lifetime is notch acuity dependent. The sharper the notch, the longer the creep lifetime is. The creep deformation is to a significant extent concentrated to the region around the notches. Different deformation on the two notches is observed. Both axial and radial strains on the failed notch are several times larger than those on the unbroken one. Linear relation between the axial and the radial strains on the notches is found. Transgranular failure is predominant, independent of stress, rupture time, and notch acuity. Adjacent to fracture, elongated grains along the stress direction, separate pores and cavities are often visible. On the unbroken notch, fewer separate cavities and cracks are only seen intergranularly for the sharper notches (notch acuity > 2). To interpret the tests for the notched creep specimens, finite element computations have been performed. A fundamental model for primary and secondary creep without fitting parameters has been used as constitutive equation. The FEM-modelling could represent the creep strain versus time curves for the notched specimens in a satisfactory way. In these curves the strain on loading is included. From the FEM-computations a stationary creep stress could be assessed, which is close to the reference stress. For a given

  1. Magnetothermoelastic creep analysis of functionally graded cylinders

    International Nuclear Information System (INIS)

    Loghman, A.; Ghorbanpour Arani, A.; Amir, S.; Vajedi, A.

    2010-01-01

    This paper describes time-dependent creep stress redistribution analysis of a thick-walled FGM cylinder placed in uniform magnetic and temperature fields and subjected to an internal pressure. The material creep, magnetic and mechanical properties through the radial graded direction are assumed to obey the simple power law variation. Total strains are assumed to be the sum of elastic, thermal and creep strains. Creep strains are time, temperature and stress dependent. Using equations of equilibrium, stress-strain and strain-displacement a differential equation, containing creep strains, for displacement is obtained. Ignoring creep strains in this differential equation a closed form solution for the displacement and initial magnetothermoelastic stresses at zero time is presented. Initial magnetothermoelastic stresses are illustrated for different material properties. Using Prandtl-Reuss relation in conjunction with the above differential equation and the Norton's law for the material uniaxial creep constitutive model, the radial displacement rate is obtained and then the radial and circumferential creep stress rates are calculated. Creep stress rates are plotted against dimensionless radius for different material properties. Using creep stress rates, stress redistributions are calculated iteratively using magnetothermoelastic stresses as initial values for stress redistributions. It has been found that radial stress redistributions are not significant for different material properties, however major redistributions occur for circumferential and effective stresses.

  2. Neutron irradiation creep in stainless steel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Schuele, Wolfgang (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy)); Hausen, Hermann (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy))

    1994-09-01

    Irradiation creep elongations were measured in the HFR at Petten on AMCR steels, on 316 CE-reference steels, and on US-316 and US-PCA steels varying the irradiation temperature between 300 C and 500 C and the stress between 25 and 300 MPa. At the beginning of an irradiation a type of primary'' creep stage is observed for doses up to 3-5 dpa after which dose the secondary'' creep stage begins. The primary'' creep strain decreases in cold-worked steel materials with decreasing stress and decreasing irradiation temperature achieving also negative creep strains depending also on the pre-treatment of the materials. These primary'' creep strains are mainly attributed to volume changes due to the formation of radiation-induced phases, e.g. to the formation of [alpha]-ferrite below about 400 C and of carbides below about 700 C, and not to irradiation creep. The secondary'' creep stage is found for doses larger than 3 to 5 dpa and is attributed mainly to irradiation creep. The irradiation creep rate is almost independent of the irradiation temperature (Q[sub irr]=0.132 eV) and linearly dependent on the stress. The total creep elongations normalized to about 8 dpa are equal for almost every type of steel irradiated in the HFR at Petten or in ORR or in EBR II. The negative creep elongations are more pronounced in PCA- and in AMCR-steels and for this reason the total creep elongation is slightly smaller at 8 dpa for these two steels than for the other steels. ((orig.))

  3. Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Wang, Lumin; Was, Gary

    2010-01-01

    Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

  4. The influence of the crust layer on RPV structural failure under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Mao, Jianfeng, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Li, Xiangqing [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Bao, Shiyi [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, Lijia [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Gao, Zengliang [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2017-05-15

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  5. The influence of the crust layer on RPV structural failure under severe accident condition

    International Nuclear Information System (INIS)

    Mao, Jianfeng; Li, Xiangqing; Bao, Shiyi; Luo, Lijia; Gao, Zengliang

    2017-01-01

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  6. Operation of the main feedwater system turbopump following plant trip with total failure of the auxiliary feedwater system

    International Nuclear Information System (INIS)

    Lucas Alvaro, A.M. de; Rosa Martinez, B. de la; Alcaide, F.; Toledano Camara, C.

    1993-01-01

    The Auxiliary Feedwater System (AF) is a safeguard system which has been designed to supply feedwater to the steam generators, cool the primary system and remove decay heat from the reactor when the main feedwater pumps fail due to loss of power or any other reason. Thus, when plant trip occurs, the AF system pumps start up automatically, allowing removal of decay heat from the reactor. However, even though this system (2 motor-driven pumps and 1 turbopump) is highly reliable, injection of water to the steam generators must be ensured when it fails completely. To do this, if plant trip has not been caused by loss of off site power or failure of the Main Feedwater System (FW) turbopumps, one of these turbopumps can be used to achieve removal of decay heat. Since a large amount of steam is consumed by these turbopumps, an analysis has been performed to determine whether one of these pumps can be used and what actions are necessary to inject water into the steam generators. Results show that, for the case in question, a FW turbopump can be used to remove decay heat from the reactor. (author)

  7. Construction of in-situ creep strain test facility for the SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel (12CrMoVW) is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92(12Cr-2W) are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.

  8. The estimation ecological risk for ground ecosystems in case of nuclear power plant failures

    International Nuclear Information System (INIS)

    Kremlenkov, D.Yu.; Kremlenkov, M.Yu.

    2004-01-01

    The results of probabilistic estimation of the damage for forest and agricultural ecosystems connected with cesium-137 and strontium-90 release during hypothetical accidents at NPPs are analyzed. The concept of radioecological risk including application of the models for radioactivity transport in atmosphere and calculation of of absorbed doses in ecosystem critical groups is used for the analysis. It is proved that the probable ecological damage expressed in terms of ecosystem destruction area depends on the scale of accidental radioisotope releases, meteorological conditions and radiation resistance of critical groups in plant associations. The conclusion is made that ecological risks expressed in the form of probable area of ecosystem destruction in the zone where dose loads lay in the range from minimal ecologically significant limit up to ecologically significant limit amount to 4-9% for conifers contaminated with cesium-137 and to 2-4% for agricultural crops contaminated with strontium-90 [ru

  9. Failure analysis of the stud bolt of a canned motor pump of heavy water plant, Talcher

    International Nuclear Information System (INIS)

    Kumar, Sunil; Sethumadhavan, V.; Sah, D.N.; Sivaramakrishnan, K.S.; Kain, Vivekanand; Gadiyar, H.S.

    1990-01-01

    Detailed investigations have been carried out on the failed stud bolts (made of DIN 1.4021 X 20Cr 13) of a canned motor pump of Heavy Water Plant, Talcher, using metallographic, microhardness testing, scanning electron microscopy and electron probe micro-analysis(EPMA) techniques. The studs had failed in a brittle manner in the mid-length portion. The origin of the fracture has been identified to be a corroded region on the stud surface. Branching cracks propagating through silicon rich inclusions have been noted. Two types of inclusions, one containing Mn and S and other containing S, Mn and O 2 have been found in the material. Clusters of large and small inclusions of the above types have been found near the origin of the fracture. It has been concluded that the fracture was caused by corrosion fatigue, initiating at the cluster of inclusions present on the surface of the stud bolts. (author). 7 refs., 5 figs

  10. The independence of irradiation creep in austenitic alloys of displacement rate and helium to dpa ratio

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States); Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    The majority of high fluence data on the void swelling and irradiation creep of austenitic steels were generated at relatively high displacement rates and relatively low helium/dpa levels that are not characteristic of the conditions anticipated in ITER and other anticipated fusion environments. After reanalyzing the available data, this paper shows that irradiation creep is not directly sensitive to either the helium/dpa ratio or the displacement rate, other than through their possible influence on void swelling, since one component of the irradiation creep rate varies with no correlation to the instantaneous swelling rate. Until recently, however, the non-swelling-related creep component was also thought to exhibit its own strong dependence on displacement rate, increasing at lower fluxes. This perception originally arose from the work of Lewthwaite and Mosedale at temperatures in the 270-350{degrees}C range. More recently this perception was thought to extend to higher irradiation temperatures. It now appears, however, that this interpretation is incorrect, and in fact the steady-state value of the non-swelling component of irradiation creep is actually insensitive to displacement rate. The perceived flux dependence appears to arise from a failure to properly interpret the impact of the transient regime of irradiation creep.

  11. Stress Calculation of a TRISO Coated Particle Fuel by Using a Poisson's Ratio in Creep Condition

    International Nuclear Information System (INIS)

    Cho, Moon-Sung; Kim, Y. M.; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.; Kim, W. K.

    2007-01-01

    KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) project since 2004, has been developing a performance analysis code for the TRISO coated particle fuel named COPA (COated Particle fuel Analysis). COPA predicts temperatures, stresses, a fission gas release and failure probabilities of a coated particle fuel in normal operating conditions. KAERI, on the other hand, is developing an ABAQUS based finite element(FE) model to cover the non-linear behaviors of a coated particle fuel such as cracking or debonding of the TRISO coating layers. Using the ABAQUS based FE model, verification calculations were carried out for the IAEA CRP-6 benchmark problems involving creep, swelling, and pressure. However, in this model the Poisson's ratio for elastic solution was used for creep strain calculation. In this study, an improvement is made for the ABAQUS based finite element model by using the Poisson's ratio in creep condition for the calculation of the creep strain rate. As a direct input of the coefficient in a creep condition is impossible, a user subroutine for the ABAQUS solution is prepared in FORTRAN for use in the calculations of the creep strain of the coating layers in the radial and hoop directions of the spherical fuel. This paper shows the calculation results of a TRISO coated particle fuel subject to an irradiation condition assumed as in the Miller's publication in comparison with the results obtained from the old FE model used in the CRP-6 benchmark calculations

  12. Detection of creep damage in a nickel base superalloy using NDE techniques

    International Nuclear Information System (INIS)

    Carreon, H.; Mora, B.; Barrera, G.

    2009-10-01

    Due to elevated temperatures, excessive stresses and severed corrosion conditions, turbine engine components are subject to creep processes that limit the components life such as a turbine bucket. The failure mechanism of a turbine bucket is related primarily to creep and corrosion and secondarily to thermal fatigue. As a result, it is desirable to assess the current conditions of such turbine component. This study uses the eddy current nondestructive evaluation technique in an effort to monitor the creep damage in a nickel base super-alloy, turbine bucket after service. The experimental results show an important electrical conductivity variation in eddy current images on the creep damage zone of nickel base super-alloy samples cut from a turbine bucket. Thermoelectric power measurements were also conducted in order to obtain a direct correlation between the presence of material changes due to creep damage and the electrical conductivity measurements. This research work shows an alternative non-destructive method in order to detect creep damage in a nickel base super-alloy turbine bucket. (Author)

  13. Evaluation of creep-fatigue/ environment interaction in Ni-base wrought alloys for HTGR application

    International Nuclear Information System (INIS)

    Hattori, Hiroshi; Kitagawa, Masaki; Ohtomo, Akira

    1986-01-01

    High Temperature Gas-cooled Reactor (HTGR) systems should be designed based on the high temperature structural strength design procedures. On the development of design code, the determination of failure criteria under cyclic loading and severe environments is one of the most important items. By using the previous experimental data for Ni-base wrought alloys, Inconel 617 and Hastelloy XR, several evaluation methods for creep-fatigue interaction were examined for their capability to predict their cyclic loading behavior for HTGR application. At first, the strainrange partitioning method, the frequency modified damage function and the linear damage summation rule were discussed. However, these methods were not satisfactory with the above experimental results. Thus, in this paper, a new fracture criterion, which is a modification of the linear damage summation rule, is proposed based on the experimental data. In this criterion, fracture is considered to occur when the sum of the fatigue damage, which is the function of the applied cyclic strain magnitude, and the modified creep damage, which is the function of the applied cyclic stress magnitude (determined as time devided by cyclic creep rupture time reflecting difference of creep damages by tensile creep and compressive creep), reaches a constant value. This criterion was successfully applied to the life prediction of materials at HTGR temperatures. (author)

  14. Characterization of load dependent creep behavior in medically relevant absorbable polymers.

    Science.gov (United States)

    Dreher, Maureen L; Nagaraja, Srinidhi; Bui, Hieu; Hong, Danny

    2014-01-01

    While synthetic absorbable polymers have a substantial history of use in medical devices, their use is expanding and becoming more prevalent for devices where long term loading and structural support is required. In addition, there is evidence that current absorbable medical devices may experience permanent deformations, warping (out of plane twisting), and geometric changes in vivo. For clinical indications with long term loading or structural support requirements, understanding the material's viscoelastic properties becomes increasingly important whereas these properties have not been used historically as preclinical indications of performance or design considerations. In this study we measured the static creep, creep recovery and cyclic creep responses of common medically relevant absorbable materials (i.e., poly(l-lactide, PLLA) and poly(l-co-glycolide, PLGA) over a range of physiologically relevant loading magnitudes. The results indicate that both PLLA and PLGA exhibit creep behavior and failure at loads significantly less than the yield or ultimate properties of the material and that significant material specific responses to loading exist. In addition, we identified a strong correlation between the extent of creep in the material and its crystallinity. Results of the study provide new information on the creep behavior of PLLA and PLGA and support the use of viscoelastic properties of absorbable polymers as part of the material selection process. © 2013 Published by Elsevier Ltd.

  15. A study on creep properties of laminated rubber bearings. Pt. 1. Creep properties and numerical simulations of thick rubber bearings

    International Nuclear Information System (INIS)

    Matsuda, Akihiro; Yabana, Shuichi

    2000-01-01

    In this report, to evaluate creep properties and effects of creep deformation on mechanical properties of thick rubber bearings for three-dimensional isolation system, we show results of compression creep test for rubber bearings of various rubber materials and shapes and development of numerical simulation method. Creep properties of thick rubber bearings were obtained from compression creep tests. The creep strain shows steady creep that have logarithmic relationships between strain and time and accelerated creep that have linear relationships. We make numerical model of a rubber material with nonlinear viscoelastic constitutional equations. Mechanical properties after creep loading test are simulated with enough accuracy. (author)

  16. A creep rupture model accounting for cavitation at sliding grain boundaries

    NARCIS (Netherlands)

    Giessen, Erik van der; Tvergaard, Viggo

    1991-01-01

    An axisymmetric cell model analysis is used to study creep failure by grain boundary cavitation at facets normal to the maximum principal tensile stress, taking into account the influence of cavitation and sliding at adjacent inclined grain boundaries. It is found that the interaction between the

  17. Failure of Aedes albopictus to overwinter following introduction and seasonal establishment at a tire recycling plant in the northeastern USA.

    Science.gov (United States)

    Andreadis, Theodore G

    2009-03-01

    In July 2006, an introduction of the Asian tiger mosquito, Aedes albopictus, was documented for the first time at a commercial tire recycling plant in northeastern Connecticut, USA. The introduction likely occurred via transport of infested tires originating from northern New Jersey or metropolitan New York City. Efforts were made to determine seasonal establishment and overwintering success by assessing adult biting and oviposition activity in the surrounding woodlands. The first adult female was collected in a CO2-baited Mosquito Magnet Liberty trap within the confines of the tire plant during the week of July 28. Additional females were collected intermittently thereafter through October 16. Host-seeking female Ae. albopictus attempting to alight on human subjects and larvae hatching from eggs collected in ovitraps placed in the woodlands surrounding the tire plant were detected weekly from August 21 through October 2, denoting seasonal establishment in the adjoining woodlands. However, no larvae of Ae. albopictus were recovered from eggs collected in ovitraps that were placed in the surrounding woodlands or in traps placed 1.0-1.6 km away, nor were any host-seeking females detected by human subjects the following season (July to October 2007), indicating that the species did not survive winter conditions to enable successful colonization. The failure of Ae. albopictus to overwinter and establish itself in the forested woodlands following several weeks of seasonal breeding and oviposition during the summer and early fall were most likely due to winter egg mortality, interspecific competition from Aedes triseriatus and Aedes japonicus, and/or other ecological barriers. Permanent establishment of Ae. albopictus in New England is unlikely despite the recurring importation of infested used tires into recycling facilities. However, continued monitoring of such facilities for potential reinvasion is warranted especially in urban/suburban environs where global warming

  18. Room temperature creep in metals and alloys

    Energy Technology Data Exchange (ETDEWEB)

    Deibler, Lisa Anne [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Characterization and Performance

    2014-09-01

    Time dependent deformation in the form of creep and stress relaxation is not often considered a factor when designing structural alloy parts for use at room temperature. However, creep and stress relaxation do occur at room temperature (0.09-0.21 Tm for alloys in this report) in structural alloys. This report will summarize the available literature on room temperature creep, present creep data collected on various structural alloys, and finally compare the acquired data to equations used in the literature to model creep behavior. Based on evidence from the literature and fitting of various equations, the mechanism which causes room temperature creep is found to include dislocation generation as well as exhaustion.

  19. Creep of parylene-C film

    KAUST Repository

    Lin, Jeffrey Chun-Hui

    2011-06-01

    The glass transition temperature of as-deposited parylene-C is first measured to be 50°C with a ramping-temperature-dependent modulus experiment. The creep behavior of parylene-C film in the primary and secondary creep region is then investigated below and above this glass transition temperature using a dynamic mechanical analysis (DMA) machine Q800 from TA instruments at 8 different temperatures: 10, 25, 40, 60, 80, 100, 120 and 150°C. The Burger\\'s model, which is the combined Maxwell model and Kelvin-Voigt model, fits well with our primary and secondary creep data. Accordingly, the results show that there\\'s little or no creep below the glass transition temperature. Above the glass transition temperature, the primary creep and creep rate increases with the temperature, with a retardation time constant around 6 minutes. © 2011 IEEE.

  20. Creep Strength of Discontinuous Fibre Composites

    DEFF Research Database (Denmark)

    Pedersen, Ole Bøcker

    1974-01-01

    relation between stress and strain rate. Expressions for the interface stress, the creep velocity profile adjacent to the fibres and the creep strength of the composite are derived. Previous results for the creep strength, sc = aVfs0 ( \\frac[( Î )\\dot] [( Î )\\dot] 0 )1/nr1 + 1/n c=Vf001n1+1n in which[( Î...... )\\dot] is the composite creep rate,V f is the fibre volume fraction,sgr 0,epsi 0 andn are the constants in the matrix creep law. The creep strength coefficient agr is found to be very weakly dependent onV f and practically independent ofn whenn is greater than about 6....

  1. Flexural creep behaviour of jute polypropylene composites

    Science.gov (United States)

    Chandekar, Harichandra; Chaudhari, Vikas

    2016-09-01

    Present study is about the flexural creep behaviour of jute fabric reinforced polypropylene (Jute-PP) composites. The PP sheet and alkali treated jute fabric is stacked alternately and hot pressed in compression molding machine to get Jute-PP composite laminate. The flexural creep study is carried out on dynamic mechanical analyzer. The creep behaviour of the composite is modeled using four-parameter Burgers model. Short-term accelerated creep testing is conducted which is later used to predict long term creep behaviour. The feasibility of the construction of a master curve using the time-temperature superposition (TTS) principle to predict long term creep behavior of unreinforced PP and Jute-PP composite is investigated.

  2. High temperature high vacuum creep testing facilities

    International Nuclear Information System (INIS)

    Matta, M.K.

    1985-01-01

    Creep is the term used to describe time-dependent plastic flow of metals under conditions of constant load or stress at constant high temperature. Creep has an important considerations for materials operating under stresses at high temperatures for long time such as cladding materials, pressure vessels, steam turbines, boilers,...etc. These two creep machines measures the creep of materials and alloys at high temperature under high vacuum at constant stress. By the two chart recorders attached to the system one could register time and temperature versus strain during the test . This report consists of three chapters, chapter I is the introduction, chapter II is the technical description of the creep machines while chapter III discuss some experimental data on the creep behaviour. Of helium implanted stainless steel. 13 fig., 3 tab

  3. Experimental investigation of creep behavior of reactor vessel lower head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1998-03-01

    The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling

  4. Creep-Data Analysis of Alloy 617 for High Temperature Reactor Intermediate Heat Exchanger

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Ryu, Woo Seog; Kim, Yong Wan; Yin, Song Nan

    2006-01-01

    The design of the metallic components such as hot gas ducts, intermediate heat exchanger (IHX) tube, and steam reformer tubes of very high temperature reactor (VHTR) is principally determined by the creep properties, because an integrity of the components should be preserved during a design life over 30 year life at the maximum operating temperature up to 1000 .deg. C. For designing the time dependent creep of the components, a material database is needed, and an allowable design stress at temperature should be determined by using the material database. Alloy 617, a nicked based superalloy with chromium, molybdenum and cobalt additions, is considered as a prospective candidate material for the IHX because it has the highest design temperature. The alloy 617 is approved to 982 .deg. C (1800 .deg. F) and other alloys approved to 898 .deg. C (1650 .deg. C), such as alloy 556, alloy 230, alloy HX, alloy 800. Also, the alloy 617 exhibits the highest level of creep strength at high temperatures. Therefore, it is needed to collect the creep data for the alloy 617 and the creep-rupture life at the given conditions of temperature and stress should be predicted for the IHX construction. In this paper, the creep data for the alloy 617 was collected through literature survey. Using the collected data, the creep life for the alloy 617 was predicted based on the Larson-Miller parameter. Creep master curves with standard deviations were presented for a safety design, and failure probability for the alloy 617 was obtained with a time coefficient

  5. Light water reactor lower head failure analysis

    International Nuclear Information System (INIS)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response

  6. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  7. Progress Report on Long Hold Time Creep Fatigue of Alloy 617 at 850°C

    International Nuclear Information System (INIS)

    Carroll, Laura Jill

    2015-01-01

    Alloy 617 is the leading candidate material for an intermediate heat exchanger for the very high temperature reactor. To evaluate the behavior of this material in the expected service conditions, strain-controlled cyclic tests that include long hold times up to 240 minutes at maximum tensile strain were conducted at 850°C. In terms of the total number of cycles to failure, the fatigue resistance decreased when a hold time was added at peak tensile strain. Increases in the tensile hold duration degraded the creep-fatigue resistance, at least to the investigated strain controlled hold time of up to 60 minutes at the 0.3% strain range and 240 minutes at the 1.0% strain range. The creep-fatigue deformation mode is considered relative to the lack of saturation, or continually decreasing number of cycles to failure with increasing hold times. Additionally, preliminary values from the 850°C creep-fatigue data are calculated for the creep-fatigue damage diagram and have higher values of creep damage than those from tests at 950°C.

  8. Creep and rupture of an ODS alloy with high stress rupture ductility. [Oxide Dispersion Strengthened

    Science.gov (United States)

    Mcalarney, M. E.; Arsons, R. M.; Howson, T. E.; Tien, J. K.; Baranow, S.

    1982-01-01

    The creep and stress rupture properties of an oxide (Y2O3) dispersion strengthened nickel-base alloy, which also is strengthened by gamma-prime precipitates, was studied at 760 and 1093 C. At both temperatures, the alloy YDNiCrAl exhibits unusually high stress rupture ductility as measured by both elongation and reduction in area. Failure was transgranular, and different modes of failure were observed including crystallographic fracture at intermediate temperatures and tearing or necking almost to a chisel point at higher temperatures. While the rupture ductility was high, the creep strength of the alloy was low relative to conventional gamma prime strengthened superalloys in the intermediate temperature range and to ODS alloys in the higher temperature range. These findings are discussed with respect to the alloy composition; the strengthening oxide phases, which are inhomogeneously dispersed; the grain morphology, which is coarse and elongated and exhibits many included grains; and the second phase inclusion particles occurring at grain boundaries and in the matrix. The creep properties, in particular the high stress dependencies and high creep activation energies measured, are discussed with respect to the resisting stress model of creep in particle strengthened alloys.

  9. Tensile creep of beta phase zircaloy-2

    International Nuclear Information System (INIS)

    Burton, B.; Reynolds, G.L.; Barnes, J.P.

    1977-08-01

    The tensile creep and creep rupture properties of beta-phase zircaloy-2 are studied under vacuum in the temperature and stress range 1300-1550 K and 0.5-2 MN/m 2 . The new results are compared with previously reported uniaxial and biaxial data. A small but systematic difference is noted between the uniaxial and biaxial creep data and reasons for this discrepancy are discussed. (author)

  10. Viscoelastic creep of high-temperature concrete

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Marchertas, A.H.; Bazant, Z.P.

    1985-01-01

    Presented in this report is the analytical model for analysis of high temperature creep response of concrete. The creep law used is linear (viscoelastic), the temperature and moisture effects on the creep rate and also aging are included. Both constant and transient temperature as well as constant and transient moisture conditions are considered. Examples are presented to correlate experimental data with parameters of the analytical model by the use of a finite element scheme

  11. Studies on the causes of failures in titanium tube condensers of nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Uh Chul; Kim, Joung Soo; Jang, Soon Sik; Park, Su Hoon; Chung, Han Sub; Hwang, Seong Sik

    1990-03-01

    All condenser tubes in domestic nuclear power plants using sea water as condenser coolant are Ti material, while water boxes and tubesheets are, respectively, made of carbon steel and Cu-alloys. In this case, to prevent galvanic corrosion occuring between these materials, cathodic protection by an impressed direct current method is being used. Under inappropriate conditions, however, over-protection can generate hydrogen gas on Ti-tube surfaces, which may cause hydrogen embrittlement to the Ti tubes. Therefore, proper conditions for appropriate cathodic protection must be obtained through experiments. Galvanic corrosion between these three different materials was investigated, first. When SA 285 is coupled with Cu or Ti in sea water, under the same surface area, galvanic corrosion rate of SA 285 significantly increases (For example, corrosion rate increment, 2 x 10 -1 → 8 ma for SA 285 coupled with Cu). It was also affected by Cu more than by Ti. In addition, as the surface area ratio of SA 285 to Ti decreases, the corrosion rate of SA 285 increases. This is similar to SA 285-Cu couple. In this case, corrosion rate of SA 285 is higher than that of SA 285-Ti system. For Cu-Ti coulpe, as the surface area ratio of Cu to Ti increases, current density increases slightly up to 1.5 μA/cm 2 , which is very low. (author)

  12. Creep-Fatigue Life Design with Various Stress and Temperature Conditions on the Basis of Lethargy Coefficient

    International Nuclear Information System (INIS)

    Park, Jung Eun; Yang, Sung Mo; Han, Jae Hee; Yu, Hyo Sun

    2011-01-01

    High temperature and stress are encounted in power plants and vehicle engines. Therefore, determination of the creep-fatigue life of a material is necessary prior to fabricating equipment. In this study, life design was determined on the basis of the lethargy coefficient for different temperatures, stress and rupture times. SP-Creep test data was compared with computed data. The SP-Creep test was performed to obtain the rupture time for X20CrMoV121 steel. The integration life equation was considered for three cases with various load, temperature and load-temperature. First, the lethargy coefficient was calculated by using the obtained rupture stress and the rupture time that were determined by carrying out the SP-Creep test. Next, life was predicted on the basis of the temperature condition. Finally, it was observed that life decreases considerably due to the coupling effect that results when fatigue and creep occur simultaneously

  13. Influence of stress on creep deformation properties of 9-12Cr ferritic creep resistant steels

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, K.; Sawada, K.; Kushima, H. [National Institute for Materials Science (Japan)

    2008-07-01

    Creep deformation property of 9-12Cr ferritic creep resistant steels was investigated. With decrease in stress, a magnitude of creep strain at the onset of accelerating creep stage decreased from about 2% in the short-term to less than 1% in the longterm. A time to 1% total strain was observed in the transient creep stage in the short term regime, however, it shifted to the accelerating creep stage in the long-term regime. Life fraction of the times to 1% creep strain and 1% total strain tended to increase with decrease in stress. Difference in stress dependence of the minimum creep rate was observed in the high- and low-stress regimes with a boundary condition of 50% of 0.2% offset yield stress. Stress dependence of the minimum creep rate in the high stress regime was equivalent to a strain rate dependence of the flow stress evaluated by tensile test, and a magnitude of stress exponent, n, in the high stress regime decreased with increase in temperature from 20 at 550 C to 10 at 700 C. On the other hand, n value in the low stress regime was about 5, and creep deformation in the low stress regime was considered to be controlled by dislocation climb. Creep rupture life was accurately predicted by a region splitting method by considering a change in stress dependence of creep deformation. (orig.)

  14. Surveillance and monitoring experiences of chemical water-steam cycle in combined-cycle power plants (CCCs in Spanish acronyms) for early diagnosis of failures

    International Nuclear Information System (INIS)

    Santaolalla losada, E.

    2010-01-01

    In the present work is made a brief comparative review of the treatments effected in water of the cycle, their purpose, types, monitoring and tracking. Moreover, are described the most important failure mechanisms that can be related to the chemistry of cycle with the aim to establish monitoring guidelines to increase the reliability of the power plant and enable an early diagnosis. (Author)

  15. Final Report for Project 13-4791: New Mechanistic Models of Creep-Fatigue Crack Growth Interactions for Advanced High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Kruzic, Jamie J [Oregon State Univ., Corvallis, OR (United States); Siegmund, Thomas [Purdue Univ., West Lafayette, IN (United States); Tomar, Vikas [Purdue Univ., West Lafayette, IN (United States)

    2018-03-20

    This project developed and validated a novel, multi-scale, mechanism-based model to quantitatively predict creep-fatigue crack growth and failure for Ni-based Alloy 617 at 800°C. Alloy 617 is a target material for intermediate heat exchangers in Generation IV very high temperature reactor designs, and it is envisioned that this model will aid in the design of safe, long lasting nuclear power plants. The technical effectiveness of the model was shown by demonstrating that experimentally observed crack growth rates can be predicted under both steady state and overload crack growth conditions. Feasibility was considered by incorporating our model into a commercially available finite element method code, ABAQUS, that is commonly used by design engineers. While the focus of the project was specifically on an alloy targeted for Generation IV nuclear reactors, the benefits to the public are expected to be wide reaching. Indeed, creep-fatigue failure is a design consideration for a wide range of high temperature mechanical systems that rely on Ni-based alloys, including industrial gas power turbines, advanced ultra-super critical steam turbines, and aerospace turbine engines. It is envisioned that this new model can be adapted to a wide range of engineering applications.

  16. Multiaxial creep behavior of 304 stainless steel

    International Nuclear Information System (INIS)

    Findley, W.N.; Mark, R.

    1975-07-01

    Tests in combined tension-torsion, pure tension and pure torsion, were conducted at elevated temperature (about 1100 0 F). Most of these tests were repeats of previous experiments where friction in the extensometer caused anomalous creep behavior. The existence of a creep surface at 12.5 ksi effective Mises stress was explored. Work on a compression creep apparatus continued. Creep and recovery data were fitted to the equation epsilon/sub ij/ = epsilon 0 /sub ij/ + e + /sub ij/t/sup n//sub ij/ by means of a least squares method. (5 tables, 10 fig) (auth)

  17. Low stress creep behaviour of zirconium

    International Nuclear Information System (INIS)

    Prasad, N.

    1989-01-01

    Creep behaviour of alpha zirconium of grain size varying between 16 and 55 μm has been investigated in the temperature range 813 to 1003K at stresses upto 5.5 MNm -2 using high sensitive spring specimen geometry. Creep experiments on specimens of 50 μm grain size revealed a transition from lattice diffusion controlled viscous creep at temperatures greater than 940K to grain boundary diffusion controlled viscous creep at lower temperatures. Tests conducted on either side of the transition suggest the dominance of Nabarro-Herring and Coble creep processes respectively. Evidence for power-law creep has been observed in practically all the creep tests. Based on the experimental data obtained in the present study and those recently reported by Novotny et al (1985), Langdon creep mechanism maps have bee n constructed at 873 and 973K. With the help of these maps for zirconium and those published for titanium the low stress creep behaviour of zirconium and titanium are compared. (author). 22 refs., 11 figs., 3 tabs

  18. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Reactor Pressure Vessels

    International Nuclear Information System (INIS)

    Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

    2013-01-01

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  19. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Potirniche, Gabriel [Univ. of Idaho, Moscow, ID (United States); Barlow, Fred D. [Univ. of Idaho, Moscow, ID (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Rink, Karl [Univ. of Idaho, Moscow, ID (United States)

    2013-11-26

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  20. Risk assessment of Giardia from a full scale MBR sewage treatment plant caused by membrane integrity failure.

    Science.gov (United States)

    Zhang, Yu; Chen, Zhimin; An, Wei; Xiao, Shumin; Yuan, Hongying; Zhang, Dongqing; Yang, Min

    2015-04-01

    Membrane bioreactors (MBR) are highly efficient at intercepting particles and microbes and have become an important technology for wastewater reclamation. However, many pathogens can accumulate in activated sludge due to the long residence time usually adopted in MBR, and thus may pose health risks when membrane integrity problems occur. This study presents data from a survey on the occurrence of water-borne Giardia pathogens in reclaimed water from a full-scale wastewater treatment plant with MBR experiencing membrane integrity failure, and assessed the associated risk for green space irrigation. Due to membrane integrity failure, the MBR effluent turbidity varied between 0.23 and 1.90 NTU over a period of eight months. Though this turbidity level still met reclaimed water quality standards (≤5 NTU), Giardia were detected at concentrations of 0.3 to 95 cysts/10 L, with a close correlation between effluent turbidity and Giardia concentration. All β-giardin gene sequences of Giardia in the WWTP influents were genotyped as Assemblages A and B, both of which are known to infect humans. An exponential dose-response model was applied to assess the risk of infection by Giardia. The risk in the MBR effluent with chlorination was 9.83×10(-3), higher than the acceptable annual risk of 1.0×10(-4). This study suggested that membrane integrity is very important for keeping a low pathogen level, and multiple barriers are needed to ensure the biological safety of MBR effluent. Copyright © 2015. Published by Elsevier B.V.

  1. Development of an accelerated creep testing procedure for geosynthetics.

    Science.gov (United States)

    1997-09-01

    The report presents a procedure for predicting creep strains of geosynthetics using creep tests at elevated temperatures. Creep testing equipment was constructed and tests were performed on two types of geosynthetics: High Density Polyethylene (HDPE)...

  2. Study on the creep constitutive equation of Hastelloy X, (1)

    International Nuclear Information System (INIS)

    Hada, Kazuhiko; Mutoh, Yasushi

    1983-01-01

    A creep constitutive equation of Hastelloy X was obtained from available experimental data. A sensitivity analysis of this creep constitutive equation was carried out. As the result, the following were revealed: (i) Variations in creep behavior with creep constitutive equation are not small. (ii) In a simpler stress change pattern, variations in creep behavior are similar to those in the corresponding fundamental creep characteristics (creep strain curve, stress relaxation curve, etc.). (iii) Cumulative creep damage estimated in accordance with ASME Boiler and Pressure Vessel Code Case N-47 from a stress history predicted by ''the standard creep constitutive equation'' which predicts the average behavior of creep strain curve data is not thought to be on the safe side on account of uncertainties in creep damage caused by variations in creep strain curve. (author)

  3. Model for transient creep of southeastern New Mexico rock salt

    International Nuclear Information System (INIS)

    Herrmann, W.; Wawersik, W.R.; Lauson, H.S.

    1980-11-01

    In a previous analysis, existing experimental data pertaining to creep tests on rock salt from the Salado formation of S.E. New Mexico were fitted to an exponential transient creep law. While very early time portions of creep strain histories were not fitted very well for tests at low temperatures and stresses, initial creep rates in particular generally being underestimated, the exponential creep law has the property that the transient creep strain approaches a finite limit with time, and is therefore desirable from a creep modelling point of view. In this report, an analysis of transient creep is made. It is found that exponential transient creep can be related to steady-state creep through a universal creep curve. The resultant description is convenient for creep analyses where very early time behavior is not important

  4. Prediction of thermoplastic failure of a reactor pressure vessel under a postulated core melt accident

    International Nuclear Information System (INIS)

    Duijvestijn, G.; Birchley, J.; Reichlin, K.

    1997-01-01

    This paper presents the lower head failure calculations performed for a postulated accident scenario in a commercial nuclear power plant. A postulated one inch break in the primary coolant circuit leads to dryout and subsequent meltdown of the core. The reference plant is a pressurized water reactor without penetrations in the reactor vessel lower head. The molten core material accumulates in the lower head, eventually causing failure of the vessel. The analysis investigates flow conditions in the melt pool, temperature evolution in the reactor vessel wall, and structure mechanical evaluation of the vessel under strong thermal loads and a range of internal pressures. The calculations were performed using the ADINA finite element codes. The analysis focusses on the failure processes, time and mode of failure. The most likely mode of failure at low pressure is global rupture due to gradual accumulation of creep strain over a large part of the heated area. In contrast, thermoplasticity becomes important at high pressure or following a pressure spike and can lead to earlier local failure. In situations in which part of the heat load is concentrated over a small area, resulting in a hot spot, local failure occurs, but not until the temperatures are close to the melting point. At low pressure, in particular, the hot spot area remains intact until the structure is molten across more than half of the thickness. (author) 14 figs., 16 refs

  5. Premature failure of dissimilar metal weld joint at intermediate temperature superheater tube

    Directory of Open Access Journals (Sweden)

    Mohammed Al Hajri

    2015-04-01

    Full Text Available Dissimilar metal weld (DMW joint between alloyed steel (AS and stainless steel (SS failed at one of intermediate temperature superheater (ITSH tube in steam/power generation plant boiler. The premature failure was detected after a relatively short time of operation (8 years where the crack propagated circumferentially from AS side through the ITSH tube. Apart from physical examination, microstructural studies based on optical microscopy, SEM and EDX analysis were performed. The results of the investigation point out the limitation of Carbides precipitation at the alloyed steel/welding interface. This is synonym of creep stage I involvement in the failure of ITSH. Improper post-welding operation and bending moment are considered as root causes of the premature failure.

  6. Creep properties of a thermally grown alumina

    Energy Technology Data Exchange (ETDEWEB)

    Kang, K.J. [Department of Mechanical Engineering, Chonnam National University, Kwangju 500-757 (Korea, Republic of)], E-mail: kjkang@chonnam.ac.kr; Mercer, C. [Materials Department, University of California, Santa Barbara, CA 93106-5050 (United States)

    2008-04-15

    A unique test system has been developed to measure creep properties of actual thermally grown oxides (TGO) formed on a metal foil. The thickness of TGO, load and displacement can be monitored in situ at high temperature. Two batches of FeCrAlY alloys which differ from each other in contents of yttrium and titanium were selected as the {alpha}-Al{sub 2}O{sub 3} TGO forming materials. The creep tests were performed on {alpha}-Al{sub 2}O{sub 3} of thickness 1-4 {mu}m, thermally grown at 1200 deg. C in air. The strength of the substrate was found to be negligible, provided that the TGO and substrate thickness satisfy: h{sub TGO} > 1 {mu}m and H{sub sub} {<=} 400 {mu}m. The steady-state creep results for all four TGO thicknesses obtained on batch I reside within a narrow range, characterized by a parabolic creep relation. It is nevertheless clear that the steady-state creep rates vary with TGO thickness: decreasing as the thickness increases. For batch II, the steady-state creep rates are higher and now influenced more significantly by TGO thickness. In comparison with previous results of the creep properties for bulk polycrystalline {alpha}-Al{sub 2}O{sub 3} at a grain size of {approx}2 {mu}m, the creep rates for the TGO were apparently higher, but both were significantly affected by yttrium content. The higher creep rate and dependency on the TGO thickness led to a hypothesis that the deformation of the TGO under tensile stress at high temperature was not a result of typical creep mechanisms such as diffusion of vacancies or intra-granular motion of dislocations, but a result of inter-grain growth of TGO. Results also indicate that the amount of yttrium may influence the growth strain as well as the creep rate.

  7. Defect accumulation in welded joints of 12Kh1MF steel steam pipelines during creep

    International Nuclear Information System (INIS)

    Anokhov, A.E.; Alekhova, I.A.

    1982-01-01

    Technique for investigation of micropore accumulation in 12Kh1MF steel welded joints is proposed. The micropore density in different zones of welded joints in non-uniform. It is shown that failure localization in welded joint in the softening zone during the creep takes place due to the micropore priority accumulation in this zone. It is found out that accumulation of residual deformations in different zones of 12Kh1MF welded joints under creep runs more uniformly as the level of working stresses decreases and test duration increases

  8. Shakedown and ratchetting below the creep range

    International Nuclear Information System (INIS)

    Ponter, A.R.S.

    1983-01-01

    The report reviews current understanding of the behaviour of structure subject to variable mechanical and thermal loading below the creep range through a comparison of theoretical solutions and experimental studies. The particular characteristics of the austenitic stainless steels are emphasized in components subject to moderate primary loads and large thermal loads. The review shows that a clear classification of types of thermal loading is required in design. Two main classes, termed category A and B, exist which arise not from the magnitude of the thermal stresses but their extent through the material volume of the structure. In category A situations, the Bree plate problem being the prime example, the maximum thermal stresses occur over a volume of the structure which does not contain a mechanism of failure. As a result very large thermal stresses may be withstood without ratchetting occurring for sufficiently small mechanical loads. For category B situations, the maximum thermal stress occur within a volume of material which contains a mechanism of deformation. In such cases, the capacity of the structure to withstand thermal loading is limited by a variation of the maximum thermal stress at a material point of 2σsub(γ) where σsub(γ) is a suitably defined yield stress. This situation seems to be the most typical problem of the Liquid Metal Fast Reactor and the ''3Sm'' limit in the ASME III code restriction on secondary stress cannot be exceeded if ratchetting is to be prevented

  9. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    International Nuclear Information System (INIS)

    Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-01-01

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  10. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-04-15

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  11. The prediction of creep damage in Type 347 weld metal: part II creep fatigue tests

    International Nuclear Information System (INIS)

    Spindler, M.W.

    2005-01-01

    Calculations of creep damage under conditions of strain control are often carried out using either a time fraction approach or a ductility exhaustion approach. In part I of this paper the rupture strength and creep ductility data for a Type 347 weld metal were fitted to provide the material properties that are used to calculate creep damage. Part II of this paper examines whether the time fraction approach or the ductility exhaustion approach gives the better predictions of creep damage in creep-fatigue tests on the same Type 347 weld metal. In addition, a new creep damage model, which was developed by removing some of the simplifying assumptions that are made in the ductility exhaustion approach, was used. This new creep damage model is a function of the strain rate, stress and temperature and was derived from creep and constant strain rate test data using a reverse modelling technique (see part I of this paper). It is shown that the new creep damage model gives better predictions of creep damage in the creep-fatigue tests than the time fraction and the ductility exhaustion approaches

  12. Creep and creep recovery of concrete subjected to triaxial compressive stresses at elevated temperature

    International Nuclear Information System (INIS)

    Ohnuma, Hiroshi; Abe, Hirotoshi

    1979-01-01

    In order to design rationally the vessels made of prestressed concrete for nuclear power stations and to improve the accuracy of high temperature creep analysis, the Central Research Institute of Electric Power Industry had carried out the proving experiments with scale models. In order to improve the accuracy of analysis, it is important to grasp the creep behavior of the concrete subjected to triaxial compressive stresses at high temperature as the basic property of concrete, because actual prestressed concrete vessels are in such conditions. In this paper, the triaxial compression creep test at 60 deg. C using the concrete specimens with same mixing ratio as the scale models is reported. The compressive strength of the concrete at the age of 28 days was 406 kg/cm 2 , and the age of the concrete at the time of loading was 63 days. Creep and creep recovery were measured for 5 months and 2 months, respectively. The creep of concrete due to uniaxial compression increased with temperature rise, and the creep strain at 60 deg. C was 2.54 times as much as that at 20 deg. C. The effective Poisson's ratio in triaxial compression creep was 0.15 on the average, based on the creep strain due to uniaxial compression at 60 deg. C. The creep recovery rate in high temperature, triaxial compression creep was 33% on the average. (Kako, I.)

  13. A method of creep rupture data extrapolation based on physical processes

    International Nuclear Information System (INIS)

    Leinster, M.G.

    2008-01-01

    There is a need for a reliable method to extrapolate generic creep rupture data to failure times in excess of the currently published times. A method based on well-understood and mathematically described physical processes is likely to be stable and reliable. Creep process descriptions have been developed based on accepted theory, to the extent that good fits with published data have been obtained. Methods have been developed to apply these descriptions to extrapolate creep rupture data to stresses below the published values. The relationship creep life parameter=f(ln(sinh(stress))) has been shown to be justifiable over the stress ranges of most interest, and gives realistic results at high temperatures and long times to failure. In the interests of continuity with past and present practice, the suggested method is intended to extend existing polynomial descriptions of life parameters at low stress. Where no polynomials exist, the method can be used to describe the behaviour of life parameters throughout the full range of a particular failure mode in the published data

  14. Integrated Design Software Predicts the Creep Life of Monolithic Ceramic Components

    Science.gov (United States)

    1996-01-01

    Significant improvements in propulsion and power generation for the next century will require revolutionary advances in high-temperature materials and structural design. Advanced ceramics are candidate materials for these elevated-temperature applications. As design protocols emerge for these material systems, designers must be aware of several innate features, including the degrading ability of ceramics to carry sustained load. Usually, time-dependent failure in ceramics occurs because of two different, delayedfailure mechanisms: slow crack growth and creep rupture. Slow crack growth initiates at a preexisting flaw and continues until a critical crack length is reached, causing catastrophic failure. Creep rupture, on the other hand, occurs because of bulk damage in the material: void nucleation and coalescence that eventually leads to macrocracks which then propagate to failure. Successful application of advanced ceramics depends on proper characterization of material behavior and the use of an appropriate design methodology. The life of a ceramic component can be predicted with the NASA Lewis Research Center's Ceramics Analysis and Reliability Evaluation of Structures (CARES) integrated design programs. CARES/CREEP determines the expected life of a component under creep conditions, and CARES/LIFE predicts the component life due to fast fracture and subcritical crack growth. The previously developed CARES/LIFE program has been used in numerous industrial and Government applications.

  15. Creep-Fatigue Relationsihps in Electroactive Polymer Systems and Predicted Effects in an Actuator Design

    Science.gov (United States)

    Vinogradov, Aleksandra M.; Ihlefeld, Curtis M.; Henslee, Issac

    2009-01-01

    The paper concerns the time-dependent behavior of electroactive polymers (EAP) and their use in advanced intelligent structures for space exploration. Innovative actuator design for low weight and low power valves required in small plants planned for use on the moon for chemical analysis is discussed. It is shown that in-depth understanding of cyclic loading effects observed through accelerated creep rates due to creep-fatigue interaction in polymers is critical in terms of proper functioning of EAP based actuator devices. In the paper, an overview of experimental results concerning the creep properties and cyclic creep response of a thin film piezoelectric polymer polyvinylidene fluoride (PVDF) is presented. The development of a constitutive creep-fatigue interaction model to predict the durability and service life of electroactive polymers is discussed. A novel method is proposed to predict damage accumulation and fatigue life of polymers under oyclic loading conditions in the presence of creep. The study provides a basis for ongoing research initiatives at the NASA Kennedy Space Center in the pursuit of new technologies using EAP as active elements for lunar exploration systems.

  16. Creep of Li2O

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Arthur, B.; Lui, Y.Y.

    1985-01-01

    The objective of this effort was to obtain data on the performance of lithium ceramic materials during fast neutron irradiation in support of solid breeder blanket designs. Li 2 O has been observed to swell (greater than or equal to 4%) under fast reactor irradiation. Fortunately, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory creep experiments were conducted between 500 to 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data

  17. Creep of Li2O

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Liu, Y.Y.; Arthur, B.

    1984-11-01

    The tritium breeding material with the highest lithium atom density, Li 2 O has been observed to incur significant swelling (>4%) under fast reactor irradiation. Such swelling, if unrestrained leads to either unacceptable, induced-strains in adjacent structural material or undesirable design compromises. Fortunately, however, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory dilational creep experiments were conducted on unirradiated Li 2 O between 500 and 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data

  18. Significance of primary irradiation creep in graphite

    CSIR Research Space (South Africa)

    Erasmus, C

    2013-05-01

    Full Text Available Traditionally primary irradiation creep is introduced into graphite analysis by applying the appropriate amount of creep strain to the model at the initial time-step. This is valid for graphite components that are subjected to high fast neutron flux...

  19. Measurement of soil creep by inclinometer

    Science.gov (United States)

    Robert R. Ziemer

    1977-01-01

    Abstract - Continued inclinometer measurements at borehole sites installed in 1964 in northern California suggest that previously reported rates of soil creep are excessively high. Upon analysis of 35 access casings located in forested and grassland sites, no consistent direction of soil movement could be detected. In addition, no significant rate of soil creep could...

  20. Creep of parylene-C film

    KAUST Repository

    Lin, Jeffrey Chun-Hui; Deng, Peigang; Lam, Gilbert; Lu, Bo; Lee, Yi-Kuen; Tai, Yu-Chong

    2011-01-01

    The glass transition temperature of as-deposited parylene-C is first measured to be 50°C with a ramping-temperature-dependent modulus experiment. The creep behavior of parylene-C film in the primary and secondary creep region is then investigated

  1. A phenomenological theory of transient creep

    International Nuclear Information System (INIS)

    Ajaja, O.; Ardell, A.J.

    1979-01-01

    A new creep theory is proposed which takes into account the strain generated during the annihilation of dislocations. This contribution is found to be very significant when recovery is appreciable, and is mainly responsible for the decreasing creep rate associated with the normal primary creep of class II materials. The theory provides excellent semiquantitative rationalization for the types of creep curves presented in the preceding paper. In particular, the theory predicts a change in the shape of the primary creep curve from normal to inverted as recovery becomes less important, i.e. as the applied stress and/or temperature decrease(s). It also predicts a minimum creep rate under certain circumstances, hence pseudo-tertiary behaviour. These different types of creep curves are predicted even though the net dislocation density decreases monotonically with time in all cases. Qualitative rationalization is presented for the inverted transient which always follows a stress drop in class II materials, as well as for the inverted primary and sigmoidal creep behaviour of class I solid solutions. (author)

  2. Creep of titanium--silicon alloys

    International Nuclear Information System (INIS)

    Paton, N.E.; Mahoney, M.W.

    1976-01-01

    Operative creep mechanisms in laboratory melts of Ti-5Zr-0.5Si and Ti-5Zr-0.5Si have been investigated as a function of microstructure, creep stress, and temperature. From creep rate data and transmission electron microscopy results, it has been shown that an important creep strengthening mechanism at 811 0 K in Si-bearing Ti alloys is clustering of solute atoms on dislocations. All of the alloys investigated showed anomalously high apparent activation energies and areas for creep and a high exponent (n) in the Dorn equation. In addition, the effect of heat treatment was investigated and it is shown that the highest creep strength was obtained by using a heat treatment which retained the maximum amount of silicon in solution. This is consistent with the proposed creep strengthening mechanism. An investigation of the creep behavior of several other Si containing alloys including two commercial alloys, Ti-11 and IMI-685 indicated similar results. 12 fig., 6 tables

  3. Towards self-healing creep resistant steels

    NARCIS (Netherlands)

    Van der Zwaag, S.; Zhang, S.; Fang, H.; Bruck, E.; Van Dijk, N.H.

    2016-01-01

    We report the main findings of our work on the behaviour of binary Fe-Cu and Fe-Au model alloys designed to explore routes to create new creep resistant steels having an in-built ability to autonomously fill creep induced porosity at grain boundaries. The alloying elements were selected on the basis

  4. Analysis of irradiation creep and the structural integrity of fusion in-vessel components

    International Nuclear Information System (INIS)

    Karditsas, Panayiotis J.

    2000-01-01

    This paper presents a brief review of the irradiation creep mechanism, analyses of the effect on the performance and behaviour of fusion in-vessel components, and discusses procedures for the estimation of in-service time (or lifetime) of components under combined creep-fatigue. The irradiation creep models and proposed theories are examined and analysed to produce a creep law relevant to fusion conditions. The necessary material data, constitutive equations and other parameters needed for estimation of in-service time from the combination of creep and fatigue damage are identified. Wherever possible, design curves are proposed for stress and strain. Time dependent non-linear elastoplastic example calculations are performed, for a typical first wall structure under power plant loading conditions, assuming austenitic and martensitic steel as structural materials, including material irradiation creep. The results of calculations for the stress and strain history of the first wall are used together with the proposed cumulative damage expressions derived in this study to estimate the in-service time, including the effects of stress relaxation due to creep, reduction of ductility (or fracture strain) and helium-to-displacement-damage ratio. The calculations give a displacement damage of ∼70 dpa for the 316 austenitic steel and ∼110-130 dpa for the martensitic steel. Provided there are no power transients, for a design strain of 0.5%, the in-service time is estimated to be ∼3 years for the 316 steel case (at 2.2 MW/m 2 wall load) and the high wall loading martensitic steel (5.0 MW/m 2 case), and ∼5.3 years for the martensitic steel at lower wall load (2.2 MW/m 2 case). The difficulty in defending these results lies in the uncertainty arising from the limited database and experience of the material properties, especially the creep constitutive law, when exposed to fusion environments

  5. Creep equations for gas turbine materials

    International Nuclear Information System (INIS)

    Kloos, K.H.; Granacher, J.; Preussler, T.

    1988-01-01

    The long-term high-temperature deformation behaviour of typical gas turbine materials can be described on the basis of a differentiated evaluation which takes the results from thermal tension tests, short-term creep tests with continuous extension measurement, long-term creep tests with discontinuous extension measurement as well as annealing tests with contraction measurement into account. By this, especially the 'negative creeping' can be controlled. Equations were developed for individual materials of the type IN-738 LC, IN-939, IN-100 and FSX-414, which describe the high-temperature deformation behaviour with consideration to the primary and secondary creeping and partly the tertiary creeping. The equations are valid in the entire application-relevant range, i.e. up to 100 000 h in the case of industrial turbine materials. (orig.) [de

  6. Creep resistant high temperature martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Hawk, Jeffrey A.; Jablonski, Paul D.; Cowen, Christopher J.

    2017-01-31

    The disclosure provides a creep resistant alloy having an overall composition comprised of iron, chromium, molybdenum, carbon, manganese, silicon, nickel, vanadium, niobium, nitrogen, tungsten, cobalt, tantalum, boron, copper, and potentially additional elements. In an embodiment, the creep resistant alloy has a molybdenum equivalent Mo(eq) from 1.475 to 1.700 wt. % and a quantity (C+N) from 0.145 to 0.205. The overall composition ameliorates sources of microstructural instability such as coarsening of M.sub.23C.sub.6carbides and MX precipitates, and mitigates or eliminates Laves and Z-phase formation. A creep resistant martensitic steel may be fabricated by preparing a melt comprised of the overall composition followed by at least austenizing and tempering. The creep resistant alloy exhibits improved high-temperature creep strength in the temperature environment of around 650.degree. C.

  7. Deformation mechanisms in cyclic creep and fatigue

    International Nuclear Information System (INIS)

    Laird, C.

    1979-01-01

    Service conditions in which static and cyclic loading occur in conjunction are numerous. It is argued that an understanding of cyclic creep and cyclic deformation are necessary both for design and for understanding creep-fatigue fracture. Accordingly a brief, and selective, review of cyclic creep and cyclic deformation at both low and high strain amplitudes is provided. Cyclic loading in conjunction with static loading can lead to creep retardation if cyclic hardening occurs, or creep acceleration if softening occurs. Low strain amplitude cyclic deformation is understood in terms of dislocation loop patch and persistent slip band behavior, high strain deformation in terms of dislocation cell-shuttling models. While interesting advances in these fields have been made in the last few years, the deformation mechanisms are generally poorly understood

  8. Assessment of concrete creep and shrinkage

    International Nuclear Information System (INIS)

    Trivedi, Neha; Singh, R.K.

    2012-01-01

    B-3 model prediction of concrete creep and shrinkage strains on cylindrical specimen and BARC Containment test model (BARCOM) are presented. Experimental shrinkage strain is shown to be in agreement with B-3 model predictions for cylindrical specimen and BARCOM. Creep strain in cylindrical specimen is found to be in agreement with B-3 model. In BARCOM for wall cast in different pores, creep strain is in agreement with B-3 model in hoop direction however in longitudinal direction, observed creep strain in higher than B-3 model. For dome structure cast in a single pour, experimental creep strain shows confirmity with B-3 model both in hoop and longitudinal directions. The study on concrete aging and average longitudinal shrinkage strain is carried out. (author)

  9. Creep lifetime assessements of ferritic pipeline welds

    International Nuclear Information System (INIS)

    Ainsworth, R.A.; Goodall, I.W.; Miller, D.A.

    1995-01-01

    The low alloy ferritic steam pipework in Advanced Gas Cooled reactor (AGR) power stations operates at temperatures in the creep range. An inspection strategy for continued operation of the pipework has been developed based on estimation of the creep rupture life of pipework weldments and fracture mechanics for demonstrating acceptance of defects. This strategy is described in outline. The estimation of creep rupture life is described in more detail. Validation for the approach is illustrated by comparison with pressure vessel tests and with metallographic examination of components removed from service. The fracture mechanics methods are also described. It is shown that the amount of creep crack growth is dependent on the life fraction at which the assessment is made; crack growth being rapid as the creep rupture life is approached. (author). 3 refs., 5 figs., 1 tab

  10. Creep and relaxation behavior of Inconel-617

    International Nuclear Information System (INIS)

    Osthoff, W.; Ennis, P.J.; Nickel, H.; Schuster, H.

    1984-01-01

    The static and dynamic creep behavior of Inconel alloy 617 has been determined in constant load creep tests, relaxation tests, and stress reduction tests in the temperature range 1023 to 1273 K. The results have been interpreted using the internal stress concept: The dependence of the internal stress on the applied stress and test temperature was determined. In a few experiments, the influence of cold deformation prior to the creep test on the magnitude of the internal stress was also investigated. It was found that the experimentally observed relaxation behavior could be more satisfactorily described using the Norton creep equation modified by incorporation of the internal stress than by the conventional Norton creep equation

  11. Documentation for the viscoplastic and creep program

    DEFF Research Database (Denmark)

    Bellini, Anna

    2004-01-01

    of this workpackage is to simulate creep behavior of aluminum cast samples subjected to high temperature. In this document a two-state variables unified model is applied in order to simulate creep behavior and time-dependent metallurgical changes. The fundamental assumption of the unified theory is that creep...... is run using the material data obtained through the mentioned experimental study. The results obtained for the simulation of tensile tests and of creep tests are compared with experimental curves, showing a good agreement. Moreover, the document describes the results obtained during the first...... is quite stable and convergence can be reached also with big time steps. Keywords: Viscoplasticity, creep, unified constitutive model, aluminum, high temperature....

  12. Creep resistant high temperature martensitic steel

    Science.gov (United States)

    Hawk, Jeffrey A.; Jablonski, Paul D.; Cowen, Christopher J.

    2015-11-13

    The disclosure provides a creep resistant alloy having an overall composition comprised of iron, chromium, molybdenum, carbon, manganese, silicon, nickel, vanadium, niobium, nitrogen, tungsten, cobalt, tantalum, boron, and potentially additional elements. In an embodiment, the creep resistant alloy has a molybdenum equivalent Mo(eq) from 1.475 to 1.700 wt. % and a quantity (C+N) from 0.145 to 0.205. The overall composition ameliorates sources of microstructural instability such as coarsening of M.sub.23C.sub.6 carbides and MX precipitates, and mitigates or eliminates Laves and Z-phase formation. A creep resistant martensitic steel may be fabricated by preparing a melt comprised of the overall composition followed by at least austenizing and tempering. The creep resistant alloy exhibits improved high-temperature creep strength in the temperature environment of around 650.degree. C.

  13. Creep collapse of TAPS fuel cladding

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Anand, A.K.

    1975-01-01

    Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported (internally) at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : (1) inelastic collapse of perfectly circular tubes (2) plastic instability in oval tubes (3) effect of creep on ovality. Creep is considered to be a non-linear combination of the following : (a) thermal creep (b) intresenic creep (c) stress aided radiation enhanced (d) stress free growth (4) Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. (author)

  14. Analysis of Current HT9 Creep Correlations and Modification

    International Nuclear Information System (INIS)

    Lee, Cheol Min; Sohn, Dongseong; Cheon, Jin Sik

    2014-01-01

    It has high thermal conductivity, high mechanical strength and low irradiation induced swelling. However high temperature creep of HT9 has always been a life limiting factor. Above 600 .deg. C, the dislocation density in HT9 is decreased and the M 23 C 6 precipitates coarsen, these processes are accelerated if there is irradiation. Finally microstructural changes at high temperature lead to lower creep strength and large creep strain. For HT9 to be used as a future cladding, creep behavior of the HT9 should be predicted accurately based on the physical understanding of the creep phenomenon. Most of the creep correlations are composed of irradiation creep and thermal creep terms. However, it is certain that in-pile thermal creep and out-of-pile thermal creep are different because of the microstructure changes induced from neutron irradiation. To explain creep behavior more accurately, thermal creep contributions other than neutron irradiation should be discriminated in a creep correlation. To perform this work, existing HT9 creep correlations are analyzed, and the results are used to develop more accurate thermal creep correlation. Then, the differences between in-pile thermal creep and out-of-pile thermal creep are examined

  15. High temperature creep-fatigue design

    International Nuclear Information System (INIS)

    Tavassoli, A. A. F.; Fournier, B.; Sauzay, M.

    2010-01-01

    Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

  16. High temperature creep-fatigue design

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A. A. F.; Fournier, B.; Sauzay, M. [CEA Saclay, DEN DMN, F-91191 Gif Sur Yvette (France)

    2010-07-01

    Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowable and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed. (authors)

  17. Illegal gene flow from transgenic creeping bentgrass: the saga continues.

    Science.gov (United States)

    Snow, Allison A

    2012-10-01

    Ecologists have paid close attention to environmental effects that fitness-enhancing transgenes might have following crop-to-wild gene flow (e.g. Snow et al. 2003). For some crops, gene flow also can lead to legal problems,especially when government agencies have not approved transgenic events for unrestricted environmental release.Creeping bentgrass (Agrostis stolonifera), a common turf grass used in golf courses, is the focus of both areas of concern. In 2002, prior to expected deregulation (still pending), The Scotts Company planted creeping bentgrass with transgenic resistance to the herbicide glyphosate,also known as RoundUp, on 162 ha in a designated control area in central Oregon (Fig. 1).Despite efforts to restrict gene flow, wind-dispersed pollen carried transgenes to florets of local A. stolonifera and A. gigantea as far as 14 km away, and to sentinel plants placed as far as 21 km away (Watrud et al. 2004).Then, in August 2003, a strong wind event moved transgenic seeds from wind rows of cut bentgrass into nearby areas. The company’s efforts to kill all transgenic survivors in the area failed: feral glyphosate-resistant populations of A. stolonifera were found by Reichman et al.(2006), and 62% of 585 bentgrass plants had the telltale CP4 EPSPS transgene in 2006 (Zapiola et al. 2008; Fig. 2).Now, in this issue, the story gets even more interesting as Zapiola & Mallory-Smith (2012) describe a transgenic,intergeneric hybrid produced on a feral, transgenic creeping bentgrass plant that received pollen from Polypogon monspeliensis (rabbitfoot grass). Their finding raises a host of new questions about the prevalence and fitness of intergeneric hybrids, as well as how to evaluate the full extent of gene flow from transgenic crops.

  18. The effect of hydrostatic pressure on the creep life of a 2.25% Cr1% Mo Steel

    International Nuclear Information System (INIS)

    Lonsdale, D.; Flewitt, P.E.J.

    1979-01-01

    The effect of superimposed hydrostatic pressures on the creep life of a 2.25% Cr 1% Mo steel, with a bainitic microstructure, is examined. Creep tests have been carried out at 923K with uniaxial stresses in the range 55 to 80 MPa and superimposed hydrostatic pressures up to 35 MPa. Optical and scanning electron microscopy have been used to evaluate the contribution of the hydrostatic stress to grain boundary cavitation. Increasing the hydrostatic pressure for a given uniaxial stress suppresses cavitation, but little changes the secondary creep rate, thereby excending the creep life. Furthermore, the time to failure depends on both the applied uniaxial stress and the hydrostatic pressure and not simply the principal stress. (orig.) [de

  19. Revision of Drucker-Prager cap creep modelling of pebble beds in fusion blankets

    International Nuclear Information System (INIS)

    Hofer, D.; Kamlah, M.; Hermsmeyer, S.

    2004-01-01

    A continuum model commonly used in soil mechanics analysis is compiled by use of a finite element software and has been used to simulate the thermomechanical behaviour of pebble beds. The Drucker-Prager Cap theory accounts for inelastic volume change, cap hardening, nonlinear elasticity and pressure dependent shear failure. The hardening mechanism allows for defining the hydrostatic pressure yield stress as a function of the volumetric inelastic strain. Volumetric creep is considered in order to simulate the pebble bed behaviour at high temperatures. Here, the strain hardening option has been used for the consolidation creep mechanism. The model has been calibrated using the fitting curves of the oedometric test given by Reimann et al. The fitted data has been used to calculate a pebble bed with simplified boundary conditions loaded by non-uniform volumetric heating. This calculation demonstrated that the model is capable of representing creep behaviour under volumetric heating conditions. (author)

  20. The creep life of superheater and reheater tubes under varying pressure conditions in operational boilers

    International Nuclear Information System (INIS)

    Mizen, D.C.; Plastow, B.

    1975-01-01

    The first of each manufacturer's 500 MW boilers supplied to the CEGB (Central Electricity Generating Board) have been subjected to an extensive programme of tests for performance optimization and safe operation. Around 250 thermocouples on superheater and reheater tubes have in each case been monitored as part of the exercise. The readings are corrected and used to compute creep rupture damage based on internationally agreed stress rupture data and a simple cumulative damage concept. Comparison of the design creep rupture life and the cumulative life consumed has in several applications been invaluable in influencing operating procedures and arranging tube modifications or replacements, so that loss of generation by creep rupture failure is minimized. (author)

  1. More plant availability by local and integral strain measurement

    Energy Technology Data Exchange (ETDEWEB)

    Hulshof, H J.M.; Welberg, P G.M. [KEMA Nederland B.V., Arnhem (Netherlands); Bruijn, L.E. de [E.ON Benelux Generation N.V., Rotterdam (Netherlands). Power Plant Maasvlakte

    2002-07-01

    Industrial installations that are under pressure and are operating at high temperatures have a limited life due to creep and fatigue. It is, therefore, of critical importance to know the location of any possible weak spots in the installation. Welds in steam pipes, especially the heat-affected zones in these welds, are such weak spots in the long term. The material deforms and cracks may develop, with significant failure in the worst case. To avoid safety risks, unforeseen plant shutdown and, as a consequence, high costs for unavailability and repair, periodic inspections and strain measurements are recommended. KEMA's SPICA (Speckle Image Correlation Analysis) system is able to measure on-stream (during operation) deformation due to creep in critical areas, like the heat-affected zone in welds. (orig.)

  2. ESTIMATION MODEL OF RESIDUAL LIFE-TIME OF LOCOMOTIVE FRAME BOGIE WITH ALLOWANCE FOR CREEP

    Directory of Open Access Journals (Sweden)

    V. R. Skalskyi

    2014-12-01

    Full Text Available Purpose. The problem of determining the residual life of frame bogie elements of locomotives is a great importance for predicting their work safely and avoidance potential failures on the track. This especially concern cases when such elements have creep-fatigue cracks which grow under action of cyclic loading with excerpts T1 in the cycle and reach their critical size. Here the question of the propagation of such defects (cracks arises, their kinetics and about the period of subcritical cracks growth. The aim is to develop a calculation model for determination the period of subcritical creep-fatigue cracks growth in the bogies frames of electric locomotive. The model takes into account the basic parameters of load, geometry of the construction element and cracks. Methodology. The calculation model for determination the period of subcritical creep-fatigue cracks growth in structural elements of frame under conditions of variable load time has been formulated. It is based on the first law of thermodynamics concerning to mechanics of solids slow fracture at low temperature creep and variable loadings. It is assumed that the period of unsteady creep dominates here (the first section of the creep curve. Low-temperature creep is creep of materials at temperatures T0 < 0,5Tmp, where Tmp − the melting point of the material. Findings. The analytical formula for the determination of the stress intensity factor of truck bolster with technological hole has been obtained. It is shown that by experimentally established constants of the material using the proposed analytical relations can easily determine residual resource of the bogie frame elements. Originality. The new mathematical model for describing the kinetics of creep-fatigue cracks growth in the frames bogies of electric locomotive under variable in time loadings with various time excerpts and on this base the period determination of subcritical crack growth has been proposed. Practical value

  3. Critical review of creep FRAPCON-3 model under dry storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L.E. [Unit of Nuclear Safety Research, CIEMAT, Avda. Complutense 22, Madrid, Madrid 28040 (Spain)

    2009-06-15

    There is a general agreement that cladding creep rupture is the most likely and limiting failure mechanism of spent fuel in dry storage compared to other potential mechanisms, like stress corrosion cracking and/or delayed hydride cracking. Nevertheless, occurrence of creep rupture is very improbable since both decay heat and hoop stress tend to decrease throughout dry storage. In spite of this, the current trend to higher burn up levels needs further attention that ensures safe storage of spent fuel irradiated over 45 GWd/MTU. An extensive work has been carried out during the last four decades in the area of in-reactor creep modelling. Unfortunately, the in-reactor conditions are so different from those prevailing under dry storage, that all the experience gained cannot be extrapolated in a straightforward manner. On the other side, as creep tests simulating conditions throughout a 20-40 year dry storage are impractical, post-irradiation cladding creep behaviour has been modelled by means of time-temperature dependent laws developed on the basis of currently available zirconium alloys data. Additionally, some tests have been exploring the effect of irradiation, hydrogen distribution and material composition on the materials creep behaviour. Adaptation of fuel performance codes initially developed for normal and off-normal reactor operation is not an easy task either. Creep modelling is usually dependent of host codes because a good part of its validation and update has been carried out in an integral way, and as a consequence its independent performance assessment is not an easy task. This work examines the current capability of FRAPCON-3 to model creep behaviour under dry storage conditions. To do so, a review of its major fundamentals has been done and its range of applicability discussed. Once its main approximations and drawbacks have been identified, an attempt to overcome some of them has been intended by implementing an alternative expression for creep under

  4. Creep avalanches on San Andreas Fault and their underlying mechanism from 19 years of InSAR and seismicity

    Science.gov (United States)

    Khoshmanesh, M.; Shirzaei, M.

    2017-12-01

    failure stress up to 0.45 bar/yr. While creeping segments are suggested to act as barriers and arrest rupture, our study implies that SSEs on these zones may trigger seismic events on adjacent locked parts.

  5. A methodology to analyze the creep behaviour of nuclear fuel waste containers

    International Nuclear Information System (INIS)

    Dutton, R.

    1995-12-01

    The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs

  6. A methodology to analyze the creep behaviour of nuclear fuel waste containers

    Energy Technology Data Exchange (ETDEWEB)

    Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1995-12-01

    The concept for the disposal of used-fuel waste from CANDU reeactors operating in Canada comprises a system of natural and engineered barriers surrounding the waste in a mined vault situated at a depth of 500 - 1000 m in plutonic rock of the Canadian Shield. The fuel would be packaged in a highly durable metal container, within a matrix of compacted particulate. The design of the container takes into account that it would be subjected to an external hydrostatic pressure. Consideration of the rate of radioactive decay of the radionuclides contained in the fuel, suggests that the lifetime of the container should be at least 500 years. Consequently, the role of creep deformation, and the possibility of creep rupture of the container shell, must be included in the assessment of time-dependent mechanical integrity. This report describes an analytical approach that can be used to quantify the long-term creep properties of the container material and facilitate the engineering design. The overall objective is to formulate a constitutive creep equation that provides the required input for a finite element computer model being developed to analyze the elastic-plastic behaviour of the container. Alternative forms of such equations are reviewed. It is shown that the capability of many of these equations to extrapolate over long time scales is limited by their empirical nature. Thus, the recommended equation is based on current mechanistic understanding of creep deformation and creep rupture. A criterion for determining the onset of material failure by creep rupture, that could be used in the design of containers with extended structural integrity, is proposed. Interpretation and extrapolation will be supported by the complementary Deformation and Fracture Mechanism Maps. (author) 103 refs., 2 tabs., 54 figs.

  7. Computational Design of Creep-Resistant Alloys and Experimental Validation in Ferritic Superalloys

    Energy Technology Data Exchange (ETDEWEB)

    Liaw, Peter

    2014-12-31

    A new class of ferritic superalloys containing B2-type zones inside parent L21-type precipitates in a disordered solid-solution matrix, also known as a hierarchical-precipitate strengthened ferritic alloy (HPSFA), has been developed for high-temperature structural applications in fossil-energy power plants. These alloys were designed by the addition of the Ti element into a previously-studied NiAl-strengthened ferritic alloy (denoted as FBB8 in this study). In the present research, systematic investigations, including advanced experimental techniques, first-principles calculations, and numerical simulations, have been integrated and conducted to characterize the complex microstructures and excellent creep resistance of HPSFAs. The experimental techniques include transmission-electron microscopy, scanningtransmission- electron microscopy, neutron diffraction, and atom-probe tomography, which provide detailed microstructural information of HPSFAs. Systematic tension/compression creep tests revealed that HPSFAs exhibit the superior creep resistance, compared with the FBB8 and conventional ferritic steels (i.e., the creep rates of HPSFAs are about 4 orders of magnitude slower than the FBB8 and conventional ferritic steels.) First-principles calculations include interfacial free energies, anti-phase boundary (APB) free energies, elastic constants, and impurity diffusivities in Fe. Combined with kinetic Monte- Carlo simulations of interdiffusion coefficients, and the integration of computational thermodynamics and kinetics, these calculations provide great understanding of thermodynamic and mechanical properties of HPSFAs. In addition to the systematic experimental approach and first-principles calculations, a series of numerical tools and algorithms, which assist in the optimization of creep properties of ferritic superalloys, are utilized and developed. These numerical simulation results are compared with the available experimental data and previous first

  8. RNA-Seq analysis of the Sclerotinia homoeocarpa--creeping bentgrass pathosystem.

    Directory of Open Access Journals (Sweden)

    Angela M Orshinsky

    Full Text Available Sclerotinia homoeocarpa causes dollar spot disease, the predominate disease on highly-maintained turfgrass. Currently, there are major gaps in our understanding of the molecular interactions between S. homoeocarpa and creeping bentgrass. In this study, 454 sequencing technology was used in the de novo assembly of S. homoeocarpa and creeping bentgrass transcriptomes. Transcript sequence data obtained using Illumina's first generation sequencing-by-synthesis (SBS were mapped to the transcriptome assemblies to estimate transcript representation in different SBS libraries. SBS libraries included a S. homoeocarpa culture control, a creeping bentgrass uninoculated control, and a library for creeping bentgrass inoculated with S. homoeocarpa and incubated for 96 h. A Fisher's exact test was performed to determine transcripts that were significantly different during creeping bentgrass infection with S. homoeocarpa. Fungal transcripts of interest included glycosyl hydrolases, proteases, and ABC transporters. Of particular interest were the large number of glycosyl hydrolase transcripts that target a wide range of plant cell wall compounds, corroborating the suggested wide host range and saprophytic abilities of S. homoeocarpa. Several of the multidrug resistance ABC transporters may be important for resistance to both fungicides and plant defense compounds. Creeping bentgrass transcripts of interest included germins, ubiquitin transcripts involved in proteasome degradation, and cinnamoyl reductase, which is involved in lignin production. This analysis provides an extensive overview of the S. homoeocarpa-turfgrass pathosystem and provides a starting point for the characterization of potential virulence factors and host defense responses. In particular, determination of important host defense responses may assist in the development of highly resistant creeping bentgrass varieties.

  9. Statistical investigations of the failure behaviour of components in the AVR experimental nuclear power plant. Vol. 2

    International Nuclear Information System (INIS)

    Meyna, A.; Mock, R.; Tietze, A.; Hennings, W.

    1989-08-01

    From operational records of the years 1977 to 1986, service life distributions of helium valves in gas circuits of the AVR were determined. Results are constant failure rates in the range from 3 to 6x10 -6 /h and, for some populations, indications of time dependent failure rates. Nonparametric methods showed only limited efficiency. For a Bayesian approach the necessary prior information was missing. Furthermore, the main failure causes could be determined. (orig./HP) [de

  10. Prediction of the creep properties of discontinuous fibre composites from the matrix creep law

    International Nuclear Information System (INIS)

    Bilde-Soerensen, J.B.; Boecker Pedersen, O.; Lilholt, H.

    1975-02-01

    Existing theories for predicting the creep properties of discontinuous fibre composites with non-creeping fibres from matrix creep properties, originally based on a power law, are extended to include an exponential law, and in principle a general matrixlaw. An analysis shows that the composite creep curve can be obtained by a simple displacement of the matrix creep curve in a log sigma vs. log epsilon diagram. This principle, that each point on the matrix curve has a corresponding point on the composite curve,is given a physical interpretation. The direction of displacement is such that the transition from a power law toan exponential law occurs at a lower strain rate for the composite than for the unreinforced matrix. This emphasizes the importance of the exponential creep range in the creep of fibre composites. The combined use of matrix and composite data may allow the creep phenomenon to be studied over a larger range of strain rates than otherwise possible. A method for constructing generalized composite creep diagrams is suggested. Creep properties predicted from matrix data by the present analysis are compared with experimental data from the literature. (author)

  11. Deformation by grain boundary sliding and slip creep versus diffusional creep

    International Nuclear Information System (INIS)

    Ruano, O A; Sherby, O D; Wadsworth, J.

    1998-01-01

    A review is presented of the debates between the present authors and other investigators regarding the possible role of diffusional creep in the plastic flow of polycrystalline metals at low stresses. These debates are recorded in eleven papers over the past seventeen years. ln these papers it has been shown that the creep rates of materials in the so-called diffusional creep region are almost always higher than those predicted by the diffusional creep theory. Additionally, the predictions of grain size effects and stress exponents from diffusional creep theory are often not found in the experimental data. Finally, denuded zones have been universally considered to be direct evidence for diffusional creep; but, those reported in the literature are shown to be found only under conditions where a high stress exponent is observed. Also, the locations of the denuded zones do not match those predicted. Alternative mechanisms are described in which diffusion-controlled dislocation creep and/or grain boundary sliding are the dominant deformation processes in low-stress creep. It is proposed that denuded zones are formed by stress-directed grain boundary migration with the precipitates dissolving in the moving grain boundaries. The above observations have led us to the conclusion that grain boundary sliding and slip creep are in fact the principal mechanisms for observations of plastic flow in the so-called diffusional creep regions

  12. A method to assess collision hazard of falling rock due to slope collapse application of DEM on modeling of earthquake triggered slope failure for nuclear power plants

    International Nuclear Information System (INIS)

    Nakase, Hitoshi; Cao, Guoqiang; Tabei, Kazuto; Tochigi, Hitoshi; Matsushima, Takashi

    2015-01-01

    Risk evaluation of slope failure against nuclear power plants, which is induced by unexpectedly large earthquakes, has been urgent need for disaster prevention measures. Specially, for risk evaluation of slope failure, understanding of information such as traveling distances, collision velocities, and collision energies is very important. Discrete Element Method (DEM) such as particle simulation method contributes important role on predicting the detailed behavior of slope failure physics. In this study, instead of accurately predicting the complicated behavior of sliding and falling for each rock, we introduce the DEM modeling to evaluate the average traveling distance of collapsed rocks and its statistical variability. First, we conduct the validation test of the proposed DEM model on the basis of reconstruction of experiment results. Next, we conducted the parametric studies to examine sensitivities of important parameters. Finally, validity of the proposed method is evaluated and its applicability and technical assignments are also discussed. (author)

  13. Contribution of dislocation creep to the radiational creep of materials

    International Nuclear Information System (INIS)

    Borodin, V.A.; Ryazanov, A.I.

    1986-01-01

    The authors propose a model of the orientational dependences of the preferences of discrete linear dislocations in which the influence of the external load on the step concentration at the dislocations is taken into account. The use of this model, taking into account the mechanism of stress-induced anisotropy of the elastic interaction between point defects and dislocations, not only permits a correct qualitative explanation of the dependence of the rate of radiational creep on the basic irradiation parameters (dose, stress, temperature) but also allows approximate quantitative agreement with experimental results to be obtained. At sufficiently high stress, the theory predicts conditions of the formation of an ensemble of dislocational loops with a specific direction of the Burgers vector

  14. Some aspects of thermomechanical fatigue of AISI 304L stainless steel: Part I. creep- fatigue damage

    Science.gov (United States)

    Zauter, R.; Christ, H. J.; Mughrabi, H.

    1994-02-01

    Thermomechanical fatigue (TMF) tests on the austenitic stainless steel AISI 304L have been conducted under “true≓ plastic-strain control in vacuum. This report considers the damage oc-curring during TMF loading. It is shown how the temperature interval and the phasing (in-phase, out-of-phase) determine the mechanical response and the lifetime of the specimens. If creep-fatigue interaction takes place during in-phase cycling, the damage occurs inside the ma-terial, leading to intergranular cracks which reduce the lifetime considerably. Out-of-phase cy-cling inhibits creep-induced damage, and no lifetime reduction occurs, even if the material is exposed periodically to temperatures in the creep regime. A formula is proposed which allows prediction of the failure mode, depending on whether creep-fatigue damage occurs or not. At a given strain rate, the formula is able to estimate the temperature of transition between pure fatigue and creep-fatigue damage.

  15. Creep Behavior of High-Strength Concrete Subjected to Elevated Temperatures.

    Science.gov (United States)

    Yoon, Minho; Kim, Gyuyong; Kim, Youngsun; Lee, Taegyu; Choe, Gyeongcheol; Hwang, Euichul; Nam, Jeongsoo

    2017-07-11

    Strain is generated in concrete subjected to elevated temperatures owing to the influence of factors such as thermal expansion and design load. Such strains resulting from elevated temperatures and load can significantly influence the stability of a structure during and after a fire. In addition, the lower the water-to-binder (W-B) ratio and the smaller the quantity of aggregates in high-strength concrete, the more likely it is for unstable strain to occur. Hence, in this study, the compressive strength, elastic modulus, and creep behavior were evaluated at target temperatures of 100, 200, 300, 500, and 800 °C for high-strength concretes with W-B ratios of 30%, 26%, and 23%. The loading conditions were set as non-loading and 0.33f cu . It was found that as the compressive strength of the concrete increased, the mechanical characteristics deteriorated and transient creep increased. Furthermore, when the point at which creep strain occurred at elevated temperatures after the occurrence of transient creep was considered, greater shrinkage strain occurred as the compressive strength of the concrete increased. At a heating temperature of 800 °C, the 80 and 100 MPa test specimens showed creep failure within a shrinkage strain range similar to the strain at the maximum load.

  16. Creep crack growth by grain boundary cavitation under monotonic and cyclic loading

    Science.gov (United States)

    Wen, Jian-Feng; Srivastava, Ankit; Benzerga, Amine; Tu, Shan-Tung; Needleman, Alan

    2017-11-01

    Plane strain finite deformation finite element calculations of mode I crack growth under small scale creep conditions are carried out. Attention is confined to isothermal conditions and two time histories of the applied stress intensity factor: (i) a monononic increase to a plateau value subsequently held fixed; and (ii) a cyclic time variation. The crack growth calculations are based on a micromechanics constitutive relation that couples creep deformation and damage due to grain boundary cavitation. Grain boundary cavitation, with cavity growth due to both creep and diffusion, is taken as the sole failure mechanism contributing to crack growth. The influence on the crack growth rate of loading history parameters, such as the magnitude of the applied stress intensity factor, the ratio of the applied minimum to maximum stress intensity factors, the loading rate, the hold time and the cyclic loading frequency, are explored. The crack growth rate under cyclic loading conditions is found to be greater than under monotonic creep loading with the plateau applied stress intensity factor equal to its maximum value under cyclic loading conditions. Several features of the crack growth behavior observed in creep-fatigue tests naturally emerge, for example, a Paris law type relation is obtained for cyclic loading.

  17. Modelling of microstructural creep damage in welded joints of 316L stainless steel

    International Nuclear Information System (INIS)

    Bouche, G.

    2000-01-01

    Welded joints of 316L stainless steel under service conditions at elevated temperature are known to be preferential sites of creep damage, as compared to the base material. This damage results in the formation of cavities and the development of creep cracks which can lead to a premature failure of welded components. The complex two-phase microstructure of 316L welds was simulated by manually filling a mould with longitudinal deposited weld beads. The moulded material was then aged during 2000 hours at 600 deg. C. High resolution Scanning Electron Microscopy was largely used to examine the microstructure of the simulated material before and after ageing. Smooth and notched creep specimens were cut from the mould and tested at 600 deg. C under various stress levels. A comparison of the lifetime versus nominal stress curves for the base and welded materials shows a greater dependence of the welded material to creep phenomena. Observation and EBSD analysis show that damage is preferentially located along the austenite grain boundaries. The stress and strain fields in the notched specimens were calculated by finite element method. A correlation of this field to the observed damage was made in order to propose a predictive law relating the creep damage to the mechanical conditions applied locally. Further mechanical tests and simulation on CT specimens and mode II tubular specimens allowed validating the model under various multiaxial loading conditions. (author)

  18. Creep Behavior of High-Strength Concrete Subjected to Elevated Temperatures

    Directory of Open Access Journals (Sweden)

    Minho Yoon

    2017-07-01

    Full Text Available Strain is generated in concrete subjected to elevated temperatures owing to the influence of factors such as thermal expansion and design load. Such strains resulting from elevated temperatures and load can significantly influence the stability of a structure during and after a fire. In addition, the lower the water-to-binder (W–B ratio and the smaller the quantity of aggregates in high-strength concrete, the more likely it is for unstable strain to occur. Hence, in this study, the compressive strength, elastic modulus, and creep behavior were evaluated at target temperatures of 100, 200, 300, 500, and 800 °C for high-strength concretes with W–B ratios of 30%, 26%, and 23%. The loading conditions were set as non-loading and 0.33fcu. It was found that as the compressive strength of the concrete increased, the mechanical characteristics deteriorated and transient creep increased. Furthermore, when the point at which creep strain occurred at elevated temperatures after the occurrence of transient creep was considered, greater shrinkage strain occurred as the compressive strength of the concrete increased. At a heating temperature of 800 °C, the 80 and 100 MPa test specimens showed creep failure within a shrinkage strain range similar to the strain at the maximum load.

  19. Failure data specialization in quantitative risk assessments of process plants; Especializacao de dados de falha em analise quantitativa de riscos de plantas de processo

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Antonio C.O. [Bayer S.A., Sao Paulo, SP (Brazil); Melo, P.F. Frutuoso e [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia

    2005-07-01

    The aim of this paper is to show the Bayesian inference in reliability studies, which are used to failures rates updating in safety analyses. It is developed the impact of its using in quantitative risks assessments for industrial process plants. With this approach we find a structured and auditable way of showing the difference between an industrial installation with a good project and maintenance structure from another one that shows a low level of quality in these areas. In general the evidence from failures rates and as follow the frequency of occurrence from scenarios, which the risks taken in account in ERA, are taken from generics data banks, instead of, the installation in analysis. When using the plant data we need special effort to develop a data bank, that is, a maintenance managing system, which allows the data insertion as for example the SAP{sup R} and its PM module. (author)

  20. Case study on unusual occurrence of wire rope failure of 15/5t EOT crane at one of the plant, Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Siddu, P.V.; Subhakar, A.; Devender Rao, D.; Srichandan, T.T.; Harilingam, S.; Mahesh Kumar, V.V.; Kiran Kumar, I.; Thakur, Sudhir

    2016-01-01

    At NFC, Production of fuel bundles and other structural involves a large number of material handling activities for transfer of materials from one plant to another. One of the widely used materials handling equipment at NFC is the Electric Overhead Travelling (EOT) Crane. These EOT cranes will be tested yearly as per the statutory requirements in the presence of the competent person. Also, regular inspection and maintenance activities will be carried by the plant maintenance crew. Even with preventive maintenance and monitoring, there could be accidents due to unsafe conditions, unsafe acts or both. However, all the unusual occurrence, incidents and accidents can be prevented. This paper deals about the unusual occurrence of wire rope failure of 15/5T EOT crane at one of the plant in NFC; recommendations given by the Safety Committee, AERB and action taken by the plant to prevent reoccurrence of unusual occurrence. (author)