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Sample records for tube rupture analysis

  1. Analysis of Ruptured Heater Tube of Degasser Condenser in Wolsong Unit 4

    International Nuclear Information System (INIS)

    Kim, Hong Pyo; Kim, J. S.; Lim, Y. S.; Kim, S. S.; Hwang, S. S.; Kim, D. J.; Kim, S. W.; Jeong, M. K.; Hong, J. H.

    2007-08-01

    In a degasser condenser in Wolsong unit 4, the cracks were found in the heater tube no. 6 and no. 7. To avoid additional damages in the specimen during a decontamination process for the previous analysis, the cracks were analyzed without any decontamination process in this work. We performed the investigation of the ruptured surface morphology, the EDS analysis of the ruptured surface, the microstructural analysis of Alloy 800H sheath tube and literature survey to find the failure mechanism. From the results, it was expected that the sheath tube has been exposed in a steam condition as the coolant level was decreased in the degasser condenser, leading to the rupture of the sheath tube

  2. Realistic analysis of steam generator tube rupture accident in Angra-1 reactor

    International Nuclear Information System (INIS)

    Fontes, S.W.F.

    1989-01-01

    This paper presents the analysis of different scenarios for a Steam Generator Tube Rupture accident (SGTR) in Angra-1 NPP. The results and conclusions will be used as support in the preparation of the emergency situation programs for the plant. For the analysis a SGTR simulation was performed with RETRAN-02 code. The results indicated that the core integrity and the plant itself will not affect by small ruptures in SG tubes. For large ruptures the analysis demonstrated that the accident may have harmful consequences if the operator do not actuate effectively since the initial moments of the accidents. (author) [pt

  3. The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kweon, Young Chul

    2002-01-01

    A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR 1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR 1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet

  4. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  5. Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Larson, T.K.

    1978-01-01

    Tests which simulated rupture of steam generator tubes during loss-of-coolant experiments in a PWR type system have been conducted in the Semiscale Mod-1 system. Analysis of test data indicates that high rod cladding temperatures occured only for a band of tube ruptures (between 12 and 20 tubes) and that the peak cladding temperatures attained within this band were strongly dependent on the magnitude of the tube rupture flow rates. Maximum cladding temperature of about 1255 K was observed for tests which simulated tube ruptures within this narrow band. (author)

  6. Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

    Energy Technology Data Exchange (ETDEWEB)

    Durbec, V.; Pitner, P.; Pages, D. [Electricite de France, 78 - Chatou (France). Research and Development Div.; Riffard, T. [Electricite de France, 69 - Villeurbanne (France). Engineering and Construction Div.; Flesch, B. [Electricite de France, 92 - Paris la Defense (France). Generation and Transmission Div.

    1997-10-01

    Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author) 8 refs.

  7. Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

    International Nuclear Information System (INIS)

    Durbec, V.; Pitner, P.; Pages, D.; Riffard, T.; Flesch, B.

    1997-10-01

    Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author)

  8. Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Larson, T.K.

    1978-01-01

    Examination of the U-tubes in the steam generators of some large commercial pressurized water reactors (PWR) has revealed the existence of leakage and in some cases structural weakening of the tubes. This structural weakening enhances the possibility of tubes rupturing during a hypothesized loss-of-coolant accident (LOCA). Considerable interest has been shown in the analysis of tube ruptures concurrent with a hypothesized LOCA since the presence of tube ruptures has the potential to influence the system thermal-hydraulic response and could foreseeably result in a more severe core thermal behavior than might otherwise occur. To experimentally investigate the influence of steam generator tube ruptures on the thermal-hydraulic response of PWR type system, a series of experiments was conducted in the Semiscale Mod-1 system by EG and G Idaho, Inc., for the U.S. Nuclear Regulatory Commission and the Department of Energy. The primary objective of the experiments was to obtain data which could be used to evaluate the influence of the simulated tube ruptures on the system and core thermal-hydraulic response for a range of tube ruptures that was expected to provide the potential for high cladding temperatures in the Semiscale facility. The experiments were conducted assuming a variety in the number of tubes ruptured during large break loss-of-coolant conditions. The number of experiments conducted permitted determination of the range of tube ruptures for which high peak cladding temperatures could result in the Semiscale Mod-1 system. The paper contains a description of the Semiscale Mod-1 system and a discussion of the steam generator tube rupture tests conducted. The experimental results from the test series and the thermal-hydraulic phenomena found to influence the core thermal response during the experiments are discussed

  9. Pressure tube rupture in a closed tank

    International Nuclear Information System (INIS)

    Khater, H.A.; Hadaller, G.I.; Stern, F.

    1985-06-01

    A study has been prepared on the feasibility of conducting pressure tube/calandria tube rupture tests in a closed tank, simulating a scaled-down calandria vessel. The study includes: i) a review of previous work, ii) an analytical investigation of the scaling problem of the calandria vessel and relevant in-core structures, iii) selection of a method for initiating pressure tube/calandria tube rupture, iv) a set of specifications for the test assembly, v) general arrangement drawings, vi) a proposal for a test matrix, vii) a survey and evaluation of existing facilities which could provide the required high pressure, temperature and fluid inventory, and viii) a cost estimate for the detailed design and construction, instrumentation, data acquisition and reduction, testing and reporting. The study concludes that it is both technically and practically feasible to conduct pressure tube rupture tests in a closed tank

  10. Steam generator tube rupture simulation using extended finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurin; Natesan, Ken

    2016-08-15

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  11. Steam generator tube rupture simulation using extended finite element method

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

    2016-01-01

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  12. Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

    International Nuclear Information System (INIS)

    Liao, Y.; Guentay, S.

    2009-01-01

    This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.

  13. Evaluation of tube rupture simulation test (TRUST-1) for FBR steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Hayashida, Yoshihiko; Hamada, Hirotsugu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-06-01

    The intermediate water leak in an FBR Steam Generator (SG) causes a high temperature and corrosive sodium-water reaction jet. In such cases, it is necessary to evaluate the wastage and overheating rupture behavior of heat transfer tubes. Especially, in the large SG that aims at high temperature of sodium and high temperature/pressure of water, the establishment of the rational evaluation method is important. In this paper, as a basic experiment to make clear the phenomenon of overheating rupture, tests and analysis of Tube Rupture Simulation Test-1 (TRUST-1) were conducted. TRUST-1 simulates the overheating rupture of the tube made of Mod.9Cr-1Mo steel by nitrogen gas pressurization and quick induction heating. The result of TRUST-1 are as follows: (1) The breaking strength predicted by the internal pressure is larger than the tensile strength of the tube material. (2) The margin of the breaking strength from the tensile strength of the tube material has a tendency of decreasing with the heating rate, especially in the lower temperature region. (3) Using an theoretical formula that is deduced from the steady creep model and appropriate experimental coefficients that are determined by the test data, the breaking strength can be reasonably evaluated. (author)

  14. Corrosion and Rupture of Steam Generator Tubings in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-15

    This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned.

  15. Corrosion and Rupture of Steam Generator Tubings in PWRs

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-01

    This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned

  16. Analysis of induced steam generator tube rupture using MAAP 4.0

    International Nuclear Information System (INIS)

    Kenton, M.; Epstein, M.; Henry, R.E.; Paik, C.; Fuller, E.

    1996-01-01

    The nuclear industry has initiated a program of Steam Generator Degradation Specific Management (SGDSM) to cope with the various types of corrosion that have been observed in pressurized water reactor (PWR) steam generators. In parallel, the U.S. Nuclear Regulatory Commission is promulgating revised rules on steam generator tube integrity. To support these efforts, the Electric Power Research Institute has sponsored calculations with the MAAP 4 code. The principal objective of these calculations is to estimate the peak temperatures experienced by the steam generator tubes during high-pressure severe accidents. These results are used to evaluate the potential for degraded tubes to leak or rupture. Attention was focused on station blackout (SBO) accidents with loss of turbine-driven auxiliary feedwater because these generally result in the greatest threat to the tubes

  17. Using MAAP 4.0 to determine risks from steam generator tube leaks or ruptures

    International Nuclear Information System (INIS)

    Fuller, E.L.; Kenton, M.A.

    1996-01-01

    As part of the Electric Power Research Institute (EPRI) program on steam generator degradation specific management (SGDSM), the nuclear industry is investigating the effects on plant risk of severe accidents involving steam generator tube leaks or ruptures. Such accidents fall into three classes: those caused by spontaneous, steam generator tube ruptures (SGTRs) that subsequently result in core damage; those caused by design-basis accidents that lead to induced tube ruptures and subsequent core damage; and those that progress to core damage, such as a station blackout (SBO), with subsequent induced tube leakage or rupture. In each case, the potential exists for a significant fraction of the fission products released from a damaged core to reach the environment through the leaking or ruptured tubes

  18. CFD modeling of a boiler's tubes rupture

    International Nuclear Information System (INIS)

    Rahimi, Masoud; Khoshhal, Abbas; Shariati, Seyed Mehdi

    2006-01-01

    This paper reports the results of a study on the reason for tubes damage in the superheater Platen section of the 320 MW Bisotoun power plant, Iran. The boiler has three types of superheater tubes and the damage occurs in a series of elbows belongs to the long tubes. A three-dimensional modeling was performed using an in-house computational fluid dynamics (CFD) code in order to explore the reason. The code has ability of simultaneous solving of the continuity, the Reynolds-Averaged Navier-Stokes (RANS) equations and employing the turbulence, combustion and radiation models. The whole boiler including; walls, burners, air channels, three types of tubes, etc., was modeled in the real scale. The boiler was meshed into almost 2,000,000 tetrahedral control volumes and the standard k-ε turbulence model and the Rosseland radiation model were used in the model. The theoretical results showed that the inlet 18.9 MPa saturated steam becomes superheated inside the tubes and exit at a pressure of 17.8 MPa. The predicted results showed that the temperature of the steam and tube's wall in the long tubes is higher than the short and medium size tubes. In addition, the predicted steam mass flow rate in the long tube was lower than other ones. Therefore, it was concluded that the main reason for the rupture in the long tubes elbow is changing of the tube's metal microstructure due to working in a temperature higher than the design temperature. In addition, the structural fatigue tension makes the last elbow of the long tube more ready for rupture in comparison with the other places. The concluded result was validated by observations from the photomicrograph of the tube's metal samples taken from the damaged and undamaged sections

  19. The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

    1985-07-01

    On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

  20. A thin-lip rupture of carbon steel superheater boiler tube

    International Nuclear Information System (INIS)

    Khalil, E.O.; Alzoye, K.S.; Elwaer, A.M.

    1993-01-01

    A ruptured A 42 medium carbon steel tube was collected by the engineering department in one of our steam power stations. Inspection of ruptured tube revealed a thin - lip fracture with brownish thin layer of oxide film on inner tube surfaces. There was no evidence of pitting, the outer surfaces of the tube exhibited a general oxidized conditions. A micro section taken near the fracture surface consists of ferrite and martensite, the amount of martensite decreased as we away from the fracture surface. Presence of martensite phase in the microstructure indicates that the tube material has been overheated. An erosion corrosion mechanism in conjunction with overheated. An erosion corrosion mechanism in conjunction with overheating resulted in strength deterioration with consequent premature failure. 4 fig., 1 tab

  1. Steam generator tube rupture (SGTR) scenarios

    International Nuclear Information System (INIS)

    Auvinen, A.; Jokiniemi, J.K.; Laehde, A.; Routamo, T.; Lundstroem, P.; Tuomisto, H.; Dienstbier, J.; Guentay, S.; Suckow, D.; Dehbi, A.; Slootman, M.; Herranz, L.; Peyres, V.; Polo, J.

    2005-01-01

    The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000-2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER)

  2. 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

  3. Study on tube rupture strength evaluation method for rapid overheating

    International Nuclear Information System (INIS)

    Komine, Ryuji; Wada, Yusaku

    1998-08-01

    A sodium-water reaction derived from the single tube break in steam generator might overheat neighbor tubes rapidly under internal pressure loadings. If the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. In the present study this phenomenon was recognized as the fracture of cylindrical tube with the large deformation due to overheating, and the evaluation method was investigated based on both of experimental and analytical approaches. The results obtained are as follows. (1) As for the nominal stress estimation, it was clarified through the experimental data and the detailed FEM elasto-plastic large deformation analysis that the formula used in conventional designs can be applied. (2) Within the overheating temperature limits of tubes, the creep effect is dominant, even if the loading time is too short. So the strain rate on the basis of JIS elevated temperature tensile test method for steels and heat-resisting alloys is too late and almost of total strain is composed by creep one. As a result the time dependent effect cannot be evaluated under JIS strain rate condition. (3) Creep tests in shorter time condition than a few minutes and tensile tests in higher strain rate condition than 10%/min of JIS are carried out for 2 1/4Cr-1Mo(NT) steel, and the standard values for tube rupture strength evaluation are formulated. (4) The above evaluation method based on both of the stress estimation and the strength standard values application is justified by using the tube burst test data under internal pressure. (5) The strength standard values on Type 321 ss is formulated in accordance with the procedure applied for 2 1/4Cr-1Mo(NT) steel. (author)

  4. CFD modeling of a boiler's tubes rupture

    Energy Technology Data Exchange (ETDEWEB)

    Rahimi, Masoud; Khoshhal, Abbas; Shariati, Seyed Mehdi [Chemical Engineering Department, Faculty of Engineering, Razi University, Kermanshah (Iran)

    2006-12-15

    This paper reports the results of a study on the reason for tubes damage in the superheater Platen section of the 320MW Bisotoun power plant, Iran. The boiler has three types of superheater tubes and the damage occurs in a series of elbows belongs to the long tubes. A three-dimensional modeling was performed using an in-house computational fluid dynamics (CFD) code in order to explore the reason. The code has ability of simultaneous solving of the continuity, the Reynolds-Averaged Navier-Stokes (RANS) equations and employing the turbulence, combustion and radiation models. The whole boiler including; walls, burners, air channels, three types of tubes, etc., was modeled in the real scale. The boiler was meshed into almost 2,000,000 tetrahedral control volumes and the standard k-{epsilon} turbulence model and the Rosseland radiation model were used in the model. The theoretical results showed that the inlet 18.9MPa saturated steam becomes superheated inside the tubes and exit at a pressure of 17.8MPa. The predicted results showed that the temperature of the steam and tube's wall in the long tubes is higher than the short and medium size tubes. In addition, the predicted steam mass flow rate in the long tube was lower than other ones. Therefore, it was concluded that the main reason for the rupture in the long tubes elbow is changing of the tube's metal microstructure due to working in a temperature higher than the design temperature. In addition, the structural fatigue tension makes the last elbow of the long tube more ready for rupture in comparison with the other places. The concluded result was validated by observations from the photomicrograph of the tube's metal samples taken from the damaged and undamaged sections. (author)

  5. Dynamic response of INTOR/NET blankets after coolant tube rupture

    International Nuclear Information System (INIS)

    Klippel, H.T.

    1985-01-01

    The dynamic response of different water-cooled liquid Li 17 Pb 83 breeder blanket modules has been calculated to study the potential of these modules in case of coolant tube rupture. Numerical calculations with the code PISCES have been carried out taking into account the fluid-structure interaction and the elasto-plastic behaviour of the structural material. The results show that for inert coolant characteristics the proposed conceptual designs for NET and INTOR have sufficient resistance against coolant tube rupture but when taking into account energy release due to chemical reaction of water with LiPb-alloy up to doubling of the wall thickness has to be envisaged to guarantee structural reliability. (orig.)

  6. Risk assessment of severe accident-induced steam generator tube rupture

    International Nuclear Information System (INIS)

    1998-03-01

    This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC's Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs

  7. Steam generator tubes rupture probability estimation - study of the axially cracked tube case

    International Nuclear Information System (INIS)

    Mavko, B.; Cizelj, L.; Roussel, G.

    1992-01-01

    The objective of the present study is to estimate the probability of a steam generator tube rupture due to the unstable propagation of axial through-wall cracks during a hypothetical accident. For this purpose the probabilistic fracture mechanics model was developed taking into account statistical distributions of influencing parameters. A numerical example considering a typical steam generator seriously affected by axial stress corrosion cracking in the roll transition area, is presented; it indicates the change of rupture probability with different assumptions focusing mostly on tubesheet reinforcing factor, crack propagation rate and crack detection probability. 8 refs., 4 figs., 4 tabs

  8. Tracheal rupture after misplacement of Sengstaken-Blakemore tube ...

    African Journals Online (AJOL)

    The balloon were immediately deflated and a chest X-ray was performed, showing the tube in the right bronchus airway (A), so it was withdrawn. Right pneumothorax appeared and was treated with an intercostal drainage. The patient required orotracheal intubation and a CT scan was performed to show the rupture level ...

  9. Five Tubes Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

    2010-12-01

    In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-CL-02 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Compared to the case of a single U-tube rupture test, opening frequency of the MSSVs in the intact steam generator (SG-2) was highly reduced after 500 seconds in the present SGTR-CL-02 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

  10. Failure analysis of the boiler water-wall tube

    OpenAIRE

    S.W. Liu; W.Z. Wang; C.J. Liu

    2017-01-01

    Failure analysis of the boiler water-wall tube is presented in this work. In order to examine the causes of failure, various techniques including visual inspection, chemical analysis, optical microscopy, scanning electron microscopy and energy dispersive spectroscopy were carried out. Tube wall thickness measurements were performed on the ruptured tube. The fire-facing side of the tube was observed to have experienced significant wall thinning. The composition of the matrix material of the tu...

  11. Consequences of pressure tube rupture on in-core components

    International Nuclear Information System (INIS)

    Hill, P.G.; Hauptmann, E.G.; Lee, V.

    1982-12-01

    An investigation has been made of the consequences of pressure tube rupture in calandria vessels of heavy water cooled and moderated reactors. The study included a review of previous experimental and analytical work, as well as supplementary investigations carried out to examine the validity of previous assumptions and findings. The central questions considered were: the possibility of a propagating pressure tube failure; damage to the calandria vessel; and damage to the shut-off-rod guide tubes of the reactor shut-down system. The results of the investigation do not indicate mechanisms of sufficient strength to cause propagating failure in a well-designed, well-operated reactor following a tube burst under normal operating conditions. However, not all the details of the physical processes involved in a tube burst have been revealed by existing experimental and analytical work

  12. Analysis of multiple-tube ruptures in both steam generators for the Three Mile Island-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1985-01-01

    The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; conditions for isolation of the SGs were not reached; and core subcooling was not lost in either case but the upper head was voided in Case 2. Comparison of the cooldown rates in the two cases after 1200 s (20 min) shows that these rates are equal (i.e., restart of the RCPs did not change the primary-system cooldown rate). However, in Case 2, a steam bubble was formed in the upper head, which did not disappear during the simulated time. One of the immediate actions in the guidelines was to fill both SGs to 95% level. This step was almost unnecessary because the tube-rupture flow was large enough that the 15.4 - mK/s (100 - 0 F h) cooldown-rate limit was exceeded and AFW could not be injected. Also, the guidelines did not address the SG overfill issue

  13. Probability of a steam generator tube rupture due to the presence of axial through wall cracks

    International Nuclear Information System (INIS)

    Mavko, B.; Cizelj, L.

    1991-01-01

    Using the Leak-Before-Break (LBB) approach to define tube plugging criteria a possibility to operate with through wall crack(s) in steam generator tubes may be considered. This fact may imply an increase in tube rupture probability. Improved examination techniques (in addition to the 100% tube examination) have been developed and introduced to counterbalance the associated risk. However no estimates of the amount of total increase or decrease of risk due to the introduction of LBB have been made. A scheme to predict this change of risk is proposed in the paper, based on probabilistic fracture mechanics analysis of axial cracks combined with available data of steam generator tube nondestructive examination reliability. (author)

  14. Creep and creep rupture properties of cladding tube (type 316) in high temperature sodium

    International Nuclear Information System (INIS)

    Atsumo, H.

    1977-01-01

    The thin walled small sized seamless AISI 316 steel tubes, which are designated to be domestically used as the fuel cladding tube for sodium cooled fast breeder reactors in Japan, are irradiated in the following sodium of high temperature in the range of 370 deg. C to 700 deg. C, and receive gradually increased internal pressure caused by the fission produced gas generating from the nuclear fuel burn-up inside the cladding tube. Consequently, the creep behavior of fuel cladding tubes under a high temperature sodium environment is an important problem which must be determined and clarified together with their characteristic features under irradiation and in air. In relation to the creep performance of fuel cladding tubes made of AISI 316 steel and other comparable austenitic stainless steels, hardly any studies are found that are made systematically to examine the effect of sodium with sodium purity as parameter or any comparative studies with in-air data at various different temperatures. The present research work was aimed to obtain certain basic design data relating to in-sodium creep performance of the domestic made fuel cladding tubes for fast breeder reactors, and also to gain further date as considered necessary under several sodium conditions. That is, together with establishment of the technology for tensile creep test and internal pressure creep rupture test in flowing sodium of high temperature, a series of tests and studies were performed on the trial made cladding tubes of AISI Type-316 steel. In the first place, two kinds of purity conditions of sodium, close to the actual reactor-operating condition, (oxygen concentration of 10 ppm and 5 ppm respectively) were established, and then uniaxial tensile creep test and rupture test under various temperatures were performed and the resulting data were compared and evaluated against the in-air data. Then, secondly, an internal pressure creep rupture test was conducted under a single purity sodium environment

  15. Five Tubes Rupture at Hot Side of Steam Generator Simulation Test Report Using the ATLAS

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Hyun Sik; Cho Seok

    2010-12-01

    In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-HL-05 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the SPACE code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. On the contrary to the case of a single U-tube rupture test, the MSSV of the intact steam generator was not opened any more after 1500 seconds in the present SGTR-HL-05 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

  16. CATHENA simulations of steam generator tube rupture

    International Nuclear Information System (INIS)

    Abdul-Razzak, A.; Lin, M.R.; Wright, A.C.D.

    1997-01-01

    The CATHENA thermalhydraulic computer code was used to simulate various scenarios following a CANDU 9 steam generator tube rupture (SGTR) event. The analysis included cases with class IV power and emergency core cooling system (ECCS) available and other cases with subsequent loss of class IV power (LCIVP) or impairment of ECCS injection. Two main approaches were followed in the analysis of each case. In the first approach, D 2 O feed was credited to provide conservative data for input to radionuclide release and dose calculations. Also operator actions are credited. The other approach is designed to give conservative predictions with respect to the acceptance criteria of fuel and fuel channel integrity and to prove that in case of such event, the operator will have enough time to mitigate the consequences. This is done by not crediting makeup for the inventory loss and relying on the automatic operation of safety systems. The analysis of the cases of the first approach provided the required data for radionuclide release and dose calculations and gave a good insight into the required sequence of operator timely actions to mitigate the consequences of such event. On the other hand, the cases of the second approach confirmed compliance with regulatory requirements for pressure tube and fuel integrity. The runs with ECCS available, showed the ECCS injection is effective in filling and cooling the core and that regulatory requirement's for fuel and channel integrity are met. In the event of ECCS impairment, the earliest indication of late fuel heat-up is late enough to provide the operator with an adequate time to act in mitigating the consequences of this event. (author)

  17. Esophageal rupture caused by explosion of an automobile tire tube: a case report.

    Science.gov (United States)

    Yu, Yongkang; Ding, Sheng; Zheng, Yifeng; Li, Wei; Yang, Lie; Zheng, Xiushan; Liu, Xiaoyan; Jiang, Jianqing

    2013-08-23

    There have been no reports in the literature of esophageal rupture in adults resulting from an explosion of an automobile tire. We report the first case of just such an occurrence after an individual bit into a tire, causing it to explode in his mouth. A 47-year-old Han Chinese man presented with massive hemorrhage in his left eye after he accidentally bit an automobile tire tube which burst into his mouth. He was diagnosed with esophageal rupture based on a chest computed tomography scan and barium swallow examination. Drainage of empyema (right chest), removal of thoracic esophagus, exposure of cervical esophagus, cardiac ligation and gastrostomy were performed respectively. After that, esophagogastrostomy was performed. Successful anastomosis was obtained at the neck with no postoperative complications 3 months after the surgery. The patient was discharged with satisfactory outcomes. We present this case report to bring attention to esophageal rupture in adults during the explosion of an automobile tire tube in the mouth.

  18. Steam generator tube rupture risk impact on design and operation of French PWR plants

    International Nuclear Information System (INIS)

    Depond, G.; Sureau, H.

    1984-01-01

    The experience of steam generator tube leaks incidents in PWR plants has resulted in an increase of EDF analysis leading to improvements in design and post-accidental operation for new projects and operating plants. The accident consequences are minimized for each of the NSSS three barriers: first barrier: safeguard systems design and operating procedures relying upon core safety allow to maintain a low level of primary radioactivity, second barrier: steam generator design and periodic inspection allow to reduce tube ruptures risks and third barrier: atmospheric releases are reduced as a result of optimal recovery procedures, detection improvements and atmospheric steam valves design improvements. (orig.)

  19. Failure analysis of the boiler water-wall tube

    Directory of Open Access Journals (Sweden)

    S.W. Liu

    2017-10-01

    Full Text Available Failure analysis of the boiler water-wall tube is presented in this work. In order to examine the causes of failure, various techniques including visual inspection, chemical analysis, optical microscopy, scanning electron microscopy and energy dispersive spectroscopy were carried out. Tube wall thickness measurements were performed on the ruptured tube. The fire-facing side of the tube was observed to have experienced significant wall thinning. The composition of the matrix material of the tube meets the requirements of the relevant standards. Microscopic examinations showed that the spheroidization of pearlite is not very obvious. The failure mechanism is identified as a result of the significant localized wall thinning of the boiler water-wall tube due to oxidation.

  20. Steam generator tube failures

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

  1. Indian Point 2 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    Dayan, A.

    1985-01-01

    Analyses were conducted with RETRAN-02 to study consequences of steam generator tube rupture (SGTR) events. The Indian Point, Unit 2, power plant (IP2, PWR) was modeled as a two asymmetric loops, consisting of 27 volumes and 37 junctions. The break section was modeled once, conservatively, as a 150% flow area opening at the wall of the steam generator cold leg plenum, and once as a 200% double-ended tube break. Results revealed 60% overprediction of breakflow rates by the traditional conservative model. Two SGTR transients were studied, one with low-pressure reactor trip and one with an earlier reactor trip via over temperature ΔT. The former is more typical to a plant with low reactor average temperature such as IP2. Transient analyses for a single tube break event over 500 seconds indicated continued primary subcooling and no need for steam line pressure relief. In addition, SGTR transients with reactor trip while the pressurizer still contains water were found to favorably reduce depressurization rates. Comparison of the conservative results with independent LOFTRAN predictions showed good agreement

  2. Code Assessment of SPACE 2.19 using LSTF Steam Generator Tube Rupture Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The SPACE is a best estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. The present work is on the line of expanding the work by Kim et al. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run LSTF SB-SG-06 experiment simulating a Steam Generator Tube Rupture (SGTR) transient. In order to identify the predictability of SPACE 2.19, the LSTF steam generator tube rupture test was simulated. To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR and the RELAP5/ MOD3.1 are used. The calculation results indicate that the SPACE 2.19 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator relief valve.

  3. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

  4. Development of tube rupture evaluation code for FBR steam generator (II). Modification of heat transfer model in sodium side

    International Nuclear Information System (INIS)

    Hamada, H.; Kurihara, A.

    2003-05-01

    The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1) The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in comparison with the previous method. (2) The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3) The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4) In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5) In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model. (author)

  5. The stress-rupture behavior of tubes made from austenitic stainless steels and Ni-based alloys subjected to internal pressure

    International Nuclear Information System (INIS)

    Schaefer, L.; Kempe, H.

    1983-12-01

    The report outlines the stress-rupture results obtained on tubes tested as possible fuel rod cladding tubes for fast breeder reactors cooled with sodium, steam or gas. For the rupture elongations of some specimens showing a pronounced burst, higher values than in earlier reports are now indicated because of better evaluation techniques. The choice and comparisons of materials are explained, the calculations of stresses and strains are described, and reference is made to the own studies carried out to date of the parameters influencing creep-rupture behaviour. Minor modifications of the composition of an alloy and of the mechanical-thermal treatment of materials, respectively, are seen to produce clearcut changes in the stress-rupture properties. (orig.) [de

  6. Comparative Analyses on OPR1000 Steam Generator Tube Rupture Event Emergency Operational Guideline

    International Nuclear Information System (INIS)

    Lee, Sang Won; Bae, Yeon Kyoung; Kim, Hyeong Teak

    2006-01-01

    The Steam Generator Tube Rupture (SGTR) event is one of the important scenarios in respect to the radiation release to the environment. When the SGTR occurs, containment integrity is not effective because of the direct bypass of containment via the ruptured steam generator to the MSSV and MSADV. To prevent this path, the Emergency Operational Guideline of OPR1000 indicates the use of Turbine Bypass Valves (TBVs) as an effective means to depressurize the main steam line and prevent the lifting of MSSV. However, the TBVs are not operable when the offsite power is not available (LOOP). In this situation, the RCS cool-down is achieved by opening the both intact and ruptured SG MSADV. But this action causes the large amount of radiation release to the environment. To minimize the radiation release to the environment, KSNP EOG adopts the improved strategy when the SGTR concurrently with LOOP is occurred. However, these procedures show some duplicated procedure and branch line that might confusing the operator for optimal recovery action. So, in this paper, the comparative analysis on SGTR and SGTR with LOOP is performed and optimized procedure is proposed

  7. Analysis code for medium and small rupture accidents in ATR. LOTRAC/HEATUP

    International Nuclear Information System (INIS)

    1997-08-01

    In the evaluation of thermo-hydraulic and fuel temperature transient changes in the events which are classified in medium and small rupture accidents of reactor coolant loss that is the safety evaluation event of the ATR, the analysis code for synthetic thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC and the detailed analysis code for fuel temperature HEATUP are used, respectively. By using the LOTAC, the thermo-hydraulic behavior of reactor cooling facility and the temperature behavior of fuel at the time of blow-down are analyzed, and also the characteristics of changing reactor thermal output is analyzed, considering the functioning characteristics of emergency core cooling system. Based on the data of thermo-hydraulic behavior obtained by the LOTRAC, the time of beginning the turn-around of fuel cladding tube temperature obtained by the data of ECCS pouring characteristics, the heat transfer rate after the turn-around and so on, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The LOTRAC code, the HEATUP code, various analysis models, and rupture simulation experiment are reported. (K.I.)

  8. Radioactivity transport following steam generator tube rupture

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-03-01

    A review of the capabilities of the CITADEL computer code as well as plant experience to project radioactivity releases following a steam generator tube rupture in PWR's shows that certain experimental data are needed for reliable off-site dose predictions. This article defines five parameters which are the key for such predictions and discusses the functional dependence of these parameters on various operational variables. The above parameters can be used in conjunction with CITADEL or they can be inserted in the appropriate equations which then conveniently can be programmed as a subroutine in thermal-hydraulic system codes. A joint Westinghouse, Electric Power Research Institute and Nuclear Regulatory Commission Program aimed at obtaining the five parameters empirically is described

  9. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    International Nuclear Information System (INIS)

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  10. Fission product retention during faults involving steam generator tube rupture

    International Nuclear Information System (INIS)

    Rodliffe, R.S.

    1983-08-01

    In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

  11. CITADEL: a computer code for the analysis of iodine behavior in steam generator tube rupture accidents

    International Nuclear Information System (INIS)

    1982-04-01

    The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the code including its input requirements and the nature and form of its output. A user's guide describing the manner in which the input data are required to be set up to run the code is also provided

  12. Analysis of Communication between Main Control Room Operators in Decision-making Process in Steam Generator Tube Rupture Accident

    International Nuclear Information System (INIS)

    Petkov, M.; Petkov, G.

    2006-01-01

    The paper presents an investigation results for Main Control Room operators' reliability by Performance Evaluation of Teamwork method, based on FSS-1000 training archives in KNPP in case of Steam Generator Tube Rupture accident. The advantages of operators' teamwork are shown: a) group decision-making vs. individual one: b) positive influence of crew initiated communication consisting of orders and reports that are required by instruction. (authors)

  13. Evaluation of the cranial base in amnion rupture sequence involving the anterior neural tube: implications regarding recurrence risk.

    Science.gov (United States)

    Jones, Kenneth Lyons; Robinson, Luther K; Benirschke, Kurt

    2006-09-01

    Amniotic bands can cause disruption of the cranial end of the developing fetus, leading in some cases to a neural tube closure defect. Although recurrence for unaffected parents of an affected child with a defect in which the neural tube closed normally but was subsequently disrupted by amniotic bands is negligible; for a primary defect in closure of the neural tube to which amnion has subsequently adhered, recurrence risk is 1.7%. In that primary defects of neural tube closure are characterized by typical abnormalities of the base of the skull, evaluation of the cranial base in such fetuses provides an approach for making a distinction between these 2 mechanisms. This distinction has implications regarding recurrence risk. The skull base of 2 fetuses with amnion rupture sequence involving the cranial end of the neural tube were compared to that of 1 fetus with anencephaly as well as that of a structurally normal fetus. The skulls were cleaned, fixed in 10% formalin, recleaned, and then exposed to 10% KOH solution. After washing and recleaning, the skulls were exposed to hydrogen peroxide for bleaching and photography. Despite involvement of the anterior neural tube in both fetuses with amnion rupture sequence, in Case 3 the cranial base was normal while in Case 4 the cranial base was similar to that seen in anencephaly. This technique provides a method for determining the developmental pathogenesis of anterior neural tube defects in cases of amnion rupture sequence. As such, it provides information that can be used to counsel parents of affected children with respect to recurrence risk.

  14. Analysis of an accident with the main circulation tube rupture at the WWER-1000

    International Nuclear Information System (INIS)

    Boyadzhiev, A.I.; Stefanova, S.J.

    1984-01-01

    In connection with the forthcoming construction of a npp with the wwer-1000 reactor the loss of coolant accident associated with the main circulation tube rupture at the inlet near the reactor is analyzed. The relap4/mod6 program is used for the analysis. The data obtained show that the coolant outflow stage continues for about 25s. On the average the pressure in the circuits varies from 16 to 10 mpa per 0.1s and then it continues to decrease slowly. The pressure in the steam generator at the secondary circuits end increases approximately up to 6.9 MPa as a result of steam generator blocking and remaining coolant heating and then somewhat decreases owing to the primary circuit cooling. By the end of the fuel and can temperatures are equalized and the heat transfer coefficient is stabilized at the level of 100 w/1 (m 2 xK). It is concluded that during a loss of coolant accident at the wwer-1000 reactor in procesess of coolant blowdown in the medium power fuel elemets neither the fuel, melting temperature (3000 k), nor the critical temperature (1000 k) of plastic deformation zirconiu can initiation are attained

  15. Early Rupture of an Ultralow Duodenal Stump after Extended Surgery for Gastric Cancer with Duodenal Invasion Managed by Tube Duodenostomy and Cholangiostomy

    Directory of Open Access Journals (Sweden)

    Konstantinos Blouhos

    2013-01-01

    Full Text Available When dealing with gastric cancer with duodenal invasion, gastrectomy with distal resection of the duodenum is necessary to achieve negative distal margin. However, rupture of an ultralow duodenal stump necessitates advanced surgical skills and close postoperative observation. The present study reports a case of an early duodenal stump rupture after subtotal gastrectomy with resection of the whole first part of the duodenum, complete omentectomy, bursectomy, and D2+ lymphadenectomy performed for a pT3pN2pM1 (+ number 13 lymph nodes adenocarcinoma of the antrum. Duodenal stump rupture was managed successfully by end tube duodenostomy, without omental patching, and tube cholangiostomy. Close assessment of clinical, physical, and radiological signs, output volume, and enzyme concentration of the tube duodenostomy, T-tube, and closed suction drain, which was placed near the tube duodenostomy site to drain the leak around the catheter, dictated postoperative management of the external duodenal fistula.

  16. Steam generator secondary pH during a steam generator tube rupture

    International Nuclear Information System (INIS)

    Adams, J.P.; Peterson, E.S.

    1991-12-01

    The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL to perform an analytical assessment of secondary coolant system (SCS) pH during an SGTR. Design basis thermal and hydraulic calculations were used together with industry standard chemistry guidelines to determine the SCS chemical concentrations during an SGTR. These were used as input to the Facility for Analysis of Chemical Thermodynamics computer system to calculate the equilibrium pH in the SCS at various discrete time during an SGTR. The results of this analysis indicate that the SCS pH decreases from the initial value of 8.8 to approximately 6.5 by the end of the transient, independent of PWR design

  17. Stress analysis of HTR-10 steam generator heat exchanging tubes

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

    Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

  18. Creep rupture properties of solution annealed and cold worked type 316 stainless steel cladding tubes

    International Nuclear Information System (INIS)

    Mathew, M.D.; Latha, S.; Mannan, S.L.; Rodriguez, P.

    1990-01-01

    Austenitic stainless steels (mainly type 316 and its modifications) are used as fuel cladding materials in all current generation fast breeder reactors. For the Fast Breeder Test Reactor (FBTR) at Kalpakkam, modified type 316 stainless steel (SS) was chosen as the material for fuel cladding tubes. In order to evaluate the influence of cold work on the performance of the fuel element, the investigation was carried out on cladding tubes in three metallurgical conditions (solution annealed, ten percent cold worked and twenty percent cold worked). The results indicate that: (i) The creep strength of type 316 SS cladding tube increases with increasing cold work. (ii) The benificial effects of cold work are retained at almost all the test conditions investigated. (iii) The Larson Miller parameter analysis shows a two slope behaviour for 20PCW material suggesting that caution should be exercised in extrapolating the creep rupture life to stresses below 125 MPa. At very low stress levels, the LMP values fall below the values of the 10 PCW material. (author). 6 refs., 19 figs. , 10 tabs

  19. Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS/SMR-S code

    International Nuclear Information System (INIS)

    Kim, Hee-Kyung; Kim, Soo Hyoung; Chung, Young-Jong; Kim, Hyeon-Soo

    2013-01-01

    Highlights: ► The analysis was performed from the viewpoint of primary coolant leakage. ► The thermal hydraulic responses and the maximum leakage have been identified. ► There is no direct release into the atmosphere caused by an SGTR accident. ► SMART safety system works well against an SGTR accident. - Abstract: A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident

  20. Long-term creep rupture strength of weldment of Fe-Ni based alloy as candidate tube and pipe for advanced USC boilers

    Energy Technology Data Exchange (ETDEWEB)

    Bao, Gang; Sato, Takashi [Babcok-Hitachi K.K., Hiroshima (Japan). Kure Research Laboratory; Marumoto, Yoshihide [Babcok-Hitachi K.K., Hiroshima (Japan). Kure Div.

    2010-07-01

    A lot of works have been going to develop 700C USC power plant in Europe and Japan. High strength Ni based alloys such as Alloy 617, Alloy 740 and Alloy 263 were the candidates for boiler tube and pipe in Europe, and Fe-Ni based alloy HR6W (45Ni-24Fe-23Cr-7W-Ti) is also a candidate for tube and pipe in Japan. One of the Key issues to achieve 700 C boilers is the welding process of these alloys. Authors investigated the weldability and the long-term creep rupture strength of HR6W tube. The weldments were investigated metallurgically to find proper welding procedure and creep rupture tests are ongoing exceed 38,000 hours. The long-term creep rupture strengths of the HST weld joints are similar to those of parent metals and integrity of the weldments was confirmed based on with other mechanical testing results. (orig.)

  1. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  2. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

    1997-01-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  3. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  4. Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test)

    International Nuclear Information System (INIS)

    Gillins, R.L.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-3 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-3 was conducted from initial conditions of 15621 kPa and 555 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

  5. Experiment data report for semiscale Mod-1 test S-28-2 (steam generator tube rupture test)

    International Nuclear Information System (INIS)

    Patton, M.L.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-2 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-2 was conducted from initial conditions of 15 936 kPa and 558 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. For Test S-28-2, accumulator injection into the intact loop hot leg was provided to simulate simulate the rupture of six steam generator tubes

  6. Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

    International Nuclear Information System (INIS)

    Esparza, V.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

  7. Preliminary evaluation of steam generator tube rupture (SGTR) accident in lead cooled reactor

    International Nuclear Information System (INIS)

    Frano, R. Lo; Forasassi, G.

    2009-01-01

    In this paper some contributions are provided to the development of a European Lead-cooled System, known as the ELSY project (within EU-6 Framework Project); that will constitute a possible reference system for a large lead-cooled reactor of GEN IV. Steam generator (SG) tubing of this system type might be subject to a variety of degradation processes, such as cracking, wall thinning and potential leakage or rupture, eventually leading to the failure of one or more SG tubes that constitute a steam generator tube rupture (SGTR) accident with possible consequences for the safety of the primary systems. It is therefore of interest for the designer to know how the SG itself, as well as the vessel and internals structures, behave under impulsive loading conditions (in form of a rapid and strong increase of pressure) that can arise as consequences of the interaction between the primary and secondary coolants (lead-water interaction). The analysed initiator event, as already mentioned, is a large break (up to a double ended guillotine break) of one (or more) SG cooling tubes that may become severe enough to determine dangerous effects on the interested structures. In order to better simulate and perform the mentioned postulated SGTR accident sequence analyses, an appropriate numerical model with the available computing resources (FEM codes) was set up at the DIMNP of Pisa University. That model was used to evaluate the effects of the propagation of the blast pressure waves inside the SG structures, taking into account also the sloshing phenomenon that could be induced by the lead primary coolant motions. Therefore the SGTR effects study may be considered as a transient and non linear problem the solution of which provides the 'time histories' of hydrodynamic pressures and stresses on the reactor pressure vessel and internals walls. (author)

  8. Radioactivity release vs probability for a steam generator tube rupture accident

    International Nuclear Information System (INIS)

    Buslik, A.J.; Hall, R.E.

    1978-01-01

    A calculation of the probability of obtaining various radioactivity releases from a steam generator tube rupture (SGTR) is presented. The only radioactive isotopes considered are Iodine-131 and Xe-133. The particular accident path considered consists of a double-ended guillotine SGTR followed by loss of offsite power (LOSP). If there is no loss of offsite power, and no system fault other than the SGTR, it is judged that the consequences will be minimal, since the amount of iodine released through the condenser air ejector is expected to be quite small; this is a consequence of the fact that the concentration of iodine in the vapor released from the condenser air ejector is very small compared to that dissolved in the condensate water. In addition, in some plants the condenser air ejector flow is automatically diverted to containment or a high-activity alarm. The analysis presented here is for a typical Westinghouse PWR such as described in RESAR-3S

  9. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Sul; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Kim, Tae Wan [Incheon National University, Incheon (Korea, Republic of)

    2017-03-15

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

  10. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    International Nuclear Information System (INIS)

    Lee, Han Sul; Heo, Gyun Young; Kim, Tae Wan

    2017-01-01

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective

  11. Experimental investigation of coolant and poisoned moderator mixing due to a simulated pressure tube/calandria tube fishmouth rupturing an overpoisoned guaranteed shutdown state

    International Nuclear Information System (INIS)

    Mackinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

    1997-01-01

    During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The contents of the paper are as follows. First, the objectives of the experimental program are

  12. Analysing The Thermalhydraulic Parameters Of VVER-1000 Reactor For The Accident Of Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Luu Nam Hai; Truong Cong Thang

    2011-01-01

    To ensure the safety operation of nuclear power plant (NPP), a lot of postulated accident scenarios were considered and analysed. This research chose and analysed the accident of steam generator tube rupture (SGTR) under the actual plant conditions by using the simulation program PCTRAN. The SGTR accident is happen when the NPP is under operation with the steady state condition (power of 3000 MWth, primary pressure of 157 bar and secondary pressure of 63 bar). The accident is initiated by creating a break with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes. The result of analysis is compared with the calculation of the Shariz University, Iran using the thermal hydraulics code RELAP5/mod3.2 and the report in the PSAR data of VVER-1000. This comparison shows that it is possible for using PCTRAN to analyse accidents of VVER-1000 reactor. (author)

  13. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Parrish, K.R.

    1995-09-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2.

  14. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    International Nuclear Information System (INIS)

    Parrish, K.R.

    1995-01-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2

  15. Intentional back flow effects on ruptured steam generator cooldown during a SGTR event for KSNP

    International Nuclear Information System (INIS)

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

    2004-01-01

    For an optimum recovery from a steam generator tube rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube as early as possible to minimize the radioactive material release. However, the reactor coolant system (RCS) cooldown and depressurization to the shutdown cooling system (SCS) operation conditions using the intact SG only are hard to achieve unless the ruptured SG is properly cooled since the ruptured SG, which is isolated by operator, remains at high temperature even though the RCS has been cooled down. The effects of intentional back flow from the SG secondary side to the RCS through the ruptured U-tube on the the ruptured SG cooldown were evaluated for the pressurized light water reactor, especially for the Korean standard nuclear power plant (KSNP). In order to evaluate the back flow effect, a series of analyses was conducted using the RELAP5/MOD3 computer code. For the first stage of the analysis, the cooldown process by natural circulation in the SG secondary side was simulated for the initial conditions of the ruptured SG cooldown. In the next analysis stage, two methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated. One utilizes the steam condensation on the uncovered U-tube surface, and the other is a SG drain and fill. In the former method, SG tubes are exposed to the steam space by draining SG secondary water into the RCS in order to condense the steam directly onto the uncovered tubes. This method showed that the steam condensation decreased SG secondary pressure and temperature rapidly, demonstrating its effectiveness for cooling. However, this process has a limited applicability if the rupture is located at the lower region. The latter method, draining by back flow and filling using the feedwater system was also found to be effective in ruptured SG cooldown and depressurization even if the rupture occurred at the top of the U-tube. It is concluded that the

  16. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    International Nuclear Information System (INIS)

    Majumdar, S.; Kasza, K.

    2009-01-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  17. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.; Kasza, K. [Argonne National Laboratory, Nuclear Energy Division, Lemont, Illinois (United States)

    2009-07-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  18. Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

    International Nuclear Information System (INIS)

    Collins, B.L.; Coppin, C.E.; Sackett, K.E.

    1977-10-01

    Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-1 was conducted from initial conditions of 15 767 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

  19. Application of probabilistic fracture mechanics to estimate the risk of rupture of PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pitner, P.; Riffard, T.; Granger, B.

    1992-01-01

    This paper describes the COMPROMIS code developed by Electricite de France (EDF) to optimize the tube bundle maintenance of steam generators. The model, based on probabilistic fracture mechanics, makes it possible to quantify the influence of in-service inspections and maintenance work on the risk of an SG tube rupture, taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive detection and sizing, crack initiation and propagation, critical sizes, leak before risk of break, etc.). (authors). 5 refs., 8 figs., 3 tabs

  20. Cellular response of healing tissue to DegraPol tube implantation in rabbit Achilles tendon rupture repair: an in vivo histomorphometric study.

    Science.gov (United States)

    Buschmann, Johanna; Meier-Bürgisser, Gabriella; Bonavoglia, Eliana; Neuenschwander, Peter; Milleret, Vincent; Giovanoli, Pietro; Calcagni, Maurizio

    2013-05-01

    In tendon rupture repair, improvements such as higher primary repair strength, anti-adhesion and accelerated healing are needed. We developed a potential carrier system of an electrospun DegraPol tube, which was tightly implanted around a transected and conventionally sutured rabbit Achilles tendon. Histomorphometric analysis of the tendon tissue 12 weeks postoperation showed that the tenocyte density, tenocyte morphology and number of inflammation zones were statistically equivalent, whether or not DegraPol tube was implanted; only the collagen fibres were slightly less parallelly orientated in the tube-treated case. Comparison of rabbits that were operated on both hind legs with ones that were operated on only one hind leg showed that there were significantly more inflammation zones in the two-leg cases compared to the one-leg cases, while the implantation of a DegraPol tube had no such adverse effects. These findings are a prerequisite for using DegraPol tube as a carrier system for growth factors, cytokines or stem cells in order to accelerate the healing process of tendon tissue. Copyright © 2012 John Wiley & Sons, Ltd.

  1. 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)

  2. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    International Nuclear Information System (INIS)

    Mendler, O.J.; Takeuchi, K.; Young, M.Y.

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

  3. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mendler, O J; Takeuchi, K; Young, M Y

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

  4. Assessment of a Pressure Tube Rupture with a Poisoned Moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, S. C.; Kim, E. K.

    2005-01-01

    The postulated in-core LOCA has been analyzed and evaluated while the reactor is operating normally with a low moderator poison concentration for CANDU. However, when the reactor is operating with a relatively large amount of boron and/or gadolinium poison in the moderator, an assessment of the fuel integrity was required for the pressure tube rupture (PTR) accident. Poisoned moderator exists mainly during a startup after a prolonged shutdown lasting for more than one day. For the case of a reactor regulating system (RRS) working, the methodology of the PTR assessment with a poisoned moderator has been developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for the Wolsong Nuclear Power Plants recently. The developed methodology and results are presented

  5. 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)

  6. Analytical and experimental studies of mechanical consequences of a steam generator tube rupture

    International Nuclear Information System (INIS)

    Duc, B.; Sudreau, F.; Rassineux, B.

    1995-01-01

    Concerning to steam generator tubes support mechanical loadings due to the impact f the ruptured one, Electricite de France, with the support of Commissariat a l'Energie. Atomique, has undertaken a large study in order to evaluate the consequences of such loadings. This paper first presents the results of the tests performed on AQUITAINE 2 facility (CEA Cadarache research center) for nominal, faulted and boiler effect conditions. Those results are then compared with numerical dynamic elastoplastic analyses performed with CASTEM 2000 code (CEA system). (author). 1 ref., 14 figs

  7. 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.

  8. Experience and modeling of radioactivity transport following steam generator tube rupture

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-01-01

    A review of the capabilities of the CITADEL computer code as well as plant experience to project radioactivity releases following a steam generator tube rupture in pressurized-water reactors shows that certain experimental data are needed for reliable offsite dose predictions. This article defines five parameters that are the key for such predictions and discusses the functional dependence of these parameters on various operational variables. The above parameters can be used in conjuction with CITADEL or they can be inserted in the appropriate equations, which then can be programmed conveniently as a subroutine in thermal-hydraulic system codes. A joint Westinghouse Electric Corporation, Electric Power Research Institute, and Nuclear Regulatory Commission program aimed at obtaining the five parameters empirically is described

  9. Investigation of a steam generator tube rupture sequence using VICTORIA

    International Nuclear Information System (INIS)

    Bixler, N.E.; Erickson, C.M.; Schaperow, J.H.

    1995-01-01

    VICTORIA-92 is a mechanistic computer code for analyzing fission product behavior within the reactor coolant system (RCS) during a severe reactor accident. It provides detailed predictions of the release of radionuclides and nonradioactive materials from the core and transport of these materials within the RCS. The modeling accounts for the chemical and aerosol processes that affect radionuclide behavior. Coupling of detailed chemistry and aerosol packages is a unique feature of VICTORIA; it allows exploration of phenomena involving deposition, revaporization, and re-entrainment that cannot be resolved with other codes. The purpose of this work is to determine the attenuation of fission products in the RCS and on the secondary side of the steam generator in an accident initiated by a steam generator tube rupture (SGTR). As a class, bypass sequences have been identified in NUREG-1150 as being risk dominant for the Surry and Sequoyah pressurized water reactor (PWR) plants

  10. Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum, E-mail: bahn@pusan.ac.kr [Pusan National University, 2 Busandaehak-ro 63 beon-gil, Geumjeong-gu, Busan 609-735 (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering & Construction Co. Inc., Seongnam 463-870 (Korea, Republic of); Majumdar, Saurin [Argonne National Laboratory, Lemont, IL 60439 (United States)

    2015-11-15

    Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

  11. Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Oh, Young-Jin; Majumdar, Saurin

    2015-01-01

    Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

  12. NRC concerns about steam generator tube U-bend failures

    International Nuclear Information System (INIS)

    Dillon, R.L.

    1981-01-01

    This paper concerns itself with genralized NRC regulatory policy regarding SGT failures and staff reports and opinions which may tend to influence the developing policy specific to U-bend failures. The most significant analysis at hand in predicting NRC policy on SGT U-bend failures is Marsh's Evaluation of Steam Generator Tube Rupture Events. Marsh sets out to describe and analyze the five steam generator tube ruptures that are known to NRC. All have occurred in the period 1975 to 1980

  13. Analysis of steam generator tube rupture as a severe accident using MELCOR 1.8.4

    International Nuclear Information System (INIS)

    Yang Hongrun; Hidaka, Akihide; Sugimoto, Jun

    1999-03-01

    This report presents the results from the MELCOR 1.8.4 calculations for Steam Generator Tube Rupture (SGTR) with stuck open of all the safety valves in faulted SG as a severe accident. The calculations are based on Surry nuclear power plant. After performed using the once-through primary system model alone by 1.0x10 5 s, the calculations were conducted with both of the once-through and the hot leg countercurrent natural circulation models. The results, including event sequences, processes and progressions of core degradation, radionuclides release from core and reactor cavity, and source terms to the environment are described in detail. It is concluded that the availability of High Pressure Safety Injection (HPSI) can significantly delay the progression of core heat-up and approximately 7% of cesium iodide (CsI) can be released to the environment directly through the stuck open safety valve. Comparisons between the results from the two models are also given in this report. The present analyses also showed that during SGTR accident, the hot leg countercurrent natural circulation flow cannot be established well and therefore it has little effect on the mitigation of the core degradation. (author)

  14. The creep and stress-rupture behaviour under internal pressure of tubes made from austenitic stainless steel X8 CrNiMoNb 1616 (Material No. 1.4981)

    International Nuclear Information System (INIS)

    Schaefer, L.; Polifka, F.; Kempe, H.

    1979-05-01

    Creep and stress rupture tests have been performed at 600, 650, 700 and 750 0 C on tubes made from three different heats from the austenitic stainless steel X8 CrNiMoNb 1616 (Material No. 1.4981). The tubes were loaded by internal pressure and the tangential (hoop) creep strain was measured continuously. The results are presented in form of creep curves, stress-time to rupture curves and curves for a creep limit. The average and minimum creep rates as a function of the applied stress have been evaluated and are described with a creep law analogous to Norton's creep law. An interpolation and extrapolation of the stress-rupture-strength and the creep strength are possible using the time-temperature-parameter-plot after Larson and Miller. (orig.) [de

  15. Method of experimental and theoretical modeling for multiple pressure tube rupture for RBMK reactor

    International Nuclear Information System (INIS)

    Medvedeva, N.Y.; Goldstein, R.V.; Burrows, J.A.

    2001-01-01

    The rupture of single RBMK reactor channels has occurred at a number of stations with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighbouring channels to break. This assumption has not been justified. A chain reaction of tube breaks could over-pressurise the reactor cavity leading to catastrophic failure of the containment. To validate the claims of the RBMK Safety Cases the Electrogorsk Research and Engineering Centre, in participation with experts from the Institute of Mechanics of RAS, has developed the method of interacting multiscale physical and mathematical modelling for coupled thermophysical, hydrogasodynamic processes and deformation and break processes causing and (or) accompanying potential failures, design and beyond the design RBMK reactor accidents. To realise the method the set of rigs, physical and mathematical models and specialized computer codes are under creation. This article sets out an experimental philosophy and programme for achieving this objective to solve the problem of credibility or non-credibility for multiple fuel channel rupture in RBMK.(author)

  16. A Retrospective Analysis of Ruptured Breast Implants

    Directory of Open Access Journals (Sweden)

    Woo Yeol Baek

    2014-11-01

    Full Text Available BackgroundRupture is an important complication of breast implants. Before cohesive gel silicone implants, rupture rates of both saline and silicone breast implants were over 10%. Through an analysis of ruptured implants, we can determine the various factors related to ruptured implants.MethodsWe performed a retrospective review of 72 implants that were removed for implant rupture between 2005 and 2014 at a single institution. The following data were collected: type of implants (saline or silicone, duration of implantation, type of implant shell, degree of capsular contracture, associated symptoms, cause of rupture, diagnostic tools, and management.ResultsForty-five Saline implants and 27 silicone implants were used. Rupture was diagnosed at a mean of 5.6 and 12 years after insertion of saline and silicone implants, respectively. There was no association between shell type and risk of rupture. Spontaneous was the most common reason for the rupture. Rupture management was implant change (39 case, microfat graft (2 case, removal only (14 case, and follow-up loss (17 case.ConclusionsSaline implants have a shorter average duration of rupture, but diagnosis is easier and safer, leading to fewer complications. Previous-generation silicone implants required frequent follow-up observation, and it is recommended that they be changed to a cohesive gel implant before hidden rupture occurs.

  17. The relative impact of sizing errors on steam generator tube failure probability

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1998-01-01

    The Outside Diameter Stress Corrosion Cracking (ODSCC) at tube support plates is currently the major degradation mechanism affecting the steam generator tubes made of Inconel 600. This caused development and licensing of degradation specific maintenance approaches, which addressed two main failure modes of the degraded piping: tube rupture; and excessive leakage through degraded tubes. A methodology aiming at assessing the efficiency of a given set of possible maintenance approaches has already been proposed by the authors. It pointed out better performance of the degradation specific over generic approaches in (1) lower probability of single and multiple steam generator tube rupture (SGTR), (2) lower estimated accidental leak rates and (3) less tubes plugged. A sensitivity analysis was also performed pointing out the relative contributions of uncertain input parameters to the tube rupture probabilities. The dominant contribution was assigned to the uncertainties inherent to the regression models used to correlate the defect size and tube burst pressure. The uncertainties, which can be estimated from the in-service inspections, are further analysed in this paper. The defect growth was found to have significant and to some extent unrealistic impact on the probability of single tube rupture. Since the defect growth estimates were based on the past inspection records they strongly depend on the sizing errors. Therefore, an attempt was made to filter out the sizing errors and to arrive at more realistic estimates of the defect growth. The impact of different assumptions regarding sizing errors on the tube rupture probability was studied using a realistic numerical example. The data used is obtained from a series of inspection results from Krsko NPP with 2 Westinghouse D-4 steam generators. The results obtained are considered useful in safety assessment and maintenance of affected steam generators. (author)

  18. Creep and stress rupture behaviour of zircaloy-2 and Zr-2.5% Nb alloy tubes at 573 K

    International Nuclear Information System (INIS)

    Laha, K.; Bhanu Sankara Rao, K.; Chandravathi, K.S.; Mannan, S.L.

    1992-01-01

    Zirconium alloys are extensively used for coolant tubes of pressurised heavy water reactors. The choice of these materials is based on their good corrosion resistance in water, low capture cross section for thermal neutrons and good mechanical properties. In this paper the results of an investigation performed on the creep and rupture behaviour of indigenously produced zircaloy-2 and Zr-2.5% Nb alloy are presented. Samples for creep testing were cut longitudinally from finished pressure tubes. Creep rupture tests were carried out in air under constant load conditions at 300 C employing five stress levels in the range 300-360 MPa. Zr-2.5% Nb alloy displayed higher rupture lives at all stress levels compared to zircaloy-2. Steady state creep rate of Zr-2.5%Nb was lower than that zircaloy-2 at identical stress levels. In the stress range of the experiments, the dependence of the steady state creep rate (ε s ) on applied stress (σ) for both the alloys could be represented by a power law, ε s =A σ n The stress sensitivity (n) for Zr-2.5% Nb was lower than that of zircaloy-2. For both the alloys the time to creep rupture t r was found related to the steady state creep rate through the modified Monkman-Grant relation (ε s ) α . t r = constant. Similar value of α was obtained for both the materials. Zr-2.5%Nb exhibited higher ductility (% elongation to rupture) compared to zircaloy-2 at stress levels ≥ 320 MPa. At lower stresses significant difference in ductility was not noticed. Percentage reduction in area was lower in Zr-2.5%Nb at all stress levels indicating better resistance for necking. The time for onset of tertiary was longer for Zr-2.5% Nb alloy. The proportion of life spent by Zr-2.5% Nb in steady state creep regime was higher compared to that of zircaloy-2. Metallographic investigations on longitudinal sections in both the alloys showed large number of intragranular pores close to the fracture surface. A few number of cracks which are characteristic of

  19. Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

    2009-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

  20. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  1. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P [Nuclear Research Inst. Rez (Switzerland)

    1996-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  2. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1995-01-01

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

  3. Simulation of a main steam line break with steam generator tube rupture using trace

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

    2012-07-01

    A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

  4. Flaw analysis in steam generator tube

    International Nuclear Information System (INIS)

    Hutin, J.P.; Billon, F.

    1985-08-01

    Operating more than 30 PWR units, Electricite de France has to face several steam generator tube problems. One of the most serious difficulties is the stress corrosion cracking due to primary fluid, just above the tube sheet, in the roll transition region. With regard to availability it is, of course, a major concern; with regard to safety, the point is that tube rupture should be preceded by a significant primary-to-secondary leak during normal operation so that the reactor should be shut down before failure occurs. The demonstration of this assessment asks for experimental and analytical evidences. In 1981, Elecricite de France started a comprehensive program on that subject. A general description of this program and the main results are to be presented during the SMIRT-8 Conference. The purpose of the present paper is to develop in greater detail the analytical part of the work

  5. STAC -- a new Swedish code for statistical analysis of cracks in SG-tubes

    International Nuclear Information System (INIS)

    Poern, K.

    1997-01-01

    Steam generator (SG) tubes in pressurized water reactor plants are exposed to various types of degradation processes, among which stress corrosion cracking in particular has been observed. To be able to evaluate the safety importance of such cracking of SG-tubes one has to have a good and empirically founded knowledge about the scope and the size of the cracks as well as the rate of their continuous growth. The basis of experience is to a large extent constituted of the annually performed SG-inspections and crack sizing procedures. On the basis of this experience one can estimate the distribution of existing crack lengths, and modify this distribution with regard to maintenance (plugging) and the predicted rate of crack propagation. Finally, one can calculate the rupture probability of SG-tubes as a function of a given critical crack length. On account of the Swedish Nuclear Power Inspectorate an introductory study has been performed in order to get a survey of what has been done elsewhere in this field. The study resulted in a proposal of a computerizable model to be able to estimate the distribution of true cracks, to modify this distribution due to the crack growth and to compute the probability of tube rupture. The model has now been implemented in a compute code, called STAC (STatistical Analysis of Cracks). This paper is aimed to give a brief outline of the model to facilitate the understanding of the possibilities and limitations associated with the model

  6. Analysis of 30 breast implant rupture cases.

    Science.gov (United States)

    Tark, Kwan Chul; Jeong, Hii Sun; Roh, Tae Suk; Choi, Jong Woo

    2005-01-01

    Breast implants used for augmentation mammoplasty or breast reconstruction could rupture from various causes such as trauma or spontaneous failure. The objectives of this study were to investigate the relationships between the causes of implant rupture and the degree of capsular contracture, and then to evaluate the relative efficacies of specific signs on magnetic resonance imaging (MRI) known to be beneficial for diagnosing the rupture. A retrospective review identified patients with prosthetic implant rupture or impending rupture treated by the senior author. The 30 cases of implant rupture available for review were classified into two groups: intracapsular and extracapsular ruptures. The 30 cases of breast implant ruptures were analyzed with respect to the clinical symptoms and signs, the causes of rupture, the degree of capsular contracture, and therapeutic plans. Among the 30 cases, 14 patients who had undergone MRI during the diagnostic period were analyzed with respect to the relationships between MRI readings and operative findings. Spontaneous rupture of membranes was most common (80%), followed by failure because of trauma (7%) and valve or implant base (4%). The symptoms during implant rupture were contour deformity, palpated mass-like lesions, pain, and focal inflammation. According to the analysis of specific MRI signs, the sensitivity and specificity of the linguine sign were 87% and 100%, respectively, for intracapsular rupture. For extracapsular rupture, the sensitivity and specificity of the linguine sign were, respectively, 67% and 75%. The sensitivity and specificity of the rat-tail sign and tear drop sign were 14% and 50%, respectively. Breast implant rupture was correlated with the degree of capsular contracture in our study. Among the various specific MRI signs used in diagnosing the rupture, the linguine sign was reliable and had a high sensitivity and specificity, especially in cases of intracapsular rupture. On the other hand, the rat

  7. Evaluation of PWR response to main-steamline break with concurrent steam-generator tube rupture and small-break LOCA

    International Nuclear Information System (INIS)

    Laaksonen, J.T.; Sheron, B.W.

    1982-12-01

    In 1980, the NRC staff raised a potential safety issue involving a coincident steamline break, steam generator tube rupture, and small-break loss-of-coolant accident (LOCA). The bases for this concern were that the system response, primarily the maintenance of core cooling, was unanalyzed and the adequacy of the present guidance to operators to respond to combination LOCAs was unknown. This report discusses the staff evaluations performed to assess the system response and the adequacy of the present emergency operator guidelines. In all of the analyzed cases the primary coolant shrinkage, caused by overcooling, and the simultaneous loss of coolant can be compensated by the high pressure emergency core cooling system. The core remains covered with liquid, and the primary coolant remains subcooled, except in the vessel upper head. If the steamline break is outside the containment and cannot be isolated, the radiological consequences could be more severe than in any accident currently analyzed in a typical plant Final Safety Evaluation Report (FSAR). To decrease the risk of elevated offsite releases, an early diagnosis of the tube rupture has to be ensured. This can be done by upgrading operator instructions. The appropriate mitigating actions are in the existing instructions

  8. High-intensity focused ultrasound ablation around the tubing.

    Science.gov (United States)

    Siu, Jun Yang; Liu, Chenhui; Zhou, Yufeng

    2017-01-01

    High-intensity focused ultrasound (HIFU) has been emerging as an effective and noninvasive modality in cancer treatment with very promising clinical results. However, a small vessel in the focal region could be ruptured, which is an important concern for the safety of HIFU ablation. In this study, lesion formation in the polyacrylamide gel phantom embedded with different tubing (inner diameters of 0.76 mm and 3 mm) at varied flow speeds (17-339 cm/s) by HIFU ablation was photographically recorded. Produced lesions have decreased length (~30%) but slightly increased width (~6%) in comparison to that without the embedded tubing. Meanwhile, bubble activities during the exposures were measured by passive cavitation detection (PCD) at the varied pulse repetition frequency (PRF, 10-30 Hz) and duty cycle (DC, 10%-20%) of the HIFU bursts. High DC and low flow speed were found to produce stronger bubble cavitation whereas no significant influence of the PRF. In addition, high-speed photography illustrated that the rupture of tubing was produced consistently after the first HIFU burst within 20 ms and then multiple bubbles would penetrate into the intraluminal space of tubing through the rupture site by the acoustic radiation force. Alignment of HIFU focus to the anterior surface, middle, and posterior surface of tubing led to different characteristics of vessel rupture and bubble introduction. In summary, HIFU-induced vessel rupture is possible as shown in this phantom study; produced lesion sizes and shapes are dependent on the focus alignment to the tubing, flow speed, and tubing properties; and bubble cavitation and the formation liquid jet may be one of the major mechanisms of tubing rupture as shown in the high-speed photography.

  9. Prognostics for Steam Generator Tube Rupture using Markov Chain model

    International Nuclear Information System (INIS)

    Kim, Gibeom; Heo, Gyunyoung; Kim, Hyeonmin

    2016-01-01

    This paper will describe the prognostics method for evaluating and forecasting the ageing effect and demonstrate the procedure of prognostics for the Steam Generator Tube Rupture (SGTR) accident. Authors will propose the data-driven method so called MCMC (Markov Chain Monte Carlo) which is preferred to the physical-model method in terms of flexibility and availability. Degradation data is represented as growth of burst probability over time. Markov chain model is performed based on transition probability of state. And the state must be discrete variable. Therefore, burst probability that is continuous variable have to be changed into discrete variable to apply Markov chain model to the degradation data. The Markov chain model which is one of prognostics methods was described and the pilot demonstration for a SGTR accident was performed as a case study. The Markov chain model is strong since it is possible to be performed without physical models as long as enough data are available. However, in the case of the discrete Markov chain used in this study, there must be loss of information while the given data is discretized and assigned to the finite number of states. In this process, original information might not be reflected on prediction sufficiently. This should be noted as the limitation of discrete models. Now we will be studying on other prognostics methods such as GPM (General Path Model) which is also data-driven method as well as the particle filer which belongs to physical-model method and conducting comparison analysis

  10. Experimental evaluation of emergency operating procedures on multiple steam generator tube rupture in INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lin, Y.M.; Lee, C.H.; Chang, C.Y.; Hong, W.T.

    1997-01-01

    The multiple steam generator tube rupture (SGTR) scenario in Westinghouse type pressurized water reactor (PWR) has been investigated at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. This reduced-height and reduced-pressure test facility was designed to simulate the main features of Maanshan nuclear power plant. The SGTR test scenario assumes the double-ended break of one-, two- and six- tubes without other failures. The major operator actions follow the related symptom-oriented Emergency Operating Procedure (EOP) on the reference plant. This study focuses on the investigation of thermal-hydraulics phenomena and the adequacy of associated EOP to limit primary-to-secondary leakage. Through this study, it is found that the adequacy of current EOP in minimizing the radioactivity release demands early substantial operator involvement, especially in the multi-tubes break events. Also, the detailed mechanism of the main thermal-hydraulic phenomena during the SGTR transient are explored. (author)

  11. Depressurisation study of a tank-tubing assemble

    International Nuclear Information System (INIS)

    Freitas, R.L.

    1975-08-01

    The depressurisation of a nuclear reactor following the rupture of the primary coolant circuit is studied, using the simple analogy of the rupture of the tubing connected to a pressurised tank. The method of characteristics has been used in this theoretical analysis. The partial differential equations of conservation of mass, momentum and energy forming a hyperbolic system and defining real characteristic directions, allow the integration of these equations to be carried out along these directions. The method allows calculations to be made of the pressure, temperature, density and fluid velocity in the reactor circuit at any time after the beginning of depressurisation. A computer code MECA I has been written to calculate all the parameters after the rupture for any point in the coolant tubing. The computers used for these calculations were the IBM 360/40 and 370/145 and the Burroughs 6700. In this preliminary study, the simplest case of a system using a perfect gas coolant has been used [pt

  12. Break size effect on the transient thermal-hydraulic behavior during the steam generator tube rupture accident

    International Nuclear Information System (INIS)

    Kang, K.H.; Park, H.S.; Cho, S.; Choi, N.H.; Chu, I.C.; Yun, B.J.; Kim, K.D.; Kim, Y.S.; Baek, W.P.; Choi, K.Y.

    2011-01-01

    In order to simulate the SGTR accident of the APR1400, integral effect tests were performed by simulating a double-ended rupture of a single and five U-tubes. Following the reactor trip, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened to reduce the secondary system pressure. Break area affected the timings of the major events observed in the tests. Less heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow. (author)

  13. Comparison between smaller ruptured intracranial aneurysm and larger un-ruptured intracranial aneurysm: gene expression profile analysis.

    Science.gov (United States)

    Li, Hao; Li, Haowen; Yue, Haiyan; Wang, Wen; Yu, Lanbing; ShuoWang; Cao, Yong; Zhao, Jizong

    2017-07-01

    As it grows in size, an intracranial aneurysm (IA) is prone to rupture. In this study, we compared two extreme groups of IAs, ruptured IAs (RIAs) smaller than 10 mm and un-ruptured IAs (UIAs) larger than 10 mm, to investigate the genes involved in the facilitation and prevention of IA rupture. The aneurismal walls of 6 smaller saccular RIAs (size smaller than 10 mm), 6 larger saccular UIAs (size larger than 10 mm) and 12 paired control arteries were obtained during surgery. The transcription profiles of these samples were studied by microarray analysis. RT-qPCR was used to confirm the expression of the genes of interest. In addition, functional group analysis of the differentially expressed genes was performed. Between smaller RIAs and larger UIAs, 101 genes and 179 genes were significantly over-expressed, respectively. In addition, functional group analysis demonstrated that the up-regulated genes in smaller RIAs mainly participated in the cellular response to metal ions and inorganic substances, while most of the up-regulated genes in larger UIAs were involved in inflammation and extracellular matrix (ECM) organization. Moreover, compared with control arteries, inflammation was up-regulated and muscle-related biological processes were down-regulated in both smaller RIAs and larger UIAs. The genes involved in the cellular response to metal ions and inorganic substances may facilitate the rupture of IAs. In addition, the healing process, involving inflammation and ECM organization, may protect IAs from rupture.

  14. Initial Experience with Computed Tomography and Fluoroscopically Guided Placement of Push-Type Gastrostomy Tubes Using a Rupture-Free Balloon Catheter

    International Nuclear Information System (INIS)

    Fujita, Takeshi; Tanabe, Masahiro; Yamatogi, Shigenari; Shimizu, Kensaku; Matsunaga, Naofumi

    2011-01-01

    The purpose of this study was to evaluate the safety and feasibility of percutaneous radiologic gastrostomy placement of push-type gastrostomy tubes using a rupture-free balloon (RFB) catheter under computed tomography (CT) and fluoroscopic guidance. A total of 35 patients (23 men and 12 women; age range 57–93 years [mean 71.7]) underwent percutaneous CT and fluoroscopically guided gastrostomy placement of a push-type gastrostomy tube using an RFB catheter between April 2005 and July 2008. Technical success, procedure duration, and complications were analyzed. Percutaneous radiologic gastrostomy placement was considered technically successful in all patients. The median procedure time was 39 ± 13 (SD) min (range 24–78). The average follow-up time interval was 103 days (range 7–812). No major complications related to the procedure were encountered. No tubes failed because of blockage, and neither tube dislodgement nor intraperitoneal leakage occurred during the follow-up period. The investigators conclude that percutaneous CT and fluoroscopically guided gastrostomy placement with push-type tubes using an RFB catheter is a safe and effective means of gastric feeding when performed by radiologists.

  15. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

    2005-04-15

    Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

  16. Application of discrete scale invariance method on pipe rupture

    International Nuclear Information System (INIS)

    Rajkovic, M.; Mihailovic, Z.; Riznic, J.

    2007-01-01

    'Full text:' A process of material failure of a mechanical system in the form of cracks and microcracks, a catastrophic phenomenon of considerable technological and scientific importance, may be forecasted according to the recent advances in the theory of critical phenomena in statistical physics. Critical rupture scenario states that, in many concrete and composite heterogeneous materials under compression and materials with large distributed residual stresses, rupture is a genuine critical point, i.e., the culmination of a self-organization of damage and cracking characterized by power law signatures. The concept of discrete scale invariance leads to a complex critical exponent (or dimension) and may occur spontaneously in systems and materials developing rupture. It establishes, theoretically, the power law dependence of a measurable observable, such as the rate of acoustic emissions radiated during loading or rate of heat released during the process, upon the time to failure. However, the problem is the power law can be distinguished from other parametric functional forms such as an exponential only close to the critical time. In this paper we modify the functional renormalization method to include the noise elimination procedure and dimension reduction. The aim is to obtain the prediction of the critical rupture time only from the knowledge of the power law parameters at early times prior to rupture, and based on the assumption that the dynamics close to rupture is governed by the power law dependence of the temperature measured along the perimeter of the tube upon the time-to-failure. Such an analysis would not only enhance the precision of prediction related to the rupture mechanism but also significantly help in determining and predicting the leak rates. The prediction will be compared to experimental data on Zr-2.5%Nb made tubes. Note: The views expressed in the paper are those of the authors and do not necessary represents those of the commission. (author)

  17. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Hareux, F.; Roche, R.; Vrillon, B.

    1964-01-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [fr

  18. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  19. Single Tube Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

    2010-12-01

    In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of a single U-tube in the APR1400, the SGTR-CL-01 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Following the reactor trip induced by high steam generator level (HSGL) signal, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened. The MSSVs repeated on and off status depending on the secondary system pressure during the whole test period. Due to the break flow, the collapsed water level of the affected steam generator showed milder decrease than that of the intact steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for a SGTR simulation, especially for DVI-adapted plants

  20. ANALISIS KEJADIAN STEAM GENERATOR TUBE RUPTURE (SGTR BERDASARKAN SKENARIO MIHAMA UNIT 2

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

    Full Text Available Pada tanggal 9 Februari 1991, terjadi kecelakaan putusnya pipa pemanas pembangkit uap (Steam Generator Tube Rupture/SGTR pada PLTN Mihama Unit 2. Dari kejadian tersebut, diperoleh catatan sekuensi kecelakaan berupa aktuasi sistem proteksi dan fitur keselamatan terekayasa dalam memitigasi kebocoran dari sistem primer ke sistem sekunder. Urutan sekuensi tersebut kemudian diterapkan pada PWR standar Jepang untuk disimulasikan menggunakan program perhitungan RELAP5/SCDAP/Mod3.2. Tujuannya untuk mengevaluasi konsekuensi yang terjadi bila kecelakaan tersebut terjadi pada PWR standar Jepang. Parameter yang dibandingkan adalah laju alir kebocoran, perubahan tekanan primer dan sekunder dan perubahan level di dalam pressurizer. Hasil simulasi menunjukkan perbedaan lama waktu kejadian SGTR hingga berhentinya kebocoran yang berlangsung lebih pendek pada PWR standar Jepang. Selain itu jumlah pendingin primer yang bocor dan jumlah uap yang terlepas dari MSRV tercatat lebih besar daripada PWR Mihama unit 2. Karakter aliran kebocoran, fluktuasi tekanan primer, dan level pressurizer sedikit berbeda pada tahap-tahap awal kejadian, namun relatif sama pada tahap akhir ketika aliran kebocoran dapat dihentikan. Hasil simulasi juga menunjukkan perlunya tindakan operator secara manual yang ditunjukkan dari isolasi sistem air umpan bantu (AFW pada pembangkit uap yang bocor, aktuasi katup pelepas uap (MSRV pada pembangkit uap yang utuh dan aktuasi auxiliary spray dan power operated relief valve (PORV pada pressurizer untuk mengantisipasi kejadian sebagai bagian dari prosedur operasi darurat. Kata kunci: SGTR, PWR Mihama Unit 2, PWR standar Jepang   On February 9,1991, a Steam Generator Tube Rupture (SGTR took place at the Mihama Unit No. 2. From that event, the accident sequence representing the actuation of protection system and engineered safety feature to mitigate the leak from primary system to secondary system is recorded. That sequence is then applied on the

  1. Effect of steam corrosion on HTGR core support post strength loss. Part II. Consequences of steam generator tube rupture event

    International Nuclear Information System (INIS)

    Wichner, R.P.

    1977-01-01

    To perform the assessment, a series of eight tube-rupture events of varying severity and probability were postulated. Case 1 pertains to the situation where the moisture detection, loop isolation, and dump procedures function as planned; the remaining seven cases suppose various defects in the moisture detection system, the core auxiliary coolant system, and the integrity of the prestressed concrete reactor vessel. Core post burnoffs beneath three typical fuel zones were estimated for each postulated event from the determined impurity compositions and core post temperature history. Two separate corrosion rate expressions were assumed, as deemed most appropriate of those published for the high-oxidant level typical in tube rupture events. It was found that the nominal core post beneath the highest power factor fuel zone would lose from 0.02 to 2.5 percent of their strength, depending on an assumed corrosion rate equation and the severity of the event. The effect of hot streaking during cooldown was determined by using preliminary estimates of its magnitude. It was found that localized strength loss beneath the highest power factor zone ranges from 0.23 to 12 percent, assuming reasonably probable hot-streaking circumstances. The combined worst case, hot streaking typical for a load-following transient and most severe accident sequence, yields an estimated strength loss of from 25 to 33 percent for localized regions beneath the highest power factor zones

  2. 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.)

  3. High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Suzuki, Motoye

    1982-03-01

    In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

  4. 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)

  5. The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

    2012-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

  6. Shock tubes: compressions in the low pressure chamber

    International Nuclear Information System (INIS)

    Schins, H.; Giuliani, S.

    1986-01-01

    The gas shock tube used in these experiments consists of a low pressure chamber and a high pressure chamber, divided by a metal-diaphragm-to-rupture. In contrast to the shock mode of operation, where incident and reflected shocks in the low pressure chamber are studied which occur within 3.5 ms, in this work the compression mode of operation was studied, whose maxima occur (in the low pressure chamber) about 9 ms after rupture. Theoretical analysis was done with the finite element computer code EURDYN-1M, where the computation was carried out to 30 ms

  7. Validation of statistical models for creep rupture by parametric analysis

    Energy Technology Data Exchange (ETDEWEB)

    Bolton, J., E-mail: john.bolton@uwclub.net [65, Fisher Ave., Rugby, Warks CV22 5HW (United Kingdom)

    2012-01-15

    Statistical analysis is an efficient method for the optimisation of any candidate mathematical model of creep rupture data, and for the comparative ranking of competing models. However, when a series of candidate models has been examined and the best of the series has been identified, there is no statistical criterion to determine whether a yet more accurate model might be devised. Hence there remains some uncertainty that the best of any series examined is sufficiently accurate to be considered reliable as a basis for extrapolation. This paper proposes that models should be validated primarily by parametric graphical comparison to rupture data and rupture gradient data. It proposes that no mathematical model should be considered reliable for extrapolation unless the visible divergence between model and data is so small as to leave no apparent scope for further reduction. This study is based on the data for a 12% Cr alloy steel used in BS PD6605:1998 to exemplify its recommended statistical analysis procedure. The models considered in this paper include a) a relatively simple model, b) the PD6605 recommended model and c) a more accurate model of somewhat greater complexity. - Highlights: Black-Right-Pointing-Pointer The paper discusses the validation of creep rupture models derived from statistical analysis. Black-Right-Pointing-Pointer It demonstrates that models can be satisfactorily validated by a visual-graphic comparison of models to data. Black-Right-Pointing-Pointer The method proposed utilises test data both as conventional rupture stress and as rupture stress gradient. Black-Right-Pointing-Pointer The approach is shown to be more reliable than a well-established and widely used method (BS PD6605).

  8. Savannah River reactor process water heat exchanger tube structural integrity margin Task Number 92-005-1

    International Nuclear Information System (INIS)

    Mertz, G.E.; Barnes, D.M.; Sindelar, R.L.

    1992-02-01

    Twelve process water heat exchangers are designed to remove heat generated in the reactor tank. Each heat exchanger has approximately 9000, 1/2 inch diameter x 0.049 inches thick tubes. Minimum structural tubing requirements and the leak rate through postulated tubing defects are developed in this report A comparison of the structural requirements and the defect size calculated to produce leak rates of 0.5 lbs./day demonstrate adequate structural margins against gross tube rupture. Commercial nuclear experience with pressurized water reactor (PWR) steam generator plugging criteria are used for guidance in performing this analysis. It is important to note that the SRS reactors are low energy systems with normal operating pressures of 203 psig at 130 degree F while the PWR is a high energy system with operating pressures near 2200 psig at 600 degree F. Clearly the PVM steam generator has loadings which are more severe than the SRS heat exchangers. Consistent with the Regulatory Guide 1.121 criteria both wastage (wall thinning) and cracking are addressed. Structural limits on wall thinning and crack size are developed to preclude gross rupture. ASME Section XI criteria, with the factors of safety recommended by Regulatory Guide 1.121 are used to develop the allowable crack size criteria. Normal operating conditions (pressure, dead weight, and hydraulic drag) are considered with seismic and water hammer accident conditions. Both the wall thinning and crack size criteria are developed for the end-of-evaluation period. Allowances for corrosion, wear, or crack growth have not been included in this analysis Structurally, the tubing is over designed and can tolerate large defects with adequate margins against gross rupture. The structural margins of heat exchanger tubing are evident by contrasting the tubing's structural capacity, per the ASME Code, with its operating conditions/configuration

  9. Analysis of the ballooning deformation of an internally pressurized thin-wall tube during fast thermal transients

    International Nuclear Information System (INIS)

    Lin, E.I.H.

    1977-01-01

    A large-strain time-dependent thermoplastic analysis has been developed for the ballooning deformation of a thin-wall tube subjected to internal pressure, axial loading, and fast thermal transients. This deformation initiates with the onset of plastic instability in the material, the onset being determined by a plastic-instability criterion for strain-rate sensitive materials. The interaction among the local ballooning geometry, the state of stress, and the plastic flow process was considered, and integration of the flow equations yields the local curvature and the states of stress and strain in the vicinity of the maximum ballooning site. The effects of axial constraint and heating rate were also discussed. The analysis was applied to a LWR Zircaloy cladding subjected to a constant heating rate and a range of internal pressures. The results agree very well with experimental strain-time data obtained from tube-burst tests. In most cases, the time of rupture was accurately predicted despite the lack of complete material-property data

  10. Evaluation of maintenance strategies for steam generator tubes in pressurized waster reactors. 2. Cost and profitability analyses

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Yoshimura, S.; Yagawa, G.

    2000-01-01

    As an application of probabilistic fracture mechanics (PFM), risk-benefit analysis was carried out to evaluate maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The analysis was conducted for SG tubes made of Inconel 600, and Inconel 690 as well assuming its crack initiation and crack propagation law based on Inconel 600 data. The following results were drawn from the analysis. Improvement of inspection accuracy reduces the maintenance costs significantly and is preferable from the viewpoint of profitability due to reduction of SG tube leakage and rupture. There is a certain region of SCC properties of SG tubes where sampling inspection is effective. (author)

  11. Failure analysis on a ruptured petrochemical pipe

    Energy Technology Data Exchange (ETDEWEB)

    Harun, Mohd [Industrial Technology Division, Malaysian Nuclear Agency, Ministry of Science, Technology and Innovation Malaysia, Bangi, Kajang, Selangor (Malaysia); Shamsudin, Shaiful Rizam; Kamardin, A. [Univ. Malaysia Perlis, Jejawi, Arau (Malaysia). School of Materials Engineering

    2010-08-15

    The failure took place on a welded elbow pipe which exhibited a catastrophic transverse rupture. The failure was located on the welding HAZ region, parallel to the welding path. Branching cracks were detected at the edge of the rupture area. Deposits of corrosion products were also spotted. The optical microscope analysis showed the presence of transgranular failures which were related to the stress corrosion cracking (SCC) and were predominantly caused by the welding residual stress. The significant difference in hardness between the welded area and the pipe confirmed the findings. Moreover, the failure was also caused by the low Mo content in the stainless steel pipe which was detected by means of spark emission spectrometer. (orig.)

  12. Prospects for stronger calandria tubes

    International Nuclear Information System (INIS)

    Ells, C.E.; Coleman, C.E.; Hosbons, R.R.; Ibrahim, E.F.; Doubt, G.L.

    1990-12-01

    The CANDU calandria tubes, made of seam welded and annealed Zircaloy-2, have given exemplary service in-reactor. Although not designed as a system pressure containment, calandria tubes may remain intact even in the face of pressure tube rupture. One such incident at Pickering Unit 2 demonstrated the economic advantage of such an outcome, and a case can be made for increasing the probability that other calandria tubes would perform in a similar fashion. Various methods of obtaining stronger calandria tubes are available, and reviewed here. When the tubes are internally pressurized, the weld is the weak section of the tube. Increasing the oxygen concentration in the starting sheet, and thickening the weld, are promising routes to a stronger tube

  13. Failure analysis of boiler tube

    International Nuclear Information System (INIS)

    Mehmood, K.; Siddiqui, A.R.

    2007-01-01

    Boiler tubes are energy conversion components where heat energy is used to convert water into high pressure superheated steam, which is then delivered to a turbine for electric power generation in thermal power plants or to run plant and machineries in a process or manufacturing industry. It was reported that one of the tubes of a fire-tube boiler used in a local industry had leakage after the formation of pits at the external surface of the tube. The inner side of the fire tube was working with hot flue gasses with a pressure of 10 Kg/cm/sup 2/ and temperature 225 degree C. The outside of the tube was surrounded by feed water. The purpose of this study was to determine the cause of pits developed at the external surface of the failed boiler tube sample. In the present work boiler tube samples of steel grade ASTM AI61/ASTM A192 were analyzed using metallographic analysis, chemical analysis, and mechanical testing. It was concluded that the appearance of defects on the boiler tube sample indicates cavitation type corrosion failure. Cavitation damage superficially resembled pitting, but surface appeared considerably rougher and had many closely spaced pits. (author)

  14. 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)

  15. Consideration on evaluation of internal pressure creep rupture for tube with circumferential joint

    International Nuclear Information System (INIS)

    Nagato, Kotaro; Satoh, Keisuke

    1983-01-01

    The behavior of internal pressure creep rupture of the thin-walled cylinders with circumferential joints is affected by the combination of creep characteristics of parent materials and weld metals. In particular, the compatibility of the creep strain rate of parent materials and weld metals becomes an important controlling factor. The behavior of internal pressure creep of the welded parts in circumferential joint cylinders can be evaluated simply with the uniaxial creep data of parent materials and weld metals, considering it by approximately substituting with the creep behavior of a uniaxial longitudinal joint. The method of evaluation is, first, to analyze the breaking behavior of uniaxial longitudinal joints using the uniaxial creep characteristic values of parent materials and weld metals, and next, by combining the equation for the relation between the rupture times of uniaxial creep and internal pressure creep with the analyzed breaking behavior of uniaxial joints, the internal pressure creep rupture behavior of the cylinders with circumferential joints can be evaluated. The internal pressure creep behavior of the thin-walled cylinders with circumferential joints, their rupture life and the uniaxial creep rupture life of longitudinal joints, and the examination of Hastelloy X cylinders are reported. (Kako, I.)

  16. Large-leak sodium-water reaction analysis for steam generators

    International Nuclear Information System (INIS)

    Sakano, K.; Shindo, Y.; Hori, M.

    1975-01-01

    The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

  17. Large-leak sodium-water reaction analysis for steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sakano, K; Shindo, Y; Hori, M

    1975-07-01

    The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

  18. Intentional back flow effects on ruptured steam generator cooldown during a SGTR event for KSNP

    International Nuclear Information System (INIS)

    Seok, Jeong Park; Cheol, Woo Kim; Chul, Jin Choi; Jong, Tae Seo

    2001-01-01

    For an optimum recovery from a Steam Generator Tube Rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube(s) as early as possible in order to minimize the radioactive material release. However, the Reactor Coolant System (RCS) cooldown and depressurization to the Residual Heat Removal (RHR) System operation conditions using the intact SG only can not be readily achievable unless the affected SG is properly cooled since the isolated SG remains at high temperature even though the RCS has been cooled down. Therefore, a study on the intentional back flow from the ruptured SG secondary side to the RCS was performed to evaluate its effectiveness on the ruptured SG cooldown during a SGTR event for the pressurized light water reactor, especially for the Korean Standard Nuclear Power Plant (KSNP). In order to evaluate the intentional back flow effect, a series of analyses was conducted by using RELAP5/MOD3 computer code. In these analyses, the primary and secondary systems of KSNP are modeled including the major Nuclear Steam Supply System (NSSS) components such as the reactor vessel, steam generators, hot and cold legs, pressurizer, and reactor coolant pumps. Also, the key safety systems and control systems are modeled. Using this model, two possible methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated: the first method is a tube uncover method, and the second method is a SG drain (back flow) and fill method. (author)

  19. Simulation of steam generator plugging tubes in a PWR to analyze the operating impact

    Energy Technology Data Exchange (ETDEWEB)

    Pla, Patricia, E-mail: patricia.pla-freixa@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands); Reventos, Francesc, E-mail: francesc.reventos@upc.edu [Technical University of Catalonia (UPC), Barcelona (Spain); Martin Ramos, Manuel, E-mail: manuel.martin-ramos@ec.europa.eu [Nuclear Safety and Security Coordination Unit, Policy Support Coordination, Joint Research Centre of the European Commission, Brussels (Belgium); Sol, Ismael, E-mail: isol@anacnv.com [Asociación Nuclear Ascó-Vandellós-II (ANAV), Tarragona (Spain); Strucic, Miodrag, E-mail: miodrag.strucic@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands)

    2016-08-15

    Highlights: • Plugging a fraction of the SG tubes does not affect power output of the plant. • There is a limit to SG plugging in the range of 10–15%. • The rupture of a SG tube in a 12% plugged SG has shown no significant differences in operator actions. • A SBLOCA in a 12% plugged SG has shown no significant differences in operator actions. - Abstract: A number of nuclear power plants (NPPs) with pressurized water reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems. Several methods were attempted to correct the defects of the tubes, but eventually the only permanent solution was to plug them. The consequences of plugging the tubes are the decrease of heat transfer surface, the reduction of the flow area and subsequent reduction of the primary system mass flow and for a fraction of plugged tubes higher than a given value, the reduction of reactor output and economic losses. The objective of this paper is to analyze whether steam generator tube plugging has an impact in the effectiveness of accident management actions. An analysis with Relap5 Mod 3.3 patch03 for the Spanish reactor Ascó-2, a 3-loop 2940.6 MWth Westinghouse PWR, in which plugging of steam generator tubes are simulated, is presented in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modeling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for optimal reactor operation. To complete the study two events, in which the steam generators are used to cooldown the plant, were simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the emergency operating

  20. Review of damages of nuclear power plants steam generator's tubes and way of detecting by using eddy current method

    International Nuclear Information System (INIS)

    Stanic, D.

    1996-01-01

    Steam generator tubing integrity is very important factor for reliable and safe operation of NPP. Several different types of tube degradation mechanisms were experienced in SG operation. To avoid possible tube rupture and primary-to-secondary leak, the EC examination of tubing should be performed. Different eddy current techniques may be used for detecting defects and theirs characterization. A comparison of data analysis results with pulled tube destructive metallography results can provide valuable insights in determining the capability of existing technology and provide guidance for procedure or technology improvements. (author)

  1. Laser Welding Of Finned Tubes Made Of Austenitic Steels

    Directory of Open Access Journals (Sweden)

    Stolecki M.

    2015-09-01

    Full Text Available This paper describes the technology of welding of finned tubes made of the X5CrNi1810 (1.4301 austenitic steel, developed at Energoinstal SA, allowing one to get high quality joints that meet the requirements of the classification societies (PN-EN 15614, and at the same time to significantly reduce the manufacturing costs. The authors described an automatic technological line equipped with a Trumph disc laser and a tube production technological process. To assess the quality of the joints, one performed metallographic examinations, hardness measurements and a technological attempt to rupture the fin. Analysis of the results proved that the laser-welded finned tubes were performed correctly and that the welded joints had shown no imperfections.

  2. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor; Etude des consequences de la rupture d'un tube de force dans la cuve d'un reacteur modere a l'eau lourde et refroidi au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Hareux, F; Roche, R; Vrillon, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [French] L'eclatement d'un tube de force dans la cuve d'un reacteur de puissance modere a l'eau lourde et refroidi par un gaz sous pression est un accident qui a ete etudie experimentalement a propos d'EL-4. Un premier essai a l'echelle 1 ayant montre que l'eclatement d'un tube ne provoque pas celui des tubes voisins, des essais relatifs a la tenue de la cuve ont ete effectues sur maquettes a echelle tres reduite (l/lO). Il a ete trouve que la cuve peut supporter plusieurs eclatements de tubes sans deformations notables. La pression transitoire dans la cuve a une allure oscillatoire amortie dont le maximum (pression de pic) a fait l'objet d'une etude experimentale detaillee. Il apparait que les parametres essentiels influant sur cette pression sont: la pression du gaz contenu dans le tube de force, le volume du gaz qui participe a l'eclatement, la flexibilite de la cuve, la masse d'air empoisonnee dans la cuve, la nature du gaz explosant. Une methode generale d'estimation des pics de pression dans

  3. Multiregion analysis of creep rupture data of 316 stainless steel

    International Nuclear Information System (INIS)

    Maruyama, Kouichi; Armaki, Hassan Ghassemi; Yoshimi, Kyosuke

    2007-01-01

    A creep rupture data set of 316 stainless steel containing 319 data points at nine heats was subjected to a conventional single-region analysis and a multiregion analysis. In the former, the conventional Larson-Miller analysis was applied to the whole data set. In the latter, a data set of a single heat is divided into several data sets, so that the Orr-Sherby-Dorn (OSD) constant Q takes a unique value in each data set, and the conventional OSD analysis was applied to each divided data set. A region with a low value of Q appears in long-term creep of eight heats. Predicted values of the 10 5 h creep rupture stress of three heats were lower than the 99% confidence limit evaluated by the single-region analysis, suggesting that the single-region analysis is error prone. The multiregion analysis is necessary for the correct evaluation of the long-term creep properties of 316 stainless steel

  4. Tube manufacturing and characterization of oxide dispersion strengthened ferritic steels

    International Nuclear Information System (INIS)

    Ukai, Shigeharu; Mizuta, Shunji; Yoshitake, Tunemitsu; Okuda, Takanari; Fujiwara, Masayuki; Hagi, Shigeki; Kobayashi, Toshimi

    2000-01-01

    Oxide dispersion strengthened (ODS) ferritic steels have an advantage in radiation resistance and superior creep rupture strength at elevated temperature due to finely distributed Y 2 O 3 particles in the ferritic matrix. Using a basic composition of low activation ferritic steel (Fe-12Cr-2W-0.05C), cladding tube manufacturing by means of pilger mill rolling and subsequent recrystallization heat-treatment was conducted while varying titanium and yttria contents. The recrystallization heat-treatment, to soften the tubes hardened due to cold-rolling and to subsequently improve the degraded mechanical properties, was demonstrated to be effective in the course of tube manufacturing. For a titanium content of 0.3 wt% and yttria of 0.25 wt%, improvement of the creep rupture strength can be attained for the manufactured cladding tubes. The ductility is also adequately maintained

  5. Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

    Energy Technology Data Exchange (ETDEWEB)

    Soliman, S A; Lee, T; Ibrahim, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs.

  6. Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

    International Nuclear Information System (INIS)

    Soliman, S.A.; Lee, T.; Ibrahim, A.M.; Hodgson, S.

    1995-01-01

    This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs

  7. Total rupture of hydatid cyst of liver in to common bile duct: a case report.

    Science.gov (United States)

    Robleh, Hassan; Yassine, Fahmi; Driss, Khaiz; Khalid, Elhattabi; Fatima-Zahra, Bensardi; Saad, Berrada; Rachid, Lefriyekh; Abdalaziz, Fadil; Najib, Zerouali Ouariti

    2014-01-01

    Rupture of hydatid liver cyst into biliary tree is frequent complications that involve the common hepatic duct, lobar biliary branches, the small intrahepatic bile ducts,but rarely rupture into common bile duct. The rupture of hydatid cyst is serious life threating event. The authors are reporting a case of total rupture of hydatid cyst of liver into common bile duct. A 50-year-old male patient who presented with acute cholangitis was diagnosed as a case of totally rupture of hydatid cyst on Abdominal CT Scan. Rupture of hydatid cyst of liver into common bile duct and the gallbladder was confirmed on surgery. Treated by cholecystectomy and T-tube drainage of Common bile duct.

  8. Rupture of an expander prosthesis mimics axillary cancer recurrence.

    LENUS (Irish Health Repository)

    Ismael, T

    2005-10-01

    Regional silicone gel migration from a ruptured breast implant has been reported at different locations including the upper extremity, chest wall muscles, axilla and back. We report a patient who presented with an axillary mass that mimicked a regional recurrence 5 years after breast cancer reconstruction with a latissimus dorsi musculocutaneous flap and silicon gel expander-prosthesis. Surgical exploration revealed that the mass contained silicone gel around the port of the breast expander that had ruptured. The mass was confluent with an intracapsular silicone leak through a tract along the tube of the expander port.

  9. Four cases of spontaneous rupture of the urinary bladder

    International Nuclear Information System (INIS)

    Amano, Toshiyasu; Miwa, Sotaro; Takashima, Hiroshi; Takemae, Katsuro

    2002-01-01

    Between November 1997 and March 2001, 4 female patients from 44 to 65 years of age with a spontaneous rupture of the urinary bladder were analyzed. They complained of abdominal pain and had undergone an intra-pelvic gynecological operation (3 for uterine cancer, 1 for an ovarian cyst) several years before. The three with uterine cancer had also received radiation therapy. For their present condition, spontaneous urinary bladder rupture, their treatment was indwelling a urethral catheter. Two of them have had no recurrence of urinary bladder rupture after one month since having the urethral catheter indwelt. One, however, had to have the catheter re-indwelt due to unsuccessful suturing of the urinary bladder wall. The fourth patient had bilateral nephrostomy tubes due to severe radiation cystitis. Thus, one can infer that intra-pelvic gynecological operations and radiation therapy are major factors causing spontaneous urinary bladder rupture. While indwelling a urethral catheter may be effective for some patients with a spontaneous rupture of the urinary bladder, it may be very difficult to treat more complicated cases. (author)

  10. Creep-Rupture Behavior of Ni-Based Alloy Tube Bends for A-USC Boilers

    Science.gov (United States)

    Shingledecker, John

    Advanced ultrasupercritical (A-USC) boiler designs will require the use of nickel-based alloys for superheaters and reheaters and thus tube bending will be required. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section II PG-19 limits the amount of cold-strain for boiler tube bends for austenitic materials. In this summary and analysis of research conducted to date, a number of candidate nickel-based A-USC alloys were evaluated. These alloys include alloy 230, alloy 617, and Inconel 740/740H. Uniaxial creep and novel structural tests and corresponding post-test analysis, which included physical measurements, simplified analytical analysis, and detailed microscopy, showed that different damage mechanisms may operate based on test conditions, alloy, and cold-strain levels. Overall, creep strength and ductility were reduced in all the alloys, but the degree of degradation varied substantially. The results support the current cold-strain limits now incorporated in ASME for these alloys for long-term A-USC boiler service.

  11. A two-fluid two-phase model for thermal-hydraulic analysis of a U-tube steam generator

    International Nuclear Information System (INIS)

    Hung, Huanjen; Chieng, Chingchang; Pei, Baushei; Wang, Songfeng

    1993-01-01

    The Advanced Thermal-Hydraulic Analysis Code for Nuclear Steam Generators (ATHANS) was developed on the basis of the THERMIT-UTSG computer code for U-tube steam generators. The main features of the ATHANS model are as follows: (a) the equations are solved in cylindrical coordinates, (b) the number and the arrangement of the control volumes inside the steam generator can be chosen by the user, (c) the virtual mass effect is incorporated, and (d) the conjugate gradient squared method is employed to accelerate and improve the numerical convergence. The performance of the model is successfully validated by comparison with the test data from a Westinghouse model F steam generator at the Maanshan nuclear power plant. Better agreement with the test data can be obtained by a finer grid system using a cylindrical coordinate system and the virtual mass effect. With these advanced features, ATHANS provides the basic framework for further studies on the problems of steam generators, such as analyses of secondary-side corrosion and tube ruptures

  12. 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)

  13. STEAM GENERATOR TUBE INTEGRITY ANALYSIS OF A TOTAL LOSS OF ALL HEAT SINKS ACCIDENT FOR WOLSONG NPP UNIT 1

    Directory of Open Access Journals (Sweden)

    HEOK-SOON LIM

    2014-02-01

    Full Text Available A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS and the steam generator (SG secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  14. Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo [Korea Htydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Seoungrae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

    2014-02-15

    A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  15. Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

    International Nuclear Information System (INIS)

    Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo; Kim, Seoungrae

    2014-01-01

    A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident

  16. Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin [Power Engineering Research Institute, KEPCO Engineering and Construction, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2014-08-15

    Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor.

  17. Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

    International Nuclear Information System (INIS)

    Oh, Young-Jin; Chang, Yoon-Suk

    2014-01-01

    Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor

  18. ANL/CANTIA code for steam generator tube integrity assessment

    International Nuclear Information System (INIS)

    Revankar, S.T.; Wolf, B.; Majumdar, S.; Riznic, J.R.

    2009-01-01

    Steam generator (SG) tubes have an important safety role in CANDU type reactors and Pressurized Water Reactors (PWR) because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear plant. The SG tubes are susceptible to corrosion and damage. A failure of a single steam generator tube, or even a few tubes, would not be a serious safety-related event in a CANDU reactor. The leakage from a ruptured tube is within makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. A sufficient safety margin against tube rupture used to be the basis for a variety of maintenance strategies developed to maintain a suitable level of plant safety and reliability. Several through-wall flaws may remain in operation and potentially contribute to the total primary-to-secondary leak rate. Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits has been used for steam generator tube fitness-for-service assessment. The advantage of this type of analysis is that it avoids the excessive conservatism typically present in deterministic methodologies. However, it requires considerable effort and expense to develop all of the failure, leakage, probability of detection, and flaw growth distributions and models necessary to obtain meaningful results from a probabilistic model. The Canadian Nuclear Safety Commission (CNSC) recently developed the CANTIA methodology for probabilistic assessment of inspection strategies for steam generator tubes as a direct effect on the probability of tube failure and primary-to-secondary leak rate Recently Argonne National Laboratory has developed tube integrity and leak rate models under Integrated Steam Generator Tube Integrity Program (ISGTIP-2). These models have been incorporated in the ANL/CANTIA code. This paper presents the ANL

  19. Multi-target Wastage Phenomena on Steam Generator Tubes During an SWR Event

    International Nuclear Information System (INIS)

    Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Eoh, Jae Hyuk; Choi, Jong Hyeun; Lee, Yong Bum

    2011-01-01

    The Korean sodium cooled fast reactor, KALIMER- 600 (Korea Advanced LIquid MEtal Reactor) of which the electric output is 600MWe, was developed. The steam generator (SG) of this system is a shell-and-tube type counter-current flow heat exchanger, which is vertically oriented with fixed tube-sheets. A direct heat exchange occurs between the shell-side sodium and the tube-side water at the SG unit. Feed-water enters the inlet nozzle at the lower part of the unit and it flows upward along the helically coiled heat transfer tubes. The inflow sodium is cooled down at the bundle region and then flows out through the sodium outlet nozzle at the bottom of the unit. The typical configuration of the KALIMER-600 SG is shown in Figure 1. In a steam generator, sodium and water are separated by the heat transfer tube wall and it makes a strong pressure boundary between the shell-side sodium and the tube-side water/steam. For this reason, if there is a small hole or crack, even with a pin hole, on heat transfer tubes, a large amount of water/steam would leak into the liquid sodium due to the high pressure difference more than 150 bars, and an exothermic sodium-water chemical reaction takes place as a result. This type of sodium-water reaction (SWR) has been considered as one of the most important safety issues to be resolved. From previous studies, it was obviously figured out that the number of ruptured tubes during an SWR event is one of the most significant factors to determine the temperature and pressure transient. Any subsequent tube rupture behavior in the vicinity of the initially postulated single ruptured tube should be evaluated by considering the single- and multi-target wastage phenomena. Wastage is defined as damage to the structural material (e.g. heat transfer tubes) due to an impingement of the highly corrosive reaction product. Since the impingement may cause wastage of the neighboring heat transfer tubes, a subsequent tube failure can occur in a very short time

  20. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

    2001-01-01

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

  1. MODTURCCLAS analysis of moderator poison/coolant mixing in the calandria due to a pressure tube/calandria tube guillotine rupture during an overpoisoned guaranteed shutdown state

    International Nuclear Information System (INIS)

    Mackinnon, J.C.; Szymanski, J.K.; Balog, G.

    1996-01-01

    This paper reports the results of a study to investigate moderator poison/coolant mixing due to a guillotine rupture of a fuel channel when the reactor is in an overpoisoned guaranteed shutdown state. The analysis, performed using MODTURC C LAS, allowed for study of the mixing characteristics and the spatial and temporal evolution of the concentration fields. Results for simulated breaks at three channel locations show that the poison in the vessel is quite well mixed throughout the transient, resulting in no extensive regions of low poison concentration. MODTURC C LAS calculations show that at all three break locations investigated, the displacement of poison from the vessel through the relief ducts is less than that calculated by both the simple uniform mixing model and piston mixing model. This result is expected to hold for all break locations in the core. (author)

  2. Abridged republication of FIGO's staging classification for cancer of the ovary, fallopian tube, and peritoneum.

    Science.gov (United States)

    Prat, Jaime

    2015-10-01

    Ovarian, fallopian tube, and peritoneal cancers have a similar clinical presentation and are treated similarly, and current evidence supports staging all 3 cancers in a single system. The primary site (i.e. ovary, fallopian tube, or peritoneum) should be designated where possible. The histologic type should be recorded. Intraoperative rupture ("surgical spill") is IC1; capsule ruptured before surgery or tumor on ovarian or fallopian tube surface is IC2; and positive peritoneal cytology with or without rupture is IC3. The new staging includes a revision of stage III patients; assignment to stage IIIA1 is based on spread to the retroperitoneal lymph nodes without intraperitoneal dissemination. Extension of tumor from omentum to spleen or liver (stage IIIC) should be differentiated from isolated parenchymal metastases (stage IVB). © 2015 American Cancer Society.

  3. Reliability analysis for the creep rupture mode of failure

    International Nuclear Information System (INIS)

    Vaidyanathan, S.

    1975-01-01

    An analytical study has been carried out to relate the factors of safety employed in the design of a component to the probability of failure in the thermal creep rupture mode. The analysis considers the statistical variations in the operating temperature, stress and rupture time, and applies the life fraction damage criterion as the indicator of failure. Typical results for solution annealed type 304-stainless steel material for the temperature and stress variations expected in an LMFBR environment have been obtained. The analytical problem was solved by considering the joint distribution of the independent variables and deriving the distribution for the function associated with the probability of failure by integrating over proper regions as dictated by the deterministic design rule. This leads to a triple integral for the final probability of failure where the coefficients of variation associated with the temperature, stress and rupture time distributions can be specified by the user. The derivation is general, and can be used for time varying stress histories and the case of irradiated material where the rupture time varies with accumulated fluence. Example calculations applied to solution annealed type 304 stainless steel material have been carried out for an assumed coefficient of variation of 2% for temperature and 6% for stress. The results show that the probability of failure associated with dependent stress intensity limits specified in the ASME Boiler and Pressure Vessel Section III Code Case 1592 is less than 5x10 -8 . Rupture under thermal creep conditions is a highly complicated phenomenon. It is believed that the present study will help in quantizing the reliability to be expected with deterministic design factors of safety

  4. Real-Time Detection of Rupture Development: Earthquake Early Warning Using P Waves From Growing Ruptures

    Science.gov (United States)

    Kodera, Yuki

    2018-01-01

    Large earthquakes with long rupture durations emit P wave energy throughout the rupture period. Incorporating late-onset P waves into earthquake early warning (EEW) algorithms could contribute to robust predictions of strong ground motion. Here I describe a technique to detect in real time P waves from growing ruptures to improve the timeliness of an EEW algorithm based on seismic wavefield estimation. The proposed P wave detector, which employs a simple polarization analysis, successfully detected P waves from strong motion generation areas of the 2011 Mw 9.0 Tohoku-oki earthquake rupture. An analysis using 23 large (M ≥ 7) events from Japan confirmed that seismic intensity predictions based on the P wave detector significantly increased lead times without appreciably decreasing the prediction accuracy. P waves from growing ruptures, being one of the fastest carriers of information on ongoing rupture development, have the potential to improve the performance of EEW systems.

  5. The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Guntay, Salih; Dehbi, Abdel; Suckow, Detlef; Birchley, Jon

    2002-01-01

    Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes' models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an un-isolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (Aerosol Trapping In a Steam generator), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2

  6. Analysis of autofrettaged metal tubes

    International Nuclear Information System (INIS)

    Malik, M. Afzaal; Khan, Muddasar; Rashid, Badar; Khushnood, Shahab

    2007-01-01

    Thick-walled cylinders are widely used as compressor cylinders, pump cylinders, high pressure tubing, process reactors and vessels, nuclear reactors, isostatic vessels and gun barrels. In practice, cylinders are generally subjected to sudden and frequently drastic pressure fluctuations, such as the pressure generated in a gun barrel upon the firing of the weapon, pressure reversals in pump cylinders or in process reactors employing high-pressure piping, necessitating enhanced strength of such cylinders. A process for enhancing the strength of thick-walled cylinders has been in service, and is referred to as 'autofrettage'. It extends the service life of the cylinder. The autofrettage is achieved by increasing elastic strength of a cylinder with various methods such as hydraulic pressurization, mechanical swaging, or by utilizing the pressure of a powder gas. This research work deals with the hydraulic and mechanical autofrettage of metal tubes with the objective to attain enhanced strength. Five metal tubes are taken randomly for analysis purpose. The experimental data for five metal tubes is obtained to analyze the behavior of different parameters used during, before, and after autofrettage process. For this research, two-stage autofrettage is taken into consideration. The modeling of the metal tube is carried out in WildFire-ProEngineering, and for analysis purpose, finite element software ANSYS7 and COSMOS are used. The graphical analysis of swage autofrettage is carried out using MATLAB7. The results are validated using available experimental and numerical data. (author)

  7. A composite model for a class of electric-discharge shock tubes

    Science.gov (United States)

    Elkins, R. T.; Baganoff, D.

    1973-01-01

    A gasdynamic model is presented and analyzed for a class of shock tubes that utilize both Joule heating and electromagnetic forces to produce high-speed shock waves. The model consists of several stages of acceleration in which acceleration to sonic conditions is achieved principally through heating, and further acceleration of the supersonic flow is obtained principally through use of electromagnetic forces. The utility of the model results from the fact that it predicts a quasi-steady flow process, mathematical analysis is straightforward, and it is even possible to remove one or more component stages and still have the model related to a possible shock-tube flow. Initial experiments have been performed where the electrical discharge configuration and current level were such that Joule heating was the dominant form of energy addition present. These experiments indicate that the predictions of the model dealing with heat addition correspond quite closely to reality. The experimental data together with the theory show that heat addition to the flowing driver gas after diaphragm rupture (approach used in the model) is much more effective in producing high-speed shock waves than heating the gas in the driver before diaphragm rupture, as in the case of the arc-driven shock tube.

  8. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

    Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

  9. Metallurgical analysis of high pressure gas pipelines rupture

    International Nuclear Information System (INIS)

    Hasan, F.; Ahmed, F.

    2007-01-01

    On 6 July 2004, two parallel-running gas pipelines (18-inch and 24-inch diameters), in the main transmission network of SNGPL (a gas company in Pakistan) were ruptured. The ruptures occurred in the early hours of the morning about 8 miles downstream of the compressor station AC-4. The ruptures were indicated by the increased gas flow at the outlet of AC-4 (1), first at about 0648 hours and then again about 20 minutes later. The gas escaping from the ruptured lines had caught fire, and the flames had also 'affected' a third parallel-running pipeline of 30-inch diameter, lying next to the 24-inch line. The metallurgical examination of the two ruptured lines showed that the 24-inch line was ruptured with the help of an explosive device that had been placed on the underside of the pipe. An examination of the 18-inch line showed that this pipe had failed as a result of the heating of the pipe-wall, presumably, by the flame emanating from the 24-inch line. These two observations clearly suggested that the 24-inch line was the first to rupture (by explosives), and the fire following this rupture had heated the 18-inch pipe to a temperature where its yield strength was unable to support the inside gas pressure. The 20 minutes time interval between the two ruptures was obviously the time taken by the 18 inch pipe to be heated upto the level where it started to yield. The 30-inch line lying next to the 24-inch line was affected to the extent that its coating had been burnt-off over a length of about 40-50 feet. However, the pipe did not exhibit any signs of deshaping or deformation what-so-ever. A replica metallographic examination indicated that the microstructure of the pipe was not measurably affected by the heat. It was thus decided not to replace the affected part of the 30-inch pipe, but only to re-coat this affected portion. (author)

  10. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

    Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.; Nitheanandan, T.; Sanderson, D.B.

    1997-07-01

    The Zircaloy-2 calandria tube has been improved to guard against abnormal operating conditions. It has been strengthened by either thickening or eliminating the weld to withstand the consequences of a pressure tube rupture. To exploit the moderator as a heat sink, both surfaces have been roughened and the inside surface ridged to maximise heat-transfer from an over-heated fuel channel during a postulated loss of coolant accident. (author)

  11. Proposals for investigating instrument tube line breaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Charlton, T.R.; Loomis, G.G.; Hall, D.G.; Cozzuol, J.M.

    1985-11-01

    Questions posed by the NRC pertaining to instrument tube critical flow and applicability of the Semiscale experimental facility are evaluated. A program is recommended to investigate the issue of generic PWR safety following hypothetical rupture of instrument tubes due to consequences of seismic events

  12. Creep Rupture Life Prediction Based on Analysis of Large Creep Deformation

    Directory of Open Access Journals (Sweden)

    YE Wenming

    2016-08-01

    Full Text Available A creep rupture life prediction method for high temperature component was proposed. The method was based on a true stress-strain elastoplastic creep constitutive model and the large deformation finite element analysis method. This method firstly used the high-temperature tensile stress-strain curve expressed by true stress and strain and the creep curve to build materials' elastoplastic and creep constitutive model respectively, then used the large deformation finite element method to calculate the deformation response of high temperature component under a given load curve, finally the creep rupture life was determined according to the change trend of the responsive curve.The method was verified by durable test of TC11 titanium alloy notched specimens under 500 ℃, and was compared with the three creep rupture life prediction methods based on the small deformation analysis. Results show that the proposed method can accurately predict the high temperature creep response and long-term life of TC11 notched specimens, and the accuracy is better than that of the methods based on the average effective stress of notch ligament, the bone point stress and the fracture strain of the key point, which are all based on small deformation finite element analysis.

  13. Life prediction of steam generator tubing due to stress corrosion crack using Monte Carlo Simulation

    International Nuclear Information System (INIS)

    Hu Jun; Liu Fei; Cheng Guangxu; Zhang Zaoxiao

    2011-01-01

    Highlights: → A life prediction model for SG tubing was proposed. → The initial crack length for SCC was determined. → Two failure modes called rupture mode and leak mode were considered. → A probabilistic life prediction code based on Monte Carlo method was developed. - Abstract: The failure of steam generator tubing is one of the main accidents that seriously affects the availability and safety of a nuclear power plant. In order to estimate the probability of the failure, a probabilistic model was established to predict the whole life-span and residual life of steam generator (SG) tubing. The failure investigated was stress corrosion cracking (SCC) after the generation of one through-wall axial crack. Two failure modes called rupture mode and leak mode based on probabilistic fracture mechanics were considered in this proposed model. It took into account the variance in tube geometry and material properties, and the variance in residual stresses and operating conditions, all of which govern the propagations of cracks. The proposed model was numerically calculated by using Monte Carlo Simulation (MCS). The plugging criteria were first verified and then the whole life-span and residual life of the SG tubing were obtained. Finally, important sensitivity analysis was also carried out to identify the most important parameters affecting the life of SG tubing. The results will be useful in developing optimum strategies for life-cycle management of the feedwater system in nuclear power plants.

  14. Assessment of the residual time to rupture of fuel pins after reactor core disturbances using the Lebensanteil rule

    International Nuclear Information System (INIS)

    Schaefer, L.; Wassilew, C.

    1992-01-01

    An important aspect of disturbances in the reactor core is the way in which they affect the service life of fuel rod cladding tubes. This factor also determines whether and how long the reactor core can be continued in operation, i.e., matters of safety and economy are involved. Potential disturbances of the reactor core affect the fuel rod cladding tubes as increases in temperature and, sometimes, as mechanical stresess for limited periods of time. As thermomechanical stresses acting on a cladding tube also give rise to creep events which may limit the service life of fuel elements, it is important to know how much creep life or time to rupture is consumed in the course of a core disturbance, and what the residual life is. For this purpose, the times to rupture before and during the accident must be added up and the balance calculated. As a rule of computation, the Lebensanteil rule is used in its form expressing the time to rupture of creeping solids. The simulation of accidents with unirradiated cladding tubes revealed a drastic decrease of the residual time to rupture in those cases in which the cladding material had recrystallized. On the other hand, because of its structural stability, irradiated material turned out to be almost insensitive even under extremely severe accident conditions. The materials data so far available are sufficient for useful estimates. Presuming one of the damage accumulating processes of the creeping cladding material is predominant, there are no further validity limiting ranges concerning kind of accident, loading condition, cladding material and so on. (orig.)

  15. Use of generalized regression models for the analysis of stress-rupture data

    International Nuclear Information System (INIS)

    Booker, M.K.

    1978-01-01

    The design of components for operation in an elevated-temperature environment often requires a detailed consideration of the creep and creep-rupture properties of the construction materials involved. Techniques for the analysis and extrapolation of creep data have been widely discussed. The paper presents a generalized regression approach to the analysis of such data. This approach has been applied to multiple heat data sets for types 304 and 316 austenitic stainless steel, ferritic 2 1 / 4 Cr-1 Mo steel, and the high-nickel austenitic alloy 800H. Analyses of data for single heats of several materials are also presented. All results appear good. The techniques presented represent a simple yet flexible and powerful means for the analysis and extrapolation of creep and creep-rupture data

  16. Representative stresses for creep deformation and failure of pressurised tubes and pipes

    International Nuclear Information System (INIS)

    Cane, B.J.; Browne, R.J.

    1982-01-01

    Results of a series of tube and uniaxial tests on two casts of 1/2CrMoV pipe steel are examined to determine the representative stress which must be applied to uniaxial data in order to predict the strain rates and lives of pressurised tubes and pipes. The stress criterion for deformation is shown to correlate with the analytically derived reference stress (σsub(R)) at low pressure while at high pressures a modified reference stress (> σsub(R)) must be used. The rupture life exhibits a similar correlation such that the representative stress for rupture is given by σsub(R) at low stresses yet, at high stresses, it is greater than σsub(R) and attains a value comparable with the mean diameter hoop stress. The latter thus describes the rupture life at high pressures but significantly underestimates the life at low pressures approaching those in service. Consideration is given to the multiaxial stress rupture criterion and the effect of geometry in constant load tests. (author)

  17. Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Hong, Seong-Wan

    2011-01-01

    A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

  18. Creep-rupture, steam oxidation and recovery behaviours upon dynamic transients up to 1300 C of cold-worked 304 stainless steel tubes dedicated to nuclear core fuel cladding

    International Nuclear Information System (INIS)

    Portier, L.; Brachet, J.C.; Vandenberghe, V.; Guilbert, T.; Lezaud-Chaillioux, V.; Bernard, C.; Rabeau, V.

    2011-01-01

    An ambitious mechanical tests program was conducted on the fuel rod cladding of the CABRI facility between 2004 and 2009 to re-evaluate the cladding tubes materials behaviour. As an offspring of this major scientific investment several conclusions of interest could be drawn on the 304 stainless steel material. In particular, the specific behaviour of the materials during hypothetical and extreme 'dry-out' conditions was investigated. In such a scenario, the cladding tube materials should experience a very brief incursion at high temperatures, in a steam environment, up to 1300 C, before cladding rewetting. Some of the measurements performed in the range of interest for the safety case were on purpose developed beyond the conservatively safe domain. Some of the results obtained for these non-conventional heating rates, pressures and temperature ranges will be presented. First in order to assess the high temperature creep-rupture material behaviour under internal pressure upon dynamic transient conditions, tests have been performed on cold-worked 304 stainless cladding tubes in a steam environment, for heating rates up to 100 C*s -1 and pressure ramp rates up to 10 bar*s -1 thanks to the use of the EDGAR facility. Other tests performed at a given pressure allowed us to check the steady-state secondary creep rate of the materials in the 1100-1200 C temperature range. It was also possible to determine the rupture strength value and the failure mode as a function of the thermal and pressure loading history applied. It is worth noticing that, for very specific conditions, a surprising pure intergranular brittle failure mode of the clad has been observed. Secondly, in order to check the materials oxidation resistance of the materials, two-side steam oxidation tests have been performed at 1300 C, using the DEZIROX facility. It was shown that, thanks to the use of Ring Compression tests, the 304 cladding tube keeps significant ductility for oxidation times up to at least

  19. Hybrid friction diffusion bonding of 316L stainless steel tube-to-tube sheet joints for coil-wound heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Haneklaus, Nils; Cionea, Cristian; Reuven, Rony; Frazer, David; Hosemann, Peter; Peterson, Per F. [Dept of Nuclear Engineering, University of California, Berkeley (United States)

    2016-11-15

    Hybrid friction diffusion bonding (HFDB) is a solid-state bonding process first introduced by Helmholtz-Zentrum Geesthacht to join aluminum tube-to-tube sheet joints of Coil-wound heat exchangers (CWHE). This study describes how HFDB was successfully used to manufacture 316L test samples simulating tube-to-tube sheet joints of stainless steel CWHE for molten salt coolants as foreseen in several advanced nuclear- and thermal solar power plants. Engineering parameters of the test sample fabrication are presented and results from subsequent non-destructive vacuum decay leak testing and destructive tensile pull-out testing are discussed. The bonded areas of successfully fabricated samples as characterized by tube rupture during pull-out tensile testing, were further investigated using optical microscopy and scanning electron microscopy including electron backscatter diffraction.

  20. The effect of the number of condensed phases modeled on aerosol behavior during an induced steam generator tube rupture sequence

    International Nuclear Information System (INIS)

    Bixler, N.E.; Schaperow, J.H.

    1998-06-01

    VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS. A recently completed independent peer review of VICTORIA, while confirming the overall adequacy of the code, recommended a number of modeling improvements. One of these recommendations, to model three rather than a single condensed phase, is the focus of the work reported here. The recommendation has been implemented as an option so that either a single or three condensed phases can be treated. Both options have been employed in the study of fission product behavior during an induced steam generator tube rupture sequence. Differences in deposition patterns and mechanisms predicted using these two options are discussed

  1. Dynamic rupture analysis of reinforced concrete shells

    International Nuclear Information System (INIS)

    Rebora, B.; Zimmermann, Th.; Wolf, J.P.

    1976-01-01

    Extreme dynamic loading conditions often require the rupture analysis of reinforced and prestressed-concrete structures. The study presented in this paper extends a method of analysis of dynamic loading conditions which has proven efficient for short-time loads. Another aim is to adapt the method to thin-walled structures. It is not sufficient to work only with plastic rupture and yield surfaces locally which are compared to the elastic distribution of the stress resultants; it is essential to account for the redistribution of the latter. The method proposed consists of discretizing the structure into isoparametric three-dimensional elements with 20 nodes for the concrete and one-dimensional bar elements with three nodes for the steel. The latter can also be handled with a 'smeared' two-dimensional membrane element. In compression a three-dimensional non-linear elastic constitutive law is introduced for the concrete, and a triaxial failure surface expressed in the stress invariants is used, determining cracking and crushing. Two- and three-dimensional cracking surfaces in which no components of stress are transmitted are accounted for. The possibility exists that, during the history of loading, cracks can close up again. For steel, a yield criterion is selected. The non-linear analysis is based on the concept of initial stress. Residual loads are calculated using information in Gauss integration points. The ultimate load is reached when the algorithm does not converge. The corresponding failure modes can be interpreted as those for which a state of equilibrium is no longer possible. The equations of motion are discretized in time, using an extension of the linear acceleration method. (Auth.)

  2. A fast prediction of plant behaviour in the steam generator tube rupture accident at Mihama unit 2 using a similar case

    International Nuclear Information System (INIS)

    Gofuku, Akio; Tanaka, Yutaka; Numoto, Atsushi; Yoshikawa, Hidekazu.

    1996-01-01

    It is important to predict fast and accurately future trend of behaviour of a nuclear power plant in an emergency situation. The case-based reasoning is a strong tool for this purpose because it solves a problem by effectively using past similar cases. This study investigates the applicability of the case-based reasoning as a fast prediction technique of plant behaviour. This paper discusses a prediction of initial plant behaviour in the steam generator tube rupture accident happened at the Mihama nuclear power plant unit 2 by using the behaviour data of an accident of the same type happened at Prairie Island nuclear power plant unit 1. The prediction results coincide well with the reported plant behaviour although there are several important differences in the detailed plant specifications and operator actions between the two SGTR accidents. (author)

  3. Radiologic analysis of the medical collateral ligament rupture

    International Nuclear Information System (INIS)

    Cho, Chung Che; Lee, Chang Jun; Kim, Kun Sang; Park, Soo Soung

    1979-01-01

    The medical collateral ligament rupture is the most common injury involving the knee joint ligaments. The ruptured medical collateral ligaments of 73 cases with clinical and surgical confirmations were radiologically analyzed. The results were obtained as follows: 1. The most risky age for tearing of the medical collateral ligament was third to fifth decades (50 cases of male and 23 of females). 2. The most common cause of the medical collateral ligament rupture was traffic accident (82.2%). 3. The mean distance of medial knee joint space was 7.9 ± 2.0 mm on the normal side and 13.7 ± 4.2 mm on the affected side. 4. The mean degree of knee joint space was 10.1 ± 2.5 on the normal side and 14.7 ± 3.8 on the affected side. 5. The fibula was the bone fractured most frequently in association with the medial collateral ligament rupture (30.6%).

  4. Preliminary analysis of the rupture process of 11 March 2011 Tohoku-Oki earthquake

    Science.gov (United States)

    Vilotte, J.; Satriano, C.; Dionicio, V.; Lancieri, M.; Bernard, P.

    2011-12-01

    The great 11 March 2011 Off the Pacific Coast of Tohoku earthquake (Mw 9.1) ruptured a ~ 200 km wide mega-thrust fault, with average displacement of ~15-20 m. The earthquake triggered a large devastating tsunami as well as strong ground motion along the east Honshu coastline. Seismic activity in this area is characterized by a number of large earthquakes with Mw ~7.2-7.9 along the down-dip portion of the mega-thrust seaward of Miyagi prefecture, with only few events of magnitude greater than 8 in last hundred years. This region was also recognized to have had a large tsunami earthquake in 869 with a source area estimated further offshore. The rupture process of the Tohoku-Oki earthquake is investigated here combining teleseismic short period P-waves back-projection imaging and broadband P-wave finite fault inversions, together with a preliminary broadband analysis of the Kik-net strong motion recordings across Japan. The main features of the Tohoku-Oki rupture process imaged by the short period (1s) back-projection are: an initial 70-80s radiation phase eastward of the epicenter, with a slow (~1-1.5 km/s) along-dip rupture propagation; a short radiation phase northward of the epicenter; and ultimately a southward radiation phase with a relatively faster rupture propagation. These features are robust and consistent using both the North American and European arrays configurations. At lower periods, the back-projection analysis reveals a shift in the radiation centroid seaward toward the trench. In contrast, the broadband (1-200s) P-waves finite fault inversion exhibits a quite complementary image with a first long period radiation phase up-dip of the epicenter followed by down-dip late southwestward radiation phase that remains however poorly constraint. The robustness and the resolution of both the back-projection and the finite fault inversion analysis are carefully assessed through bootstrap analysis, and the analysis of some of the main foreshocks and aftershocks

  5. Synthesis, characterization and histomorphometric analysis of cellular response to a new elastic DegraPol® polymer for rabbit Achilles tendon rupture repair.

    Science.gov (United States)

    Buschmann, Johanna; Calcagni, Maurizio; Bürgisser, Gabriella Meier; Bonavoglia, Eliana; Neuenschwander, Peter; Milleret, Vincent; Giovanoli, Pietro

    2015-05-01

    Tendon rupture repair is a surgical field where improvements are still required due to problems such as repeat ruptures, adhesion formation and joint stiffness. In the current study, a reversibly expandable and contractible electrospun tube based on a biocompatible and biodegradable polymer was implanted around a transected and conventionally sutured rabbit Achilles tendon. The material used was DegraPol® (DP), a polyester urethane. To make DP softer, more elastic and surgeon-friendly, the synthesis protocol was slightly modified. Material properties of conventional and new DP film electrospun meshes are presented. At 12 weeks post-surgery, tenocyte and tenoblast density, nuclei and width, collagen fibre structure and inflammation levels were analyzed histomorphometrically. Additionally, a comprehensive histological scoring system by Stoll et al. (2011) was used to compare healing outcomes. Results showed that there were no adverse reactions of the tendon tissue following the implant. No differences were found whether the DP tube was applied or not for both traditional and new DP materials. As a result, the new DP material was shown to be an excellent carrier for delivery of growth factors, stem cells and other agents responsible for tendon healing. Copyright © 2015 John Wiley & Sons, Ltd.

  6. Estimation of residual life of boiler tubes using steamside oxide scale thickness

    International Nuclear Information System (INIS)

    Vikrant, K.S.N.; Ramareddy, G.V.; Pavan, A.H.V.; Singh, Kulvir

    2013-01-01

    In thermal power plants, remaining-life-estimation of boiler tubes is required at regular intervals for a safer and a better functionality of boilers. In this paper, a new method is proposed for the residual life estimation of service exposed boiler tubes using Non-Destructive Ultrasonic Oxide scale thickness measurements, average metal temperature and creep master curve. While steady state conduction heat transfer equations are solved to calculate the average metal temperature, creep master curve is generated from short term stress rupture data of rupture life less than 5000 h on a virgin material. In the present study, the residual life of T22 (2.25Cr-1Mo) service exposed Platen Superheater tube is estimated using two master creep curves, i.e. Larson-Miller Parametric (LMP) method of standard ASME T22 creep data and Wilshire approach of short term stress rupture data of T22. As the residual life is calculated from fundamental conduction heat transfer theory and creep rupture data, the proposed method can be applied for different grades of boiler materials. -- Highlights: ► Residual life is calculated from non-destructive oxide scale thickness, creep master curve and average metal temperature. ► A new method is proposed for calculating residual life using above parameters and from conduction heat transfer principles. ► The method can be applied to different boiler grades for estimating residual life and hence the method is generic

  7. Severe accident analysis of a steam generator tube rupture accident using MAAP-CANDU to support level 2 PSA for the Point Lepreau Generating Station Refurbishment Project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

    2015-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP-CANDU code was used to simulate the progression of postulated severe core damage accidents and fission product releases. This paper discusses the results for the reference case of the Steam Generator Tube Rupture initiating event. The reference case, dictated by the Level 2 Probabilistic Safety Assessment, was extreme and assumed most safety-related plant systems were not available: all steam generator feedwater; the emergency water supply; the moderator, shield and shutdown cooling systems; and all stages of emergency core cooling. The reference case also did not credit any post Fukushima lessons or any emergency mitigating equipment. The reference simulation predicted severe core damage beginning at 3.7 h, containment failure at 6.4 h, moderator boil off by 8.2 h, and calandria vessel failure at 42 h. A total release of 5.3% of the initial inventory of radioactive isotopes of Cs, Rb and I was predicted by the end of the simulation (139 h). Almost all noble gas fission products were released to the environment, primarily after the containment failure. No hydrogen/carbon monoxide burning was predicted. (author)

  8. Analysis of the ballooning deformation of an internally pressurized thin-wall tube during fast thermal transients

    International Nuclear Information System (INIS)

    Lin, E.I.H.

    1977-01-01

    A large-strain, time-dependent thermoplastic analysis of ballooning deformation was developed. The true (or lagorithmic) strains, the Von Mises yield criterion and Prandtl-Reuss flow rules were used. The constitutive equation was expressed in terms of the temperature, effective stress, strain and strain rate. Material isotropy was assumed as a first approximation; note that at high temperatures even zircaloy tends to lose a substantial amount of its low-temperature anisotropy. The axisymmetry of ballooning was also assumed, which has actually been verified by numerous experiments to be accurate throughout the course of ballooning, except in the final stage when rupture is imminent. The thin-shell approximation was made, which proved to be adequate for the standard fuel claddings and which was advantageous in that the averaged state of stress was rendered determinate. The analysis led to a set of non-linear ordinary differential equations, which was then integrated by a fifth-order Runge-Kutta routine. The general formulation allows for a direct interpretation of the experimentally-observed effects of the heating rate and cladding axial constraints on the ballooning behavior. Its implications on the flow-blockage and cladding-rupture evaluations were discussed. The analysis was applied to zircaloy claddings subjected to simulated thermal transient conditions. Most of the required material properties were taken from the existing uniaxial tensile test data. Analyses were performed at a uniform heating rate of 115 0 C/sec with internal pressures ranging from 100 to 1200 psi. Satisfactory agreement was obtained between the predictions and the diametral strain-time data available from tube-burst tests

  9. Traumatic Rupture of the Posterior Urethra. Analysis of 87 Cases at ...

    African Journals Online (AJOL)

    Traumatic Rupture of the Posterior Urethra. Analysis of 87 Cases at the Conakry University Hospital. A B Diallo, M Barry, I Bah, A T Diallo, O R Bah, A Toure, S Balde, K B Sow, S Guirassay, M B Diallo ...

  10. Fluid-Structure Interaction Analysis of Ruptured Mitral Chordae Tendineae.

    Science.gov (United States)

    Toma, Milan; Bloodworth, Charles H; Pierce, Eric L; Einstein, Daniel R; Cochran, Richard P; Yoganathan, Ajit P; Kunzelman, Karyn S

    2017-03-01

    The chordal structure is a part of mitral valve geometry that has been commonly neglected or simplified in computational modeling due to its complexity. However, these simplifications cannot be used when investigating the roles of individual chordae tendineae in mitral valve closure. For the first time, advancements in imaging, computational techniques, and hardware technology make it possible to create models of the mitral valve without simplifications to its complex geometry, and to quickly run validated computer simulations that more realistically capture its function. Such simulations can then be used for a detailed analysis of chordae-related diseases. In this work, a comprehensive model of a subject-specific mitral valve with detailed chordal structure is used to analyze the distinct role played by individual chordae in closure of the mitral valve leaflets. Mitral closure was simulated for 51 possible chordal rupture points. Resultant regurgitant orifice area and strain change in the chordae at the papillary muscle tips were then calculated to examine the role of each ruptured chorda in the mitral valve closure. For certain subclassifications of chordae, regurgitant orifice area was found to trend positively with ruptured chordal diameter, and strain changes correlated negatively with regurgitant orifice area. Further advancements in clinical imaging modalities, coupled with the next generation of computational techniques will enable more physiologically realistic simulations.

  11. Global catalog of earthquake rupture velocities shows anticorrelation between stress drop and rupture velocity

    Science.gov (United States)

    Chounet, Agnès; Vallée, Martin; Causse, Mathieu; Courboulex, Françoise

    2018-05-01

    Application of the SCARDEC method provides the apparent source time functions together with seismic moment, depth, and focal mechanism, for most of the recent earthquakes with magnitude larger than 5.6-6. Using this large dataset, we have developed a method to systematically invert for the rupture direction and average rupture velocity Vr, when unilateral rupture propagation dominates. The approach is applied to all the shallow (z earthquakes of the catalog over the 1992-2015 time period. After a careful validation process, rupture properties for a catalog of 96 earthquakes are obtained. The subsequent analysis of this catalog provides several insights about the seismic rupture process. We first report that up-dip ruptures are more abundant than down-dip ruptures for shallow subduction interface earthquakes, which can be understood as a consequence of the material contrast between the slab and the overriding crust. Rupture velocities, which are searched without any a-priori up to the maximal P wave velocity (6000-8000 m/s), are found between 1200 m/s and 4500 m/s. This observation indicates that no earthquakes propagate over long distances with rupture velocity approaching the P wave velocity. Among the 23 ruptures faster than 3100 m/s, we observe both documented supershear ruptures (e.g. the 2001 Kunlun earthquake), and undocumented ruptures that very likely include a supershear phase. We also find that the correlation of Vr with the source duration scaled to the seismic moment (Ts) is very weak. This directly implies that both Ts and Vr are anticorrelated with the stress drop Δσ. This result has implications for the assessment of the peak ground acceleration (PGA) variability. As shown by Causse and Song (2015), an anticorrelation between Δσ and Vr significantly reduces the predicted PGA variability, and brings it closer to the observed variability.

  12. Ruptured liver abscess: Analysis of 50 cases

    Directory of Open Access Journals (Sweden)

    Mohit Bhatia

    2017-01-01

    Full Text Available Background: Liver abscess (pyogenic and amebic is frequently encountered clinical condition; however, it can result in lethal outcome if there is a delay in diagnosis and treatment. Despite modalities to diagnose the condition early, still ruptured liver abscess presents with a common cause of acute abdomen in surgical emergency. In developing countries, ruptured liver abscess is a common cause of mortality. For contained abscess, nonsurgical options are considered; however, for ruptured liver abscess, surgical intervention is considered necessary. Materials and Methods: This was a retrospective study carried in Safdarjung hospital, New Delhi, between 2015 and 2016. All patients with ruptured liver abscess (clear signs of peritonitis were included in this study, and those patients having other causes of peritonitis were excluded. A preformed protocol for management was followed for all the patients, and various parameters contributing to the illness and its prognosis were evaluated and assessed. Results: Out of the fifty patients assessed, male patients were mainly affected (86%. The most affected age group was 31–40 years (64% followed by 41–50 years (22%. Right hypochondrium pain was the most common presenting complaint. Nine patients (18% had presented with signs of toxemia. Only right lobe of the liver was affected the most in 44 patients (88%. Escherichia coli was the most common organism isolated in our study in 19 patients (38%. A total of 19 patients (38% had diabetes in our study and total of 13 patients had mortality in our study. Conclusion: Ruptured liver abscess most commonly involves the right lobe of the liver. Males are affected far higher than the females; probable cause believed to be higher alcohol consumption. Most common affected age group falls between 30 and 60 years of age. If prompt treatment is started in time, mortality involved with it is evitable.

  13. How safe is defect specific maintenance of steam generator tubes?

    International Nuclear Information System (INIS)

    Dvorsek, T.; Cizelj, L.

    1995-01-01

    Outside diameter stress corrosion cracking at the tube to tube support plate intersections is assessed in the paper. The impact of defect specific maintenance on steam generator operation safety and reliability was investigated. This was performed by comparing efficiencies of defect specific and traditional maintenance strategy. The efficiency was studied through expected primary-to-secondary leak rate and tube rupture probability in a case of postulated accidental operating conditions, and number of tubes which shall be plugged using both maintenance strategies. In general, the efficiency of specific maintenance is function of particular steam generator and operating cycle. (author)

  14. Formation of a cavitation cluster in the vicinity of a quasi-empty rupture

    Science.gov (United States)

    Bol'shakova, E. S.; Kedrinskiy, V. K.

    2017-09-01

    The presentation deals with one of the experimental and numerical models of a quasi-empty rupture in the magma melt. This rupture is formed in the liquid layer of a distilled cavitating fluid under shock loading within the framework of the problem formulation with a small electromagnetic hydrodynamic shock tube. It is demonstrated that the rupture is shaped as a spherical segment, which retains its topology during the entire process of its evolution and collapsing. The dynamic behavior of the quasi-empty rupture is analyzed, and the growth of cavitating nuclei in the form of the boundary layer near the entire rupture interface is found. It is shown that rupture implosion is accompanied by the transformation of the bubble boundary layer to a cavitating cluster, which takes the form of a ring-shaped vortex floating upward to the free surface of the liquid layer. A p-κ mathematical model is formulated, and calculations are performed to investigate the implosion of a quasi-empty spherical cavity in the cavitating liquid, generation of a shock wave by this cavity, and dynamics of the bubble density growth in the cavitating cluster by five orders of magnitude.

  15. Deformation behavior of Zircaloy-4 cladding tubes under inert gas conditions in the temperature range from 600 to 12000C

    International Nuclear Information System (INIS)

    Hofmann, P.; Raff, S.; Gausmann, G.

    1981-07-01

    Within the temperature range from 600 0 to 1200 0 isothermal, isobaric creep rupture experiments were performed under inert gas with short Zircaloy-4 tube specimens in order to obtain experimental data supporting the development of the NORA cladding tube deformation model. The values of the tube inner pressure were so selected that the time-to-failure values varied between 2 and 2000 s. The corresponding creep rupture curves are indicated. Besides the temperature and the burst pressure the development of deformation over time of the tube specimens was measured. This allowed to draw diagrams of stress, strain rate and strain. On account of the type of specimen heating applied (radiation heating) the temperature difference at the cladding tube circumference is very small ( [de

  16. Metallurgical Analysis of Cracks Formed on Coal Fired Boiler Tube

    Science.gov (United States)

    Kishor, Rajat; Kyada, Tushal; Goyal, Rajesh K.; Kathayat, T. S.

    2015-02-01

    Metallurgical failure analysis was carried out for cracks observed on the outer surface of a boiler tube made of ASME SA 210 GR A1 grade steel. The cracks on the surface of the tube were observed after 6 months from the installation in service. A careful visual inspection, chemical analysis, hardness measurement, detailed microstructural analysis using optical and scanning electron microscopy coupled with energy dispersive X-ray spectroscopy were carried out to ascertain the cause for failure. Visual inspection of the failed tube revealed the presence of oxide scales and ash deposits on the surface of the tube exposed to fire. Many cracks extending longitudinally were observed on the surface of the tube. Bulging of the tube was also observed. The results of chemical analysis, hardness values and optical micrographs did not exhibit any abnormality at the region of failure. However, detailed SEM with EDS analysis confirmed the presence of various oxide scales. These scales initiated corrosion at both the inner and outer surfaces of the tube. In addition, excessive hoop stress also developed at the region of failure. It is concluded that the failure of the boiler tube took place owing to the combined effect of the corrosion caused by the oxide scales as well as the excessive hoop stress.

  17. Analysis of the State of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Bergunker, Olga

    2008-01-01

    The problem of safe operation of SG heat exchanging tubes, of both economical and effective control of their state is still important these days. Issues connected with peculiarities of methods of SG tubes inspection, automated analysis of the inspection results, tubes state analysis and development of algorithms of forecasting their state are considered in this report. The need for effective use of extensive data arrays on SG operation has led to the necessity of creating software tools for collection, storage and analysis of these data. The data-analytical system 'NPP Steam Generators' meant for data systematization and visualization as well as various types of analyses of data on eddy current inspection of WWER-440 and WWER-1000 SG tubes is presented in this report. The main possibilities of the data-analytical system (DAS), the code current state and prospects of its development are shown. The main fields of DAS application are considered and some results of its practical use are mentioned, namely, in the field of forecasting SG tubes state. (authors)

  18. Clinical diagnosis of an anterior cruciate ligament rupture : A meta-analysis

    NARCIS (Netherlands)

    Benjammse, A; Gokeler, A; van der Schans, CP

    Study Design: Meta-analysis. Objectives: To define the accuracy of clinical tests for assessing anterior cruciate ligament (ACL) ruptures. Background: The cruciate ligaments, and especially the ACL, are among the most commonly injured structures of the knee. Given the increasing injury prevalence,

  19. A reappraisal of steam generator tube rupture in the French licensing process

    International Nuclear Information System (INIS)

    Conte, M.; Gouffon, A.; Moriette, P.

    1984-10-01

    Upon the examination of the safety options submitted by EDF (Electricite de France) for a new pressurized water reactor design (N4, 1400 MWe), the French safety authorities decided that the conventionnal list of events to take under consideration should be amended as follows: failure of 1 and 2 steam generator tubes. To meet these objectives, design improvements were decided and new operating criteria were required by the technical specifications. Various preventive measures have been adopted by EDF to reduce tube degradation risks at the design stage, at the secondary feedwater quality level, and concerning also the quality control. The radiological consequences of generator tube integrity failure can be mitigated if the primary coolant activity is low, the tube flow detection is rapid, the release time is short, and the operating procedure is suitable and easily implemented [fr

  20. Unique case of esophageal rupture after a fall from height

    Directory of Open Access Journals (Sweden)

    van Berge Henegouwen Mark I

    2009-12-01

    Full Text Available Abstract Background Traumatic ruptures of the esophagus are relatively rare. This condition is associated with high morbidity and mortality. Most traumatic ruptures occur after motor vehicle accidents. Case Presentation We describe a unique case of a 23 year old woman that presented at our trauma resuscitation room after a fall from 8 meters. During physical examination there were no clinical signs of life-threatening injuries. She did however have a massive amount of subcutaneous emphysema of the chest and neck and pneumomediastinum. Flexible laryngoscopy revealed a lesion in the upper esophagus just below the level of the upper esophageal sphincter. Despite preventive administration of intravenous antibiotics and nutrition via a nasogastric tube, the patient developed a cervical abscess, which drained spontaneously. Normal diet was gradually resumed after 2.5 weeks and the patient was discharged in a reasonable condition 3 weeks after the accident. Conclusions This case report presents a high cervical esophageal rupture without associated local injuries after a fall from height.

  1. Moderator mixing after a pressure tube failure

    International Nuclear Information System (INIS)

    MacKinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

    1997-01-01

    During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system conditions investigated are of a reactor in a GSS, with coolant in the primary heat transport system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The

  2. Single chest tube drainage is superior to double chest tube drainage after lobectomy: a meta-analysis.

    Science.gov (United States)

    Zhou, Dong; Deng, Xu-Feng; Liu, Quan-Xing; Chen, Qian; Min, Jia-Xin; Dai, Ji-Gang

    2016-05-27

    In this meta-analysis, we conducted a pooled analysis of clinical studies comparing the efficacy of single chest tube versus double chest tube after a lobectomy. According to the recommendations of the Cochrane Collaboration, we established a rigorous study protocol. We performed a systematic electronic search of the PubMed, Embase, Cochrane Library and Web of Science databases to identify articles to include in our meta-analysis. A literature search was performed using relevant keywords. A meta-analysis was performed using RevMan© software. Five studies, published between 2003 and 2014, including 630 patients (314 patients with a single chest tube and 316 patients with a double chest tube), met the selection criteria. From the available data, the patients using a single tube demonstrated significantly decreased postoperative pain [weighted mean difference [WMD] -0.60; 95 % confidence intervals [CIs] -0.68-- 0.52; P tube after a pulmonary lobectomy. However, there were no significant differences in postoperative complications [OR 0.91; 95 % CIs 0.57-1.44; P = 0.67] and re-drainage rates [OR 0.81; 95 % CIs 0.42-1.58; P = 0.54]. Our results showed that a single-drain method is effective, reducing postoperative pain, hospitalization times and duration of drainage in patients who undergo a lobectomy. Moreover, the single-drain method does not increase the occurrence of postoperative complications and re-drainage rates.

  3. Analysis of MSGTR events for APR1400 by means of best estimate thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kim, Sang Jae; Chang, Keun Sun; Lee, Jae Hun

    2001-01-01

    A multiple steam generator tube rupture (MSGTR) event has never occurred in the history of commercial nuclear reactor operation while single steam generator tube rupture (SGTR) event is reported to occur every two years. As there is no history of MSGTR event, the understandings of transients and consequences of this event are not so much. In this study, a postulated MSGTR event in advanced power reactor 1400 (APR1400) is analyzed using thermal-hydraulic system code. The APR 1400 is a two-loop, 1000 MWe, PWR supposed to be built in 2009. MARS1.4 is used in this study. The present study aims to understand the effects of rupture location in heat transfer tubes and selection of affected steam generator following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 is to allow shortest time for operator action following a tubes rupture in the vicinity of hot-leg side tube sheet and to allow longest time following a tube ruptures at the tube top. The MSSV lift time for rupture at tube-top is evaluated as 24.5% larger than that for rupture at hot-leg side tube sheet. Also, the MSSV lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generator is affected. The comparison shows that the cases for both of two steam generators are affected allow longer time for operator action compared with the cases that a single steam generator is affected. Further more, the tube ruptures in the steam generator where a pressurizer is linked leads to the shortest operator response time

  4. A portable tube exciting X-ray fluorescence analysis system

    International Nuclear Information System (INIS)

    Yang Qiang; Lai Wanchang; Ge Liangquan

    2009-01-01

    Article introduced a portable tube exciting X-ray fluorescence analysis system which is based on arm architecture. Also, we designed Tube control circuit and finished preliminary application. The energy and the intensity of the photon can be adjusted continuously by using the tube. Experiments show that high excitation efficiency obtained by setting the appropriate parameters of the tube for the various elements. (authors)

  5. Tracheomegaly Secondary to Tracheotomy Tube Cuff in Amyotrophic Lateral Sclerosis: A Case Report.

    Science.gov (United States)

    Lee, Dong Hoon; Yoon, Tae Mi; Lee, Joon Kyoo; Lim, Sang Chul

    2015-10-01

    Tracheomegaly has not been reported in amyotrophic lateral sclerosis (ALS). Herein, the authors report a case of tracheomegaly secondary to tracheotomy tube cuff in a patient with ALS. To our knowledge, this is the first report of an ALS patient with tracheomegaly and of tracheomegaly being associated with tracheotomy tube cuff and home tracheotomy mechanical ventilator.The clinician should consider the possibility of tracheomegaly in the differential diagnosis, if a patient with ALS develops repeat air leakage around the tracheotomy tube or rupture of tracheotomy tube cuff.

  6. Measurement of the flow properties within a copper tube containing a deflagrating explosive

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Larry G [Los Alamos National Laboratory; Morris, John S [Los Alamos National Laboratory; Jackson, Scott I [Los Alamos National Laboratory

    2009-01-01

    We report on the propagation of deflagration waves in the high explosive (HE) PBX 9501 (95 wt % HMX, 5 wt% binder). Our test configuration, which we call the def1agration cylinder test (DFCT), is fashioned after the detonation cylinder test (DTCT) that is used to calibrate the JWL detonation product equation of state (EOS). In the DFCT, the HE is heated to a uniform slightly subcritical temperature, and is ignited at one end by a hot wire. For some configurations and initial conditions, we observe a quasi-steady wave that flares the tube into a funnel shape, stretching it to the point of rupture. This behavior is qualitatively like the DTCT, such that, by invoking certain additional approximations that we discuss, its behavior can be analyzed by the same methods. We employ an analysis proposed by G.I. Taylor to infer the pressure-volume curve for the burning, expanding flow. By comparing this result to the EOS of HMX product gas alone. we infer that only {approx}20 wt% of the HMX has burned at tube rupture. This result confirms pre-existing observations about the role of convective burning in HMX cookoff explosions.

  7. Functional rehabilitation of patients with acute Achilles tendon rupture: a meta-analysis of current evidence.

    Science.gov (United States)

    Mark-Christensen, Troels; Troelsen, Anders; Kallemose, Thomas; Barfod, Kristoffer Weisskirchner

    2016-06-01

    The optimal treatment for acute Achilles tendon rupture (ATR) is continuously debated. Recent studies have proposed that the choice of either operative or non-operative treatment may not be as important as rehabilitation, suggesting that functional rehabilitation should be preferred over traditional immobilization. The purpose of this meta-analysis of randomized controlled trials (RCTs) was to compare functional rehabilitation to immobilization in the treatment of ATR. This meta-analysis was conducted using the databases: PubMed, EMBASE, Rehabilitation and Sports Medicine Source, AMED, CINAHL, Cochrane Library and PEDro using the search terms: "Achilles tendon," "rupture," "mobilization" and "immobilization". Seven RCTs involving 427 participants were eligible for inclusion, with a total of 211 participants treated with functional rehabilitation and 216 treated with immobilization. Re-rupture rate, other complications, strength, range of motion, duration of sick leave, return to sport and patient satisfaction were examined. There were no statistically significant differences between groups. A trend favoring functional rehabilitation was seen regarding the examined outcomes. Functional rehabilitation after acute Achilles tendon rupture does not increase the rate of re-rupture or other complications. A trend toward earlier return to work and sport, and increased patient satisfaction was found when functional rehabilitation was used. The present literature is of low-to-average quality, and the basic constructs of the examined treatment and study protocols vary considerably. Larger, randomized controlled trials using validated outcome measures are needed to confirm the findings. II.

  8. Atmospheric Pressure and Abdominal Aortic Aneurysm Rupture: Results From a Time Series Analysis and Case-Crossover Study.

    Science.gov (United States)

    Penning de Vries, Bas B L; Kolkert, Joé L P; Meerwaldt, Robbert; Groenwold, Rolf H H

    2017-10-01

    Associations between atmospheric pressure and abdominal aortic aneurysm (AAA) rupture risk have been reported, but empirical evidence is inconclusive and largely derived from studies that did not account for possible nonlinearity, seasonality, and confounding by temperature. Associations between atmospheric pressure and AAA rupture risk were investigated using local meteorological data and a case series of 358 patients admitted to hospital for ruptured AAA during the study period, January 2002 to December 2012. Two analyses were performed-a time series analysis and a case-crossover study. Results from the 2 analyses were similar; neither the time series analysis nor the case-crossover study showed a significant association between atmospheric pressure ( P = .627 and P = .625, respectively, for mean daily atmospheric pressure) or atmospheric pressure variation ( P = .464 and P = .816, respectively, for 24-hour change in mean daily atmospheric pressure) and AAA rupture risk. This study failed to support claims that atmospheric pressure causally affects AAA rupture risk. In interpreting our results, one should be aware that the range of atmospheric pressure observed in this study is not representative of the atmospheric pressure to which patients with AAA may be exposed, for example, during air travel or travel to high altitudes in the mountains. Making firm claims regarding these conditions in relation to AAA rupture risk is difficult at best. Furthermore, despite the fact that we used one of the largest case series to date to investigate the effect of atmospheric pressure on AAA rupture risk, it is possible that this study is simply too small to demonstrate a causal link.

  9. Efficiency of defect specific maintenance od steam generator tubes: the case of ODSCC

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1996-01-01

    The outside diameter stress corrosion cracking at tube support plates became the dominating ageing mechanism in steam generators tubes made of Inconel 600. A variety of maintenance approaches were developed and implemented worldwide to deal with this mechanism. Despite different philosophical and physical backgrounds implemented, all of the applied approaches satisfy the relevant regulatory requirements. For our purpose, the maintenance approach consist of: (1) inspection of tubes, (2) accepting or rejecting the defective tube and (3) plugging of rejected tubes. The problem of selecting an optimal maintenance approach is raised in the paper. Consequently, a method comparing the efficiency of applicable maintenance approaches is proposed. The efficiency is defined by three parameters: (a) number of plugged tubes, (b) probability of steam generator tube rupture and (c) predicted accidental leak rates through the defects. An original probabilistic model is proposed to quantify the probability of tube rupture, while procedures available in literature were used to define the accidental leak rates. The numerical example considers the data from Krsko NPP (Westinghouse 632 MWe). The maintenance approaches analyzed include: (i) no repair at all, (ii) traditional defect depth (40%) based maintenance, (iii) alternate plugging criterion (bobbin coil voltage as defined by EPRI and U.S. NRC) and (iv) combined traditional and alternate approach. Advantages of the defect specific approaches (iii) and (iv) over the traditional one (defect depth) are clearly shown. A brief discussion on the optimization of safe life of steam generator is given. (author)

  10. Inner tubing technique used for the treatment of anastomotic aneurism.

    Science.gov (United States)

    Gaspar, Márcio Teodoro da Costa; de Mattos, Bruno Vinicius Hortences; Sofia, Milena Cristina Dias; Mulatti, Grace Carvajal; Lederman, Alex

    2016-01-01

    The authors report the case of a 66-year-old male patient diagnosed with a pseudoaneurysm of the distal aorto-aortic anastomosis treated with the inner tubing technique. The patient had been operated on 1 year before when he had an aortic prosthesis implanted as treatment for a ruptured abdominal aortic aneurysm. The inner tubing technique was developed to facilitate the treatment in bifurcated vascular lesions, where endovascular conventional prosthesis is not available.

  11. Ectopic pregnancy with tubal rupture: an analysis of 80 cases

    International Nuclear Information System (INIS)

    Ashfaq, S.; Aziz, S.; Hasan, M.; Sultan, S.; Irfan, S.M.

    2017-01-01

    Ectopic pregnancy (EP) is a major problem in obstetrics as there is evidence of increasing incidence throughout the world. It is an important cause of maternal morbidity and mortality. In Pakistan, the care seeking behaviour among female is limited that makes female vulnerable to die due to complication of ectopic pregnancy. The aim of this study is to determine the frequency of tubal rupture in ectopic pregnancy in Pakistani patients. Method: In this cross-sectional study data pertaining to age, gestational age, parity and duration of presenting symptoms were collected and analysed. Result: 80 patients were diagnosed to have ectopic pregnancy. The frequency of tubal rupture was 91.25%. It is encountered significantly more often in women with age of 26 years. More tubal rupture is found in patient with low parity, in which the frequency of tubal rupture is up to 100% and decrease up to 78.6% with increasing parity up to four. Furthermore, it is noted that increase in gestational age from 8 weeks to 10 weeks caused an increase in frequency of tubal rupture from 80 to 100% respectively. It is also noted that earlier the patient presents the lesser is the frequency of tubal rupture, as compared to late presentation beyond 3-4 days which make frequency up to 95%. Conclusion: Tubal rupture is still common cause of maternal morbidity and mortality, and is still a major challenge in gynaecological practice. Creating awareness amongst midwives and GPs regarding early diagnosis can contribute to decrease the mortality, morbidity and fertility loss related to EP. (author)

  12. Complete Right Main Bronchus Rupture in a Child: Report of a Case

    Directory of Open Access Journals (Sweden)

    Bayram Altuntas

    2014-03-01

    Full Text Available Blunt chest trauma resulting in rupture of a main bronchus is rare and probably have a high prehospital mortality.These injuries are often fatal because of respiratory distress and the high frequency of associated multiple organ injuries. A six-year-old boy was admitted our clinic due to blunt chest trauma. The tube thoracostomy was performed for the right pneumothorax at another surgical center. He was referred to our clinic due to inadequate expansion of the lung. On the physical examination, there was middle intercostal retraction, cyanosis and tachypnoea. The initial chest x-ray showed total pnemothorax on the right side and the hilum replaced by inferiorly. The rigid bronchoscopy was performed and the the rupture of main bronchus was seen. The sleeve upper lobectomy was performed. We aimed to emphasize the important of early diagnosis and treatment in the bronchial ruptures.

  13. An Analysis of Surgical Treatment for the Spontaneous Rupture of Hepatocellular Carcinoma.

    Science.gov (United States)

    Sada, Haruki; Ohira, Masahiro; Kobayashi, Tsuyoshi; Tashiro, Hirotaka; Chayama, Kazuaki; Ohdan, Hideki

    2016-01-01

    The prognosis of spontaneous rupture of hepatocellular carcinoma (HCC) remains unclear. We investigated the prognosis of patients with ruptured HCC based on the treatments and prognostic factors associated with long-term survival. The prognoses of 64 consecutive patients treated for ruptured HCC from 1986 to 2013 were analyzed according to their methods of treatment. The prognostic factors of 16 surgical patients were identified, and their overall survival (OS) and recurrence rates were compared to 1,157 surgical patients who underwent surgery for non-ruptured HCC. The surgical outcomes were also compared using a propensity score matching method. Surgery was associated with a better OS. Curative resection was the only independent prognostic factor in surgical patients with ruptured HCC (p = 0.040). Although the OS of surgical patients with non-ruptured HCC was found to be significantly better than that of the patients with ruptured HCC, no significant difference in OS was observed after propensity score matching. A curative resection should be the objective of treatment, assuming the suitability of the patient's clinical condition. When the liver function reserve and tumor extension of patients with ruptured and non-ruptured HCC are similar, then their surgical outcomes may not be significantly different. © 2015 S. Karger AG, Basel.

  14. A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

    2013-01-01

    The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

  15. Thermal creep properties of alloy D9 stainless steel and 316 stainless steel fuel clad tubes

    International Nuclear Information System (INIS)

    Latha, S.; Mathew, M.D.; Parameswaran, P.; Bhanu Sankara Rao, K.; Mannan, S.L.

    2008-01-01

    Uniaxial thermal creep rupture properties of 20% cold worked alloy D9 stainless steel (alloy D9 SS) fuel clad tubes for fast breeder reactors have been evaluated at 973 K in the stress range 125-250 MPa. The rupture lives were in the range 90-8100 h. The results are compared with the properties of 20% cold worked type 316 stainless steel (316 SS) clad tubes. Alloy D9 SS were found to have higher creep rupture strengths, lower creep rates and lower rupture ductility than 316 SS. The deformation and damage processes were related through Monkman Grant relationship and modified Monkman Grant relationship. The creep damage tolerance parameter indicates that creep fracture takes place by intergranular cavitation. Precipitation of titanium carbides in the matrix and chromium carbides on the grain boundaries, dislocation substructure and twins were observed in transmission electron microscopic investigations of alloy D9 SS. The improvement in strength is attributed to the precipitation of fine titanium carbides in the matrix which prevents the recovery and recrystallisation of the cold worked microstructure

  16. Stress analysis of a rupture disk

    International Nuclear Information System (INIS)

    Werne, R.W.

    1975-04-01

    The results of an elastic stress analysis of the rupture disk for an internal pressure of 45.5 MPa (6600 psi) indicate that the maximum von Mises stresses occur in the membrane and are on the order of 483 to 690 MPa (70,000 psi). This far exceeds the yield of the membrane material of 207 MPa (30,000 psi). These high stresses are expected since the membrane is designed to burst at that design pressure. The von Mises stresses in the rest of the body are less than 138 MPa (20,000 psi). An elastic-plastic analysis of the membrane alone subjected to the 45.5 MPa (6600 psi) pressure indicates that it becomes plastically unstable, i.e., it continues to deform under constant load. A second load case with a constant 6.9 MPa (1000 psi) pressure throughout the entire body (i.e., after release of pressure by burst of the membrane) was analyzed. The results indicate that the elastic von Mises stresses are less than 26.7 MPa (3880 psi) throughout the body. (U.S.)

  17. Safety significance of steam generator tube degradation mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G; Mignot, P [AIB-Vincotte Nuclear - AVN, Brussels (Belgium)

    1991-07-01

    Steam generator (SG) tube bundle is a part of the Reactor Coolant Pressure Boundary (RCPB): this means that its integrity must be maintained. However, operating experience shows various types of tube degradation to occur in the SG tubing, which may lead to SG tube leaks or SG tube ruptures and create a loss of primary system coolant through the SG, therefore providing a direct path to the environment outside the primary containment structure. In this paper, the major types of known SG tube degradations are described and analyzed in order to assess their safety significance with regard to SG tube integrity. In conclusion: The operational reliability and the safety of the PWR steam generator s requires a sufficient knowledge of the degradation mechanisms to determine the amount of degradation that a tube can withstand and the time that it may remain in operation. They also require the availability of inspection techniques to accurately detect and characterize the various degradations. The status of understanding of the major types of degradation summarized in this paper shows and justifies why efforts are being performed to improve the management of the steam generator tube defects.

  18. Data analysis for steam generator tubing samples

    International Nuclear Information System (INIS)

    Dodd, C.V.

    1996-07-01

    The objective of the Improved Eddy-Current ISI for Steam Generators program is to upgrade and validate eddy-current inspections, including probes, instrumentation, and data processing techniques for inservice inspection of new, used, and repaired steam generator tubes; to improve defect detection, classification and characterization as affected by diameter and thickness variations, denting, probe wobble, tube sheet, tube supports, copper and sludge deposits, even when defect types and other variables occur in combination; to transfer this advanced technology to NRC's mobile NDE laboratory and staff. This report provides a description of the application of advanced eddy-current neural network analysis methods for the detection and evaluation of common steam generator tubing flaws including axial and circumferential outer-diameter stress-corrosion cracking and intergranular attack. The report describes the training of the neural networks on tubing samples with known defects and the subsequent evaluation results for unknown samples. Evaluations were done in the presence of artifacts. Computer programs are given in the appendix

  19. Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

    International Nuclear Information System (INIS)

    Majumdar, S.

    1996-09-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents

  20. Analysis of Hydrogen Generation and Accumulation in U-233 Tube Vaults

    International Nuclear Information System (INIS)

    Ally, M.R.; Willis, K.J.

    1999-01-01

    The purpose of the 233 U Safe Storage Program is to enhance the safe storage of 233 U-bearing materials. This report describes the work done at the Oak Ridge National Laboratory's Radiochemical Development Facility (RDF) to address questions related to possible hydrogen generation and accumulation in 233 U tube vaults. The objective of this effort was to verify assumptions in the mathematical model used to estimate the hydrogen content of the gaseous atmosphere that possibly could occur inside the tube vaults in Building 3019 and to evaluate proposed measures for mitigating any hydrogen concerns. A mathematical model was developed using conservative assumptions to evaluate possible hydrogen generation and accumulation in the tube vaults. The model concluded that an equilibrium concentration would be established below the lower flammability limit (LFL) of 4.1% hydrogen. The major assumptions used in the model that were validated are as follows: (1) The shield plug does not form a seal with the tube vault wall, thus allowing the hydrogen gas to diffuse past the shield plug to the upper section of the tube vault. (2) The tube vault end-cap leaks sufficiently to allow air to be drawn into the tube vault by the off-gas system, thereby purging hydrogen from the upper section of the tube vault. (3) Any hydrogen gas generated completely mixes with the other gases present in the lower section of the tube vault and does not stratify beneath the shield plug. (4) The diffusion coefficient determined from the literature for constant diffusion of hydrogen in air is valid. The coefficient is corrected for temperatures from 0 to 25 C. Another assumption used in the model, that hydrogen generated by radiolytic decomposition of hydrogen-bearing materials (e.g., moisture and plastic) leaks from the cans under steady-state condition, as opposed to a sudden release resulting from rupture of the can(s), was beyond the scope of this investigation. Several parameters from the original

  1. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    International Nuclear Information System (INIS)

    Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

  2. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

  3. Tube Bulge Process : Theoretical Analysis and Finite Element Simulations

    International Nuclear Information System (INIS)

    Velasco, Raphael; Boudeau, Nathalie

    2007-01-01

    This paper is focused on the determination of mechanics characteristics for tubular materials, using tube bulge process. A comparative study is made between two different models: theoretical model and finite element analysis. The theoretical model is completely developed, based first on a geometrical analysis of the tube profile during bulging, which is assumed to strain in arc of circles. Strain and stress analysis complete the theoretical model, which allows to evaluate tube thickness and state of stress, at any point of the free bulge region. Free bulging of a 304L stainless steel is simulated using Ls-Dyna 970. To validate FE simulations approach, a comparison between theoretical and finite elements models is led on several parameters such as: thickness variation at the free bulge region pole with bulge height, tube thickness variation with z axial coordinate, and von Mises stress variation with plastic strain. Finally, the influence of geometrical parameters deviations on flow stress curve is observed using analytical model: deviations of the tube outer diameter, its initial thickness and the bulge height measurement are taken into account to obtain a resulting error on plastic strain and von Mises stress

  4. Rupture of primigravid uterus and recurrent rupture

    Directory of Open Access Journals (Sweden)

    Nahreen Akhtar

    2016-08-01

    Full Text Available Uterine rupture is a deadly obstetrical emergency endangering the life of both mother and fetus. In Bangladesh, majority of deliveries arc attended by unskilled traditional birth attendant and maternal mortality is still quite high. It is rare Ln developed country but unfortunately it is common in a developing country like Bangladesh. We report a case history of a patient age 32yrs from Daudkandi, Comilla admitted with H/0 previous two rupture uterus and repair with no living issue. We did caesarean section at her 31+ weeks of pregnancy when she developed Jabour pain. A baby of 1.4 kg was delivered. During cesarean section, focal rupture was noted in previous scar of rupture. Unfortunately the baby expired in neonatal ICU after 36 hours.

  5. Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

    1994-12-31

    The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

  6. Strength analysis of filament-wound composite tubes

    Directory of Open Access Journals (Sweden)

    Vasović Ivana

    2010-01-01

    Full Text Available The subject of this work is focused on strength analysis of filament-wound composite tubes made of E glass/polyester under internal pressure. The primary attention of this investigation is to develop a reliable computation procedure for stress, displacement and initial failure analysis of layered composite tubes. For that purpose we have combined the finite element method (FEM with corresponding initial failure criterions. In addition, finite element analyses using commercial code, MSC/NASTRAN, were performed to predict the behavior of filament wound structures. Computation results are compared with experiments. Good agreement between computation and experimental results are obtained.

  7. The temperature dependence of the tensile properties of thermally treated Alloy 690 tubing

    International Nuclear Information System (INIS)

    Harrod, D.L.; Gold, R.E.; Larsson, B.; Bjoerkman, G.

    1992-01-01

    Tensile tests were run in air on full tube cross-sections of 22.23 mm OD by 1.27 mm wall thickness Alloy 690 steam generator production tubes from ten (10) heats of material at eight (8) temperatures between room temperature and 760 degrees C. The tubing was manufactured to specification requirements consistent with the EPRI guidelines for Alloy 690 tubing. The room temperature stress-strain curves are described quite well by the Voce equation. Ductile fracture by dimpled rupture was observed at all test temperatures. The elevated temperature tensile properties are compared with design data given in the ASME Code

  8. Steam generator tube support plate degradation in French plants: maintenance strategy

    International Nuclear Information System (INIS)

    Gauchet, J.-P.; Gillet, N.; Stindel, M.

    1998-01-01

    This paper reports on the degradations of Steam Generator (SG) Tube Support Plates (TSPs) observed in French plants and the maintenance strategy adopted to continue operating the plant without any decrease of the required safety level. Only drilled carbon steel TSPs of early SGs are affected. Except the particular damage of the TSP8 of FESSENHEIM 2 caused by chemical cleaning procedures implemented in 1992, two main problems were observed almost exclusively on the upper TSP: Ligaments ruptured near the aseismic block located at 215 degrees. This degradation is perfectly detectable by bobbin coil inspection. It occurs very early in the life of the SG as can be seen from the records of previous inspections and no evolution of the signals was observed. This damage can be detected for 51M model SGs on several sites; Wastage of the ligaments resulting in enlargement of flow holes with in some cases complete consumption of a ligament. This damage was only observed for SGs of at GRAVELINES. This damage evolved cycle after cycle. Detailed studies were performed to analyze tubing behavior when a tube is not supported by the upper TSP because of missing ligaments. These studies evaluated the risk of vibratory instability, the behavior of both the TSP and the tubing in case of a seismic event or a LOCA and finally the behavior of the TSP in case of a Steam Line Break. Concerning vibratory instability it was possible to define zones where stability could not be demonstrated. Dampine, cables and sentinel plugs were then used when necessary to eliminate the risk of Steam Generator Tube Rupture (SGTR). For accidental conditions, it could be shown that no unacceptable damage occurs and that the core cooling function of the SG is always maintained if some tubes are plugged. From this analysis, It was possible to define the inspection programs for the different plants taking into account the specific situation of each plant regarding the damages detected. These programs include

  9. Spontaneous rupture of pheochromocytoma and its clinical features: a case report.

    Science.gov (United States)

    Maruyama, Mayumi; Sato, Haruhiro; Yagame, Mitsunori; Shoji, Sunao; Terachi, Toshiro; Osamura, Robert Yoshiyuki

    2008-09-20

    Rupture of adrenal pheochromocytoma is extremely rare and can be lethal because of dramatic changes in the circulation. We describe a 58-year-old Japanese man who suffered rupture of a pheochromocytoma. The patient was referred to our hospital because of severe hypertension (256/127 mmHg) and a left adrenal tumor. T2-weighted magnetic resonance imaging showed high signal intensity in the 50-mm left adrenal tumor. Endocrinological examinations showed elevated plasma and urinary catecholamine levels. These findings suggested that the left adrenal tumor was a pheochromocytoma. Phentolamine mesilate was administered intravenously. This resulted in a decrease of the systolic blood pressure to 100 mmHg. On the third hospital day, the patient complained of left back pain, and abdominal computed tomography showed rupture of the pheochromocytoma. Pulmonary congestion and effusion, and paralytic small-intestinal ileus occurred. Blood pressure was controlled, small-intestinal decompression was done with a Miller-Abbot tube, and body water was controlled by fluid replacement. After the general condition of the patient had became stable, laparoscopic adrenalectomy was performed. Phentolamine mesilate is a useful α-adrenergic blocker. However, care is needed with its administration, because rupture of pheochromocytoma may be related to a decrease in blood pressure induced by phentolamine mesilate.

  10. Globe Rupture

    Directory of Open Access Journals (Sweden)

    Reid Honda

    2017-07-01

    Full Text Available History of present illness: A 46-year-old male presented to the emergency department (ED with severe left eye pain and decreased vision after tripping and striking the left side of his head on the corner of his wooden nightstand. The patient arrived as an inter-facility transfer for a suspected globe rupture with a protective eye covering in place; thus, further physical examination of the eye was not performed by the emergency physician in order to avoid further leakage of aqueous humor. Significant findings: The patient’s computed tomography (CT head demonstrated a deformed left globe, concerning for ruptured globe. The patient had hyperdense material in the posterior segment (see green arrow, consistent with vitreous hemorrhage. CT findings that are consistent with globe rupture may include a collapsed globe, intraocular air, or foreign bodies. Discussion: A globe rupture is a full-thickness defect in the cornea, sclera, or both.1 It is an ophthalmologic emergency. Globe ruptures are almost always secondary to direct perforation via a penetrating mechanism; however, it can occur due to blunt injury if the force generated creates sufficient intraocular pressure to tear the sclera.2 Globes most commonly rupture at the insertions of the intraocular muscles or at the limbus. They are associated with a high rate of concomitant orbital floor fractures.2,3 Possible physical examination findings include a shallow anterior chamber on slit-lamp exam, hyphema, and an irregular “teardrop” pupil. Additionally, a positive Seidel sign, which is performed by instilling fluorescein in the eye and then examining for a dark stream of aqueous humor, is indicative of a globe rupture.4 CT is often used to assess for globe rupture; finds of a foreign body, intraocular air, abnormal contour or volume of the globe, or disruption of the sclera suggest globe rupture.2 The sensitivity of CT scan for diagnosis of globe rupture is only 75%; thus, high clinical

  11. 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.

  12. Integrity Analysis of Damaged Steam Generator Tubes

    International Nuclear Information System (INIS)

    Stanic, D.

    1998-01-01

    Variety of degradation mechanisms affecting steam generator tubes makes steam generators as one of the critical components in the nuclear power plants. Depending of their nature, degradation mechanisms cause different types of damages. It requires performance of extensive integrity analysis in order to access various conditions of crack behavior under operating and accidental conditions. Development and application of advanced eddy current techniques for steam generator examination provide good characterization of found damages. Damage characteristics (shape, orientation and dimensions) may be defined and used for further evaluation of damage influence on tube integrity. In comparison with experimental and analytical methods, numerical methods are also efficient tools for integrity assessment. Application of finite element methods provides relatively simple modeling of different type of damages and simulation of various operating conditions. The stress and strain analysis may be performed for elastic and elasto-plastic state with good ability for visual presentation of results. Furthermore, the fracture mechanics parameters may be calculated. Results obtained by numerical analysis supplemented with experimental results are the base for definition of alternative plugging criteria which may significantly reduce the number of plugged tubes. (author)

  13. Fatigue analysis of a PWR steam generator tube sheet

    International Nuclear Information System (INIS)

    Billon, F.; Buchalet, C.; Poudroux, G.

    1985-01-01

    The fatigue analysis of a PWR steam generator (S.G) tube sheet is threefold. First, the flow, pressure and temperature variations during the design transients are defined for both the primary fluid and the normal and auxiliary feedwater. Second, the flow, velocities, pressure and temperature variations of the secondary fluid at the bottom of the downcomer and above the tube sheet are determined for the transients considered. Finally, the corresponding temperatures and stresses in the tube sheet are calculated and the usage factors determined at various locations in the tube sheet. The currently available standard design transients for the primary fluid and the feedwater are too conservative to be utilized as such in the fatigue analysis of the S.G. tube sheets. Thus, a detailed examination and reappraisal of each operating transient was performed. The revised design conditions are used as inputs to the calculation model TEMPTRON. TEMPTRON determines the mixing conditions between the feedwater and the recirculation fluid from the S.G. feedwater nozzles to the center of the tube sheet via the downcomer. The fluid parameters, flow rate and velocity, temperature and pressure variations, as a function of the time during the transients are obtained. Finally, the usage factors at various locations on the tube sheet are derived using the standard ASME section III method

  14. Visual accumulation tube for size analysis of sands

    Science.gov (United States)

    Colby, B.C.; Christensen, R.P.

    1956-01-01

    The visual-accumulation-tube method was developed primarily for making size analyses of the sand fractions of suspended-sediment and bed-material samples. Because the fundamental property governing the motion of a sediment particle in a fluid is believed to be its fall velocity. the analysis is designed to determine the fall-velocity-frequency distribution of the individual particles of the sample. The analysis is based on a stratified sedimentation system in which the sample is introduced at the top of a transparent settling tube containing distilled water. The procedure involves the direct visual tracing of the height of sediment accumulation in a contracted section at the bottom of the tube. A pen records the height on a moving chart. The method is simple and fast, provides a continuous and permanent record, gives highly reproducible results, and accurately determines the fall-velocity characteristics of the sample. The apparatus, procedure, results, and accuracy of the visual-accumulation-tube method for determining the sedimentation-size distribution of sands are presented in this paper.

  15. A technique to simulate a tube break in a high-pressure gas/cooling water heat exchanger - HTR2008-58161

    International Nuclear Information System (INIS)

    Antwerpen, H. J. V.; Mulder, E. J.

    2008-01-01

    The gas cycles of most High Temperature Gas-Cooled Reactors (HTR's) reject heat to water at some stage. In the helium/water heat exchangers of HTR's with direct Brayton cycles, the helium is usually at a much higher pressure than the water. If the pressure boundary between the helium and the water fails inside the heat exchanger. the effect on the rest of the water system has to be established in order to do a proper system design. This can be done most efficiently by using a system simulation code, however, very few system simulation codes has the capability to do gas/liquid interface tracking as required for this problem. This study describes a calculation method with which a gas/liquid heat exchanger tube rupture can be calculated in a simulation code without interface tracking. The course of events after tube rupture is described and appropriate calculation models derived. A mathematical model for a pressure relief valve (PRV) was also created. The calculation models were implemented in the system simulation software Flownex and used to study a tube rupture on a 5000 kPa helium/water heat exchanger. The assembled calculation network solved stable and within reasonable time. The simulation provided insight into the course of events following the tube break. It was shown that the acceleration of water out of the helium cooler, by choked-flow helium, caused the main pressure pulses during the event. The maximum pressure in the water loop occurs on the opposite side of the helium cooler due to constructive interference of the initial pressure wave with itself. It was also shown that by changing only pipe lengths, the system could become prone to severe oscillations after a tube rupture event. (authors)

  16. Characterize kinematic rupture history of large earthquakes with Multiple Haskell sources

    Science.gov (United States)

    Jia, Z.; Zhan, Z.

    2017-12-01

    Earthquakes are often regarded as continuous rupture along a single fault, but the occurrence of complex large events involving multiple faults and dynamic triggering challenges this view. Such rupture complexities cause difficulties in existing finite fault inversion algorithms, because they rely on specific parameterizations and regularizations to obtain physically meaningful solutions. Furthermore, it is difficult to assess reliability and uncertainty of obtained rupture models. Here we develop a Multi-Haskell Source (MHS) method to estimate rupture process of large earthquakes as a series of sub-events of varying location, timing and directivity. Each sub-event is characterized by a Haskell rupture model with uniform dislocation and constant unilateral rupture velocity. This flexible yet simple source parameterization allows us to constrain first-order rupture complexity of large earthquakes robustly. Additionally, relatively few parameters in the inverse problem yields improved uncertainty analysis based on Markov chain Monte Carlo sampling in a Bayesian framework. Synthetic tests and application of MHS method on real earthquakes show that our method can capture major features of large earthquake rupture process, and provide information for more detailed rupture history analysis.

  17. Evaluation of creep rupture property of high strength ferritic/martensitic steel (PNC-FMS)

    International Nuclear Information System (INIS)

    Uehira, Akihiro; Mizuno, Tomoyasu; Ukai, Shigeharu; Yoshida, Eiichi

    1999-04-01

    High Strength Ferritic/Martensitic Steel (PNC-FMS : 11Cr-0.5Mo-2W,Nb,V), developed by JNC, is one of the candidate materials for the long-life core of large-scale fast breeder reactor. The material design base standard (tentative) of PNC-FMS was established and the creep rupture strength reduction factor in the standard was determined in 1992. This factor was based on only evaluation of decarburization effect on tensile strength after sodium exposure. In this study, creep rupture properties of PNC-FMS under out of pile sodium exposure and in pile were evaluated, using recent test results as well as previous ones. The evaluation results are summarized as follows : a. Decarburization rate constant of pressurized tubes under sodium exposure is identical with stress free specimens. b. In case of the same decarburization content under out of pile sodium exposure, creep strength tends to decrease more significantly than tensile strength. c. Creep strength under out of pile sodium exposure showed significant decrease in high temperature and long exposure time, but in pile (MOTA) creep strength showed little decrease. A new creep rupture strength reduction factor, which is the ratio of creep rupture strength under sodium exposure or in pile to in air, was made by correlating the creep rupture strength. This new method directly using the ratio of creep rupture strength was evaluated and discussed from the viewpoint of design applicability, compared with the conventional method based on decarburization effect on tensile strength. (author)

  18. Structural and leakage integrity assessment of WWER steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K.; Otruba, J. [Nuclear Research Inst., Rez (Switzerland)

    1997-12-31

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

  19. Structural and leakage integrity assessment of WWER steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K; Otruba, J [Nuclear Research Inst., Rez (Switzerland)

    1998-12-31

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

  20. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.

    1997-01-01

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction

  1. Ruptured ectopic pregnancy diagnosed with computed tomography

    International Nuclear Information System (INIS)

    Michalak, Maciej; Żurada, Anna; Biernacki, Maciej; Zygmunt, Kozielec

    2010-01-01

    The rupture of ectopic pregnancy (EP) still remains the primary and direct cause of death in the first trimester of pregnancy. Ultrasonography is known to be a modality of choice in EP diagnostics. We found a severe discrepancy between the frequency of ectopic pregnancies (EP) and the number of available computed tomography (CT) examinations. A 29-year-old woman was admitted to the emergency department with a history of abdominal pain, nausea, vomiting and collapse. Sonographic findings of a suspected EP were unclear. Moreover, not all features of intrauterine pregnancy were present. Due to the patient’s life-threatening condition, an emergency multi-slice CT with MPR and VRT reconstructions was performed, revealing symptoms of a ruptured EP. In the right adnexal area, a well-vascularized, solid-cystic abnormal mass lesion was found. Intraperitoneal hemorrhage was confirmed intraoperatively, and the right fallopian tube with a tubal EP was resected. In the surgery in situ, as well as in the pathological examination of the tumor mass, a human embryo of approximately 1.5 cm in length (beginning of the 8 th week of gestation) was found. Although ultrasonography still remains the first-line imaging examination in EP diagnostics, sometimes the findings of suspected EPs are unclear and not sufficient. The rupture of EP, with serious bleeding and symptoms of shock, may require an emergent pelvic and abdominal CT inspection. A clear correlation was found between the macroscopic CT images and the intraoperatively sampled material

  2. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-01-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  3. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  4. A three-dimensional rupture analysis of steel liners anchored to concrete pressure and containment vessels

    International Nuclear Information System (INIS)

    Bangash, Y.

    1987-01-01

    Steel liners or plates are anchored to concrete pressure and containment vessels for nuclear and offshore facilities. Due to extreme loading conditions a liner may buckle due to the pull-out or shearing of anchors from the base metal and concrete. Under certain conditions attributed to loadings, liner metal deterioration and cracking of concrete behind the liner, the liner may fail by rupture. This paper presents a three-dimensional analysis of steel-concrete elements, using finite elements analysis in which a provision is made for liner instability, anchor strength and stiffness, concrete cracking and finally liner rupture. The analysis is tested first on an octagonal slab with and without an anchored steel liner. It is then extended to concrete pressure and containment vessels. The analytical results obtained are compared well with those available from the experimental tests and other sources. (author)

  5. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L.; Wilam, M. [Vitkovice NPP Services (Switzerland); Herman, M. [Vuje, Trnava (Slovakia)

    1997-12-31

    At operated Czech and Slovak nuclear power plants the 80 % criteria for crack or other bulk defect depth is used for steam generator heat exchanging tubes plugging. This criteria was accepted as the recommendation of designer of WWER steam generators. Verification of this criteria was the objective of experimental program performed by Vitkovice, J.S.C., UJV Rez, J.S.C. and Vuje Trnava, J.S.C .. Within this program the following factors were studied: (1) Influence of secondary water chemistry on defects initiation and propagation, (2) Statistical evaluation of corrosion defects progression at operated SG, and (3) Determination of critical pressure for tube rupture as a function of eddy current indications. In this presentation items (2) and (3) are considered.

  6. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L; Wilam, M [Vitkovice NPP Services (Switzerland); Herman, M [Vuje, Trnava (Slovakia)

    1998-12-31

    At operated Czech and Slovak nuclear power plants the 80 % criteria for crack or other bulk defect depth is used for steam generator heat exchanging tubes plugging. This criteria was accepted as the recommendation of designer of WWER steam generators. Verification of this criteria was the objective of experimental program performed by Vitkovice, J.S.C., UJV Rez, J.S.C. and Vuje Trnava, J.S.C .. Within this program the following factors were studied: (1) Influence of secondary water chemistry on defects initiation and propagation, (2) Statistical evaluation of corrosion defects progression at operated SG, and (3) Determination of critical pressure for tube rupture as a function of eddy current indications. In this presentation items (2) and (3) are considered.

  7. Heat transfer and thermal stress analysis in grooved tubes

    Indian Academy of Sciences (India)

    Heat transfer and thermal stresses, induced by temperature differencesin the internally grooved tubes of heat transfer equipment, have been analysed numerically. The analysis has been conducted for four different kinds of internally grooved tubes and three different mean inlet water velocities. Constant temperature was ...

  8. Status of prototype rupture disc testing in the large leak test rig

    International Nuclear Information System (INIS)

    Amos, J.C.

    1979-09-01

    The prototype CRBRP double membrane rupture disc assembly is being performance tested in conjunction with the LLTR Series II Large Leak Program. In May 1979, the double membrane disc assembly was inadvertently activated during sodium system pressure instrument calibration. This experience indicated that the rupture disc burst at essentially the design burst pressure when a gradually increasing state pressure was applied. The area of membrane opening was found to be about 25 to 30% of the cross-sectional area. In July 1979, the disc assembly was again tested (this time in a single membrane configuration) in conjunction with the first LLTR Series II Test A-1a (inert gas injection). Test data indicated that the disc burst in about 35 ms at essentially the design burst pressure with an opening of about 30% of the cross-sectional area. The pressure immediately downstream of the disc dropped below atmospheric pressure following the rupture tube event (releasing high pressure nitrogen into sodium) for about 1.5 seconds before increasing to a maximum of 30 psig. This behavior raises a question on the adequacy of a downstream pressure device for rapid sensing of disc rupture and initiating plant shutdown following a large SWR event. 14 figures

  9. Reflood Heat Transfer in SiC and Graphene Oxide Coated Tube

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Lee, Seung Won; Bang, In Cheol

    2013-01-01

    The reflood tests have been performed flowing water into bare tube and nanoparticles coated tube at constant flow rate (3 cm/s). The quenching curves have been obtained at atmospheric pressure. Finally, Scanning Electron Microscopy (SEM) images are acquired and contact angles are measured in order to observe the surface structures and wettability effect on cooling performance. The quenching time decreases and quenching velocity increases as the coating time of nanoparticles on the tube increases, because the nanoparticles deposited on the tube destabilize and rupture the vapor film early in the effect of increased Leidenfrost point temperature. The SiC nanoparticles coated tubes have better quenching performance than GO nanoparticles coated tubes. The SEM images and contact angle observations proved the enhanced wettability and rough surface due to deposition of SiC nanoparticles. And the wettability of GO nanoparticles coated tubes shows the increase at 600 s coating. But, the wettability decreases on GO nanoparticles tube coated for 900 s despite the enhanced quenching performance. Thus, the porous structure affects to the better cooling performance in case of GO nanoparticles coated tubes

  10. Crush analysis of the foam-filled bitubal circular tube under oblique impact

    Science.gov (United States)

    Djamaluddin, F.; Abdullah, S.; Arrifin, A. K.; Nopiah, Z. M.

    2018-02-01

    This paper presents crashworthiness analysis of bitubal cylindrical tubes under different impact angular. The numerical solution of double cylindrical tubes are determined by finite element analysis (FEA). Moreover, the structure was impacted by mass block as impactor respect to longitudinal direction of the tubes. The model of structure was developed by non-linear ABAQUS sofware with variations of load angle and dimensions of tube. The outcome of this study is the respons parameters such as the peak crusing force (PCF), energy absorption (EA) and specific energy absorption (SEA), thus it can be expected this tube as the great energy absorber.

  11. Rupture disc

    International Nuclear Information System (INIS)

    Newton, R.G.

    1977-01-01

    The intermediate heat transport system for a sodium-cooled fast breeder reactor includes a device for rapidly draining the sodium therefrom should a sodium-water reaction occur within the system. This device includes a rupturable member in a drain line in the system and means for cutting a large opening therein and for positively removing the sheared-out portion from the opening cut in the rupturable member. According to the preferred embodiment of the invention the rupturable member includes a solid head seated in the end of the drain line having a rim extending peripherally therearound, the rim being clamped against the end of the drain line by a clamp ring having an interior shearing edge, the bottom of the rupturable member being convex and extending into the drain line. Means are provided to draw the rupturable member away from the drain line against the shearing edge to clear the drain line for outflow of sodium therethrough

  12. Augmented Versus Nonaugmented Repair of Acute Achilles Tendon Rupture: A Systematic Review and Meta-analysis.

    Science.gov (United States)

    Zhang, Yi-Jun; Zhang, Chi; Wang, Quan; Lin, Xiang-Jin

    2017-04-01

    Although simple end-to-end repair of the Achilles tendon is common, many augmented repair protocols have been implemented for acute Achilles tendon rupture. However, whether augmented repair is better than nonaugmented repair of an acute Achilles tendon rupture is still unknown. To conduct a meta-analysis to determine whether augmented surgical repair of an acute Achilles tendon rupture improved subjective patient satisfaction without an increase in rerupture rates. Secondary outcomes assessed included infections, ankle range of motion, calf muscle strength, and minor complications. Meta-analysis. A systematic literature search of peer-reviewed articles was conducted to identify all randomized controlled trials (RCTs) comparing augmented repair and nonaugmented repair for acute Achilles tendon rupture from January 1980 to August 2016 in the electronic databases of PubMed, Web of Science (SCI-E/SSCI/A&HCI), and EMBASE. The keywords (Achilles tendon rupture) AND (surg* OR operat* OR repair* OR augment* OR non-augment* OR end-to-end OR sutur*) were combined, and results were limited to human RCTs and controlled clinical trials published in the English language. Four RCTs involving 169 participants were eligible for inclusion; 83 participants were treated with augmented repair and 86 were treated with nonaugmented repair. Augmented repair led to similar responses when compared with nonaugmented repair for acute Achilles tendon rupture (93% vs 90%, respectively; P = .53). The rerupture rates showed no significant difference for augmented versus nonaugmented repair (7.2% vs 9.3%, respectively; P = .69). No differences in superficial and deep infections occurred in augmented (7 infections) and nonaugmented (8 infections) repair groups during postoperative follow-up ( P = .89). The average incisional infection rate was 8.4% with augmented repair and 9.3% with nonaugmented repair. No significant differences in other complications were found between augmented (7.2%) and

  13. Temperature and thermal stress analysis of a switching tube anode

    International Nuclear Information System (INIS)

    Sutton, S.B.

    1979-01-01

    In the design of high power density switching tubes which are subjected to cyclic thermal loads, the temperature induced stresses must be minimized in order to maximize the life expectancy of the tube. Following are details of an analysis performed for the Magnetic Fusion Program at the Lawrence Livermore Laboratory on a proposed tube. The tube configuration is given. The problem was simplified to one-dimensional approximations for both the thermal and stress analyses. The underlying assumptions and their implications are discussed

  14. Analysis methods for evaluating leak-before-break in U-tube steam generators

    International Nuclear Information System (INIS)

    Griesbach, T.; Cipolla, R.

    1985-01-01

    In recent years, there has been an increased incidence of cracking in steam generator tubes. As a result, there has been increased effort in assuring that cracks in steam generator tubes will leak well in advance of significant loss in structural integrity. Demonstrating a leak-before-break condition is an integrated analysis process that utilizes several engineering disciplines, specifically, materials engineering, fracture mechanics, stress analysis, and fluid mechanics. The output from a leak-before-break assessment is typically depicted in terms of available margins against failure and measurable or detectable leak rate. In this paper, the analysis methods for performing a leak-before-break analysis for the U-tubes of a recirculating steam generator are presented. The results from generic analysis for the first row U-tubes illustrates the analysis techniques. Because of realistic input values used herein, these results also suggest that large leak rates are possible from cracks in U-bend regions, yet these cracks are small relative to their critical size for failure. Hence, orderly shutdowns can be completed prior to the point when tube bursting is of concern

  15. A study of the effect of maintenance on the safety of a mechanical system subject to aging and its application to steam generator tube degradation

    International Nuclear Information System (INIS)

    Dussarte, D.

    1991-11-01

    The different degradation mechanisms to which pressurized water reactor steam generator tubes are observed to be subject may result in the risk of their rupture being greater than anticipated. Prevention of tube rupture essentially consists of inspections during outages of the units and applying appropriate criteria for the withdrawal of defective tubes from service. Planning such measures implies being able to gauge the effectiveness of the action taken. This document describes a proposed technique for quantifying the effects of the preventive maintenance we have had to develop to address this problem and, hence, to obtain material for assessing the action taken by the utility. (author)

  16. Contrastive Analysis and Research on Negative Pressure Beam Tube System and Positive Pressure Beam Tube System for Mine Use

    Science.gov (United States)

    Wang, Xinyi; Shen, Jialong; Liu, Xinbo

    2018-01-01

    Against the technical defects of universally applicable beam tube monitoring system at present, such as air suction in the beam tube, line clogging, long sampling time, etc., the paper analyzes the current situation of the spontaneous combustion fire disaster forecast of mine in our country and these defects one by one. On this basis, the paper proposes a research thought that improving the positive pressure beam tube so as to substitute the negative pressure beam tube. Then, the paper introduces the beam tube monitoring system based on positive pressure technology through theoretical analysis and experiment. In the comparison with negative pressure beam tube, the paper concludes the advantage of the new system and draws the conclusion that the positive pressure beam tube is superior to the negative pressure beam tube system both in test result and test time. At last, the paper proposes prospect of the beam tube monitoring system based on positive pressure technology.

  17. Analysis of the FFTF primary pipe rupture transients

    International Nuclear Information System (INIS)

    Perkins, K.R.; Bari, R.A.; Chen, L.C.; Albright, D.C.

    1979-01-01

    The response of the Fast Flux Test Facility (FFTF) to hypothetical ruptures of the high pressure primary piping has been analyzed using two LMFBR plant systems codes, namely IANUS and DEMO. Comparisons of the average channel temperatures predicted by the two codes show good agreement for identical transients. However, the hot channel temperatures predicted by DEMO are about 60K higher than the corresponding IANUS predictions for severe transients. This difference is attributed to the dynamic hot channel factors employed in DEMO which discount the thermal inertia of the duct walls for rapid transients. DEMO also predicts more severe transients for hot-leg ruptures in FFTF than previously reported analyses for the CRBR

  18. Structural analysis of 177-FA redesigned surveillance specimen holder tube

    International Nuclear Information System (INIS)

    Pryor, C.W.; Thoren, D.E.; Vames, G.J.; Harris, R.J.

    1976-08-01

    Because of in-service operational problems, the surveillance specimen holder tubes described in B and W topical report BAW-10051 have been redesigned. This report describes the new design and structural analysis for normal operation and upset loading conditions. The results of the analysis demonstrate the adequacy of the new surveillance specimen holder tubes for their design life of 40 years

  19. Studies of the steam generator degraded tubes behavior on BRUTUS test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chedeau, C.; Rassineux, B. [EDF/DER/MTC, Moret Sur Loing (France); Flesch, B. [EDF/EPN/DMAINT, Paris (France)] [and others

    1997-04-01

    Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture.

  20. Fracture analysis of HFIR beam tube caused by radiation embrittlement

    International Nuclear Information System (INIS)

    Chang, S.J.

    1994-01-01

    With an attempt to estimate the neutron beam tube embrittlement condition for the Oak Ridge High Flux Isotope Reactor (HFIR), fracture mechanics calculations are carried out in this paper. The analysis provides some numerical result on how the tube has been structurally weakened. In this calculation, a lateral impact force is assumed. Numerical result is obtained on how much the critical crack size should be reduced if the beam tube has been subjected to an extended period of irradiation. It is also calculated that buckling strength of the tube is increased, not decreased, with irradiation

  1. Inelastic analysis of finite length and depth cracked tubes

    International Nuclear Information System (INIS)

    Reich, M.; Gardner, D.; Prachuktam, S.; Chang, T.Y.

    1977-01-01

    Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdown. This paper concerns itself with the prediction of the failure pressures for typical PWR steam generator tubes with longitudinal finite length and finite depth cracks. Only local plastic overload failure is considered since the material is non-notch sensitive. Non-linear finite element analyses are carried out to determine the burst pressures of steam generator tubes containing longitudinal cracks located on the outer surface of the tubes. The non-linearities considered herein include elastic-plastic material behaviour and large deformations. A non-proprietary general purpose non-linear finite element program, NFAP was adopted for the analysis. Due to the asymmetric nature of the cracks, two-dimensional as well as three-dimensional finite element analyses, were performed. The analysis clearly shows that for short cracks axial effects play a significant role. For long cracks, they are not important since two-dimensional conditions predominate and failure is governed by circumferential or hoop stress conditions. (Auth.)

  2. Analysis of tube vibrations in D-4 steam generator

    International Nuclear Information System (INIS)

    Mavko, B.; Peterlin, G.; Boltezar, M.

    1983-01-01

    Accelerometer data for the most exposed tube in steam generator D-4 were recorded on magnetic tape. Procedures for calculations of the most characteristic parameters were prepared for spectral analyzer on SD 360. Parameters which most satisfactorily describe the vibrations are power spectral densities peak to peak acceleration volume and root mean square displacement. Computer program was written to calculate the natural frequencies of a multispaned tube. Procedures and the computer program will be used for independent analysis of tube vibrations in Krsko D-4 type steam generator. (author)

  3. Analysis and computer program for rupture-risk prediction of abdominal aortic aneurysms

    Directory of Open Access Journals (Sweden)

    Li Zhonghua

    2006-03-01

    Full Text Available Abstract Background Ruptured abdominal aortic aneurysms (AAAs are the 13th leading cause of death in the United States. While AAA rupture may occur without significant warning, its risk assessment is generally based on critical values of the maximum AAA diameter (>5 cm and AAA-growth rate (>0.5 cm/year. These criteria may be insufficient for reliable AAA-rupture risk assessment especially when predicting possible rupture of smaller AAAs. Methods Based on clinical evidence, eight biomechanical factors with associated weighting coefficients were determined and summed up in terms of a dimensionless, time-dependent severity parameter, SP(t. The most important factor is the maximum wall stress for which a semi-empirical correlation has been developed. Results The patient-specific SP(t indicates the risk level of AAA rupture and provides a threshold value when surgical intervention becomes necessary. The severity parameter was validated with four clinical cases and its application is demonstrated for two AAA cases. Conclusion As part of computational AAA-risk assessment and medical management, a patient-specific severity parameter 0

  4. Tocolysis after preterm premature rupture of membranes and neonatal outcome: a propensity-score analysis.

    Science.gov (United States)

    Lorthe, Elsa; Goffinet, François; Marret, Stéphane; Vayssiere, Christophe; Flamant, Cyril; Quere, Mathilde; Benhammou, Valérie; Ancel, Pierre-Yves; Kayem, Gilles

    2017-08-01

    There are conflicting results regarding tocolysis in cases of preterm premature rupture of membranes. Delaying delivery may reduce neonatal morbidity because of prematurity and allow for prenatal corticosteroids and, if necessary, in utero transfer. However, that may increase the risks of maternofetal infection and its adverse consequences. The objective of the study was to investigate whether tocolytic therapy in cases of preterm premature rupture of membranes is associated with improved neonatal or obstetric outcomes. Etude Epidémiologique sur les Petits Ages Gestationnels 2 is a French national prospective, population-based cohort study of preterm births that occurred in 546 maternity units in 2011. Inclusion criteria in this analysis were women with preterm premature rupture of membranes at 24-32 weeks' gestation and singleton gestations. Outcomes were survival to discharge without severe morbidity, latency prolonged by ≥48 hours and histological chorioamnionitis. Uterine contractions at admission, individual and obstetric characteristics, and neonatal outcomes were compared by tocolytic treatment or not. Propensity scores and inverse probability of treatment weighting for each woman were used to minimize indication bias in estimating the association of tocolytic therapy with outcomes. The study population consisted of 803 women; 596 (73.4%) received tocolysis. Women with and without tocolysis did not differ in neonatal survival without severe morbidity (86.7% vs 83.9%, P = .39), latency prolonged by ≥48 hours (75.1% vs 77.4%, P = .59), or histological chorioamnionitis (50.0% vs 47.6%, P = .73). After applying propensity scores and assigning inverse probability of treatment weighting, tocolysis was not associated with improved survival without severe morbidity as compared with no tocolysis (odds ratio, 1.01 [95% confidence interval, 0.94-1.09], latency prolonged by ≥48 hours (1.03 [95% confidence interval, 0.95-1.11]), or histological chorioamnionitis

  5. Ruptured eardrum

    Science.gov (United States)

    ... eardrum ruptures. After the rupture, you may have: Drainage from the ear (drainage may be clear, pus, or bloody) Ear noise/ ... doctor to see the eardrum. Audiology testing can measure how much hearing has been lost. Treatment You ...

  6. Transient analysis for resolving safety issues

    International Nuclear Information System (INIS)

    Chao, J.; Layman, W.

    1987-01-01

    The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

  7. Analysis of the heat transfer in double and triple concentric tube heat exchangers

    Science.gov (United States)

    Rădulescu, S.; Negoiţă, L. I.; Onuţu, I.

    2016-08-01

    The tubular heat exchangers (shell and tube heat exchangers and concentric tube heat exchangers) represent an important category of equipment in the petroleum refineries and are used for heating, pre-heating, cooling, condensation and evaporation purposes. The paper presents results of analysis of the heat transfer to cool a petroleum product in two types of concentric tube heat exchangers: double and triple concentric tube heat exchangers. The cooling agent is water. The triple concentric tube heat exchanger is a modified constructive version of double concentric tube heat exchanger by adding an intermediate tube. This intermediate tube improves the heat transfer by increasing the heat area per unit length. The analysis of the heat transfer is made using experimental data obtained during the tests in a double and triple concentric tube heat exchanger. The flow rates of fluids, inlet and outlet temperatures of water and petroleum product are used in determining the performance of both heat exchangers. Principally, for both apparatus are calculated the overall heat transfer coefficients and the heat exchange surfaces. The presented results shows that triple concentric tube heat exchangers provide better heat transfer efficiencies compared to the double concentric tube heat exchangers.

  8. Methods to diagnose acute anterior cruciate ligament rupture: a meta-analysis of instrumented knee laxity tests

    NARCIS (Netherlands)

    van Eck, Carola F.; Loopik, Miette; van den Bekerom, Michel P.; Fu, Freddie H.; Kerkhoffs, Gino M. M. J.

    2013-01-01

    The aims of this meta-analysis were to determine the sensitivity and specificity of the KT 1000 Arthrometer, Stryker Knee Laxity Tester and Genucom Knee Analysis System for ACL rupture. It was hypothesized that the KT 1000 test is the most sensitive and specific. Secondly, it was hypothesized that

  9. Analysis of prestressed double-wall tubing for LMFBR steam generators

    International Nuclear Information System (INIS)

    Uber, C.F.; Langford, P.J.

    1981-01-01

    A radial interface pressure is provided between the inner and outer tubes of each double-wall tube in a steam generator design now being developed for commercial breeder reactor plants. This paper describes a finite element analysis of the manufacturing technique used to prestress the double-wall tube. The analytical predictions are compared with experimental measurements of the residual interface pressure. Resulting residual stress states are used as the starting point for operating condition analyses. 9 refs

  10. Analysis of temperature and stress distribution of superheater tubes after attemperation or sootblower activation

    International Nuclear Information System (INIS)

    Madejski, Paweł; Taler, Dawid

    2013-01-01

    Highlights: • The CFD simulation was used to calculate 3D steam and tube wall temperature distributions in the platen superheater. • The CFD results can be used in design of superheaters made of tubes with complex cross-section. • The CFD analysis enables the proper selection of the steel grade. • The transient temperature and stress distributions were calculated using Finite Volume Method. • The detailed analysis prevents superheater tubes from excessive stresses during sootblower or attemperator activation. - Abstract: Superheaters are characterized by high metal temperatures due to higher steam temperature and low heat transfer coefficients on the tube inner surfaces. Superheaters have especially difficult operating conditions, particularly during attemperator and sootblower activations, when temperature and steam flow rate as well as tube wall temperature change with time. A detailed thermo-mechanical analysis of the superheater tubes makes it possible to identify the cause of premature high-temperature failures and aids greatly in the changes in tubing arrangement and improving start-up technology. This paper presents a thermal and strength analysis of a tube “double omega”, used in the steam superheaters in CFB boilers

  11. Testing and analysis of tube voltage and tube current in the radiation generator for mammography

    International Nuclear Information System (INIS)

    Jung, Hong Ryang; Hong, Dong Hee; Han, Beom Hui

    2014-01-01

    Breast shooting performance management and quality control of the generator is applied to the amount of current IEC(International Electrotechnical Commission) 60601-2-45 tube voltage and tube current are based on standards that were proposed in the analysis of the test results were as follows. Tube voltage according to the value of the standard deviation by year of manufacture from 2001 to 2010 as a 42-3.15 showed the most significant, according to the year of manufacture by tube amperage value of the standard deviation to 6.38 in the pre-2000 showed the most significant , manufactured after 2011 the standard deviation of the devices, the PAE(Percent Average Error) was relatively low. This latest generation device was manufactured in the breast of the tube voltage and tube diagnosed shooting the correct amount of current to maintain the performance that can be seen. The results of this study as the basis for radiography diagnosed breast caused by using the device's performance and maintain quality control, so the current Food and Drug Administration 'about the safety of diagnostic radiation generator rule' specified in the test cycle during three years of self-inspection radiation on a radiation generating device ensure safety and performance of the device using a coherent X-ray(constancy) by two ultimately able to keep the radiation dose to the public to reduce the expected effect is expected

  12. Development of rupture process analysis method for great earthquakes using Direct Solution Method

    Science.gov (United States)

    Yoshimoto, M.; Yamanaka, Y.; Takeuchi, N.

    2010-12-01

    Conventional rupture process analysis methods using teleseismic body waves were based on ray theory. Therefore, these methods have the following problems in applying to great earthquakes such as 2004 Sumatra earthquake: (1) difficulty in computing all later phases such as the PP reflection phase, (2) impossibility of computing called “W phase”, the long period phase arriving before S wave, (3) implausibility of hypothesis that the distance is far enough from the observation points to the hypocenter compared to the fault length. To solve above mentioned problems, we have developed a new method which uses the synthetic seismograms computed by the Direct Solution Method (DSM, e.g. Kawai et al. 2006) as Green’s functions. We used the DSM software (http://www.eri.u-tokyo.ac.jp/takeuchi/software/) for computing the Green’s functions up to 1 Hz for the IASP91 (Kennett and Engdahl, 1991) model, and determined the final slip distributions using the waveform inversion method (Kikuchi et al. 2003). First we confirmed whether the Green’s functions computed by DSM were accurate in higher frequencies up to 1 Hz. Next we performed the rupture process analysis of this new method for Mw8.0 (GCMT) large Solomon Islands earthquake on April 1, 2007. We found that this earthquake consisted of two asperities and the rupture propagated across the subducting Sinbo ridge. The obtained slip distribution better correlates to the aftershock distributions than existing method. Furthermore, this new method keep same accuracy of existing method (which has the advantage of calculating) with respect to direct P-wave and reflection phases near the source, and also accurately calculate the later phases such a PP-wave.

  13. Ruptured Heterotopic Tubal Pregnancy for a Patient with a History of Segmental Salpingectomy from Ectopic Pregnancy: A Case Report

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Kyung Bum; Namkung, Sook; Hong, Myung Sun; Kim, Heung Cheol; Cho, Young; Choi, Young Hee [Chuncheon Sacred Heart Hospital, Chyncheon (Korea, Republic of)

    2012-06-15

    Heterotopic pregnancy refers to the simultaneous development of an intrauterine pregnancy and an extrauterine pregnancy. We experienced a case of a ruptured heterotopic pregnancy for a patient with a history of a right segmental salpingectomy from an ectopic pregnancy. The 30-year-old patient with amenorrhea for six weeks complained of lower abdominal pain with hypovolemic shock. Transabdominal ultrasonography showed diffuse hemoperitoneum with a structure similar to an ectatic tube or a deformed cyst with no echogenic double ring or peripheral hypervascularity in the right adnexa and an intrauterine gestational sac. We considered a ruptured corpus luteum cyst as an ultrasonographic finding and found a ruptured tubal mass in the right salpinx and hemoperitoneum through an emergency laparotomy. We performed a right salpingectomy, and the histopathologic report confirmed ectopic pregnancy.

  14. Ruptured Heterotopic Tubal Pregnancy for a Patient with a History of Segmental Salpingectomy from Ectopic Pregnancy: A Case Report

    International Nuclear Information System (INIS)

    Nam, Kyung Bum; Namkung, Sook; Hong, Myung Sun; Kim, Heung Cheol; Cho, Young; Choi, Young Hee

    2012-01-01

    Heterotopic pregnancy refers to the simultaneous development of an intrauterine pregnancy and an extrauterine pregnancy. We experienced a case of a ruptured heterotopic pregnancy for a patient with a history of a right segmental salpingectomy from an ectopic pregnancy. The 30-year-old patient with amenorrhea for six weeks complained of lower abdominal pain with hypovolemic shock. Transabdominal ultrasonography showed diffuse hemoperitoneum with a structure similar to an ectatic tube or a deformed cyst with no echogenic double ring or peripheral hypervascularity in the right adnexa and an intrauterine gestational sac. We considered a ruptured corpus luteum cyst as an ultrasonographic finding and found a ruptured tubal mass in the right salpinx and hemoperitoneum through an emergency laparotomy. We performed a right salpingectomy, and the histopathologic report confirmed ectopic pregnancy.

  15. True posterior communicating artery aneurysms: are they more prone to rupture? A biomorphometric analysis.

    Science.gov (United States)

    He, Wenzhuan; Hauptman, Jason; Pasupuleti, Latha; Setton, Avi; Farrow, Maria G; Kasper, Lydia; Karimi, Reza; Gandhi, Chirag D; Catrambone, Jeffrey E; Prestigiacomo, Charles J

    2010-03-01

    Posterior communicating artery (PCoA) aneurysms can occur at the junction with the internal carotid artery, posterior cerebral artery (PCA), or the proximal PCoA itself. Hemodynamic stressors contribute to aneurysm formation and may be associated with parent vessel size and aneurysm location. This study evaluates the correlation of various biomorphometric characteristics in 2 of the aforementioned types of PCoA aneurysms. Patients with PCoA aneurysms were analyzed using CT angiography. Source images and reconstructions were used to determine which aneurysms originated purely from the PCoA and those that originated from the internal carotid artery/PCoA junction. Morphometric analysis was performed on the aneurysm, the precommunicating segment of the PCA (P(1)), the ambient segment of the PCA (P(2)), and both PCoA arteries and were correlated to clinical presentation. Parametric and nonparametric analyses were performed to test for significance. A total of 77 PCoA aneurysms were analyzed, and 10 were found to be true PCoA aneurysms (13.0%). The ipsilateral PCoA/P(1) ratio (1.77 +/- 0.44 vs 0.82 +/- 0.46, p = 0.0001) and ipsilateral P(2)/P(1) ratio (1.73 +/- 0.40 vs 1.22 +/- 0.41, p = 0.0003) were significantly larger in true PCoA aneurysms. Interestingly, aneurysm size was statistically larger in the junctional aneurysms (0.14 +/- 0.1 vs 0.072 +/- 0.04 cm(3), p = 0.03). The prevalence of ruptured aneurysms was similar in both groups (approximately 80%, p value not significant). These data suggest that true PCoA aneurysms have a larger PCoA relative to the ipsilateral P(1) segment. To the authors' knowledge, this represents the first such biomorphometric comparison of these different types of PCoA aneurysms. Although statistically smaller in size, true PCoA aneurysms also have a similar prevalence of presenting as a ruptured aneurysm, suggesting that they might be more prone to rupture than a junctional aneurysms of similar size. Further analysis will be required to

  16. Creep analysis of boiler tubes by fem | Taye | Zede Journal

    African Journals Online (AJOL)

    In this paper an analysis is developed for the determination of creep deformation of an axisymmetric boiler tubes subjected to axisymmetric loads. The stresses and the permanent strains at a particular time and at the steady state condition, resulting from loading of the tube under constant internal pressure and elevated ...

  17. Reliability assessment of creep rupture life for Gr. 91 steel

    International Nuclear Information System (INIS)

    Kim, Woo-Gon; Park, Jae-Young; Kim, Seon-Jin; Jang, Jinsung

    2013-01-01

    Highlights: • Statistical analysis of a number of creep rupture data based on Z parameter. • Determination of the constant C in LM parameter and long-term creep life prediction. • Generation of random variables for Z s and Z cr by Monte-Carlo simulation in a SCRI model. • Examples for design application were reasonably drawn from the viewpoints of reliability. - Abstract: This paper presents reliability assessment of the long-term creep life of Gr. 91 steel, which is a major structural material for high temperature structural components of Generation-IV reactor systems. A number of creep rupture data for Gr. 91 steel were collected through literature surveys, and the long-term creep life was predicted by Larson–Miller parameter. A “Z parameter” method was used to describe the magnitude of the deviation of the creep rupture data to a master curve. A “Service Condition-creep Rupture property Interference (SCRI) model” based on the Z parameter was used to simultaneously consider the scattering of the creep rupture data of materials and the fluctuations of service conditions in reliability assessment. A statistical analysis of the creep rupture data was conducted by the Z parameter. To carry out the SCRI model, a number of random variables for Z s describing service conditions and Z cr describing the dispersion of the creep rupture data were generated using a Monte-Carlo simulation technique. As examples for application, the creep rupture life under a certain service conditions of Gr. 91 steel was reasonably drawn from the viewpoints of reliability

  18. Slow rupture of frictional interfaces

    Science.gov (United States)

    Bar Sinai, Yohai; Brener, Efim A.; Bouchbinder, Eran

    2012-02-01

    The failure of frictional interfaces and the spatiotemporal structures that accompany it are central to a wide range of geophysical, physical and engineering systems. Recent geophysical and laboratory observations indicated that interfacial failure can be mediated by slow slip rupture phenomena which are distinct from ordinary, earthquake-like, fast rupture. These discoveries have influenced the way we think about frictional motion, yet the nature and properties of slow rupture are not completely understood. We show that slow rupture is an intrinsic and robust property of simple non-monotonic rate-and-state friction laws. It is associated with a new velocity scale cmin, determined by the friction law, below which steady state rupture cannot propagate. We further show that rupture can occur in a continuum of states, spanning a wide range of velocities from cmin to elastic wave-speeds, and predict different properties for slow rupture and ordinary fast rupture. Our results are qualitatively consistent with recent high-resolution laboratory experiments and may provide a theoretical framework for understanding slow rupture phenomena along frictional interfaces.

  19. Stress relaxation analysis and irradiation creep and swelling in pressure tubes

    International Nuclear Information System (INIS)

    Beeston, J.M.; Burr, T.K.

    1979-01-01

    An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

  20. Examination of the SG tube fatigue cracking at Fessenheim unit no.2 of EDF

    International Nuclear Information System (INIS)

    Boccanfuso, M.; Lorthios, J.; Thebault, Y.; Bruyere, B.; Duisabeau, L.; Herms, E.

    2015-01-01

    In February 2008, a primary-to-secondary leak occurred at Fessenheim Unit No.2 on a steam generator. A circumferential fatigue crack was observed at the upper tube support plate level of the R12C62 tube although the stability ratio evaluation performed to take into account some prior international events, concluded that this tube had no risk of fluid-elastic instability. A new tube pull process was developed and performed by AREVA in 2011 just before the SG replacement. The extraction at the uppermost TSP elevation was a first occurrence in the French EDF PWR. Destructive examinations were carried out in the EDF hot laboratory of CEIDRE/Chinon in order to characterize damage mechanisms at the initiation and propagation stage. The document relates the major results of laboratory examinations leading us to exclude the fluid-elastic instability scenario as previously reported in North-Anna (1987) and Mihama (1991) tube rupture incidents. Results analysis with particular focus on the fracture surface description using Scanning Electron microscopy observations and metallurgical investigations provide new elements concerning the aggravating factors of fatigue damage. Fracture surface investigations reveal that the circumferential crack was due to high cycle fatigue with a very low stress intensity factor. Some aggravating factors like intergranular corrosion appeared to be critical for the fatigue cracking initiation stage. The deterioration was also largely promoted by the lack of tube support at the Anti-Vibration Bars

  1. Hydrogen tube vehicle for supersonic transport: Analysis of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Miller, A.R. [Vehicle Projects LLC and Supersonic Tube Vehicle LLC, 621 17th Street, Suite 2131, Denver, CO 80293 (United States)

    2008-04-15

    I propose and analyze a concept vehicle that operates in a hydrogen atmosphere contained within a tube, or pipeline, and because of the high speed of sound in hydrogen, it delays the onset of the sound barrier. Mach 1.2 in air corresponds to only Mach 0.32 in hydrogen. The proposed vehicle, a cross between a train and an airplane, is multi-articulated, runs on a guideway, is propelled by propfans, and flies on a hydrogen aerostatic fluid film. Vehicle power is provided by onboard hydrogen-oxygen fuel cells. Hydrogen fuel is taken from the tube itself, liquid oxygen (LOX) is carried onboard, and the product water is collected and stored until the end of a run. Thus, unlike conventional vehicles, it breathes its fuel, stores its oxidant, and its weight increases during operation. Taking hydrogen fuel from the tube solves the problem of vehicular hydrogen storage, a major challenge of contemporary hydrogen fuel-cell vehicles. The foundation of the feasibility analysis is extrapolation of aerodynamic properties of a mid-sized turboprop airliner, the Bombardier Dash 8 Q400 trademark. Based on the aerodynamic analysis, I estimate that the hydrogen tube vehicle would require 2.0 MW of power to run at 1500 km/h, which is supersonic with respect to air. It would require 2.64 h to travel from New York City to Los Angeles, consuming 2330 L of onboard LOX and producing 2990 L of liquid water during the trip. Part of the feasibility analysis shows that it is possible to package the corresponding fuel-cell stacks, LOX systems, and water holding tanks in the tube vehicle. The greatest technical challenge is levitation by aerostatic hydrogen bearings. Risk of fire or detonation within the tube, similar to that of existing large natural-gas pipelines, is expected to be manageable and acceptable. (author)

  2. Analysis of severe core damage accident progression for the heavy water reactor

    International Nuclear Information System (INIS)

    Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

    2010-01-01

    In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

  3. Flow distribution analysis on the cooling tube network of ITER thermal shield

    International Nuclear Information System (INIS)

    Nam, Kwanwoo; Chung, Wooho; Noh, Chang Hyun; Kang, Dong Kwon; Kang, Kyoung-O; Ahn, Hee Jae; Lee, Hyeon Gon

    2014-01-01

    Thermal shield (TS) is to be installed between the vacuum vessel or the cryostat and the magnets in ITER tokamak to reduce the thermal radiation load to the magnets operating at 4.2K. The TS is cooled by pressurized helium gas at the inlet temperature of 80K. The cooling tube is welded on the TS panel surface and the composed flow network of the TS cooling tubes is complex. The flow rate in each panel should be matched to the thermal design value for effective radiation shielding. This paper presents one dimensional analysis on the flow distribution of cooling tube network for the ITER TS. The hydraulic cooling tube network is modeled by an electrical analogy. Only the cooling tube on the TS surface and its connecting pipe from the manifold are considered in the analysis model. Considering the frictional factor and the local loss in the cooling tube, the hydraulic resistance is expressed as a linear function with respect to mass flow rate. Sub-circuits in the TS are analyzed separately because each circuit is controlled by its own control valve independently. It is found that flow rates in some panels are insufficient compared with the design values. In order to improve the flow distribution, two kinds of design modifications are proposed. The first one is to connect the tubes of the adjacent panels. This will increase the resistance of the tube on the panel where the flow rate is excessive. The other design suggestion is that an orifice is installed at the exit of tube routing where the flow rate is to be reduced. The analysis for the design suggestions shows that the flow mal-distribution is improved significantly

  4. Morphological characteristics associated with rupture risk of multiple intracranial aneurysms.

    Science.gov (United States)

    Wang, Guang-Xian; Liu, Lan-Lan; Wen, Li; Cao, Yun-Xing; Pei, Yu-Chun; Zhang, Dong

    2017-10-01

    To identify the morphological parameters that are related to intracranial aneurysms (IAs) rupture using a case-control model. A total of 107 patients with multiple IAs and aneurysmal subarachnoid hemorrhage between August 2011 and February 2017 were enrolled in this study. Characteristics of IAs location, shape, neck width, perpendicular height, depth, maximum size, flow angle, parent vessel diameter (PVD), aspect ratio (AR) and size ratio (SR) were evaluated using CT angiography. Multiple logistic regression analysis was used to identify the independent risk factors associated with IAs rupture. Receiver operating characteristic curve analysis was performed on the final model, and the optimal thresholds were obtained. IAs located in the internal carotid artery (ICA) was associated with a negative risk of rupture, whereas AR, SR1 (height/PVD) and SR2 (depth/PVD) were associated with increased risk of rupture. When SR was calculated differently, the odds ratio values of these factors were also different. The receiver operating characteristic curve showed that AR, SR1 and SR2 had cut-off values of 1.01, 1.48 and 1.40, respectively. SR3 (maximum size/PVD) was not associated with IAs rupture. IAs located in the ICA are associated with a negative risk of rupture, while high AR (>1.01), SR1 (>1.48) or SR2 (>1.40) are risk factors for multiple IAs rupture. Copyright © 2017 Hainan Medical University. Production and hosting by Elsevier B.V. All rights reserved.

  5. Guidelines for Safety Evaluation of a Potential for PWR Steam Generator Tube Failure due to Fluid elastic Instability

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Do, Kyu Sik; Sheen, Cheol [Nuclear System Evaluation Dept., Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    It was found that both SG tube rupture events occurred at North Anna Unit 1 in 1987 and at Mihama Unit 2 in 1991 were caused by a high cycle fatigue due to fluid elastic instability. Therefore, with regard to nuclear safety it is important to design the SG properly in a conservative manner so that the potential for SG U-tube failures due to fluid elastic instability can be minimized. This article provides guidelines for assessing the potential for SG U-tube damage due to fluid elastic instability. This article described guidelines for safety evaluation of a potential for PWR steam generator tube failure due to fluid elastic instability. The guidelines address the requirements for realistically performing the SG thermal-hydraulic analysis and the modal analysis of tubes as well as the criteria for conservatively determining the added mass, the damping ratio and the fluid elastic instability coefficient. The guidelines can be used to predict the potential SG tubes which are susceptible to failure due to fluid elastic instability at operating nuclear power plants and also to evaluate the safety and structural integrity of new SG designs at the licensing review stage. Failure of a pressurized water reactor (PWR) steam generator (SG) tube leads to a leakage of contaminated primary coolant to the secondary system, which has serious safety implications such as the potential for direct release of radioactive fission products to the environment and the loss of coolant. Excessive tube vibration excited by dynamic forces of internal or external fluid flow is called flow-induced vibration (FIV). Among the FIV mechanisms, the so-called fluid elastic instability of SG tubes in cross flow is the most important safety issue in the design of SGs because it may cause severe tube failure in a very short time.

  6. Mechanical properties of cladding tubes made from type 1.4970 SS after irradiation in a Rapsodie-bundle

    International Nuclear Information System (INIS)

    Schaefer, L.

    1980-08-01

    The mechanical properties of pin sections are tested in tensile and stress-rupture tests. The dependence of the tensile properties of the irradiation temperature, of the dosis of fast neutrons and of the deformation rate is described. The results are as expected except for a maximum of the yield strength at 400 0 C and a fluence of 2 x 10 22 (nsub(s)/cm 2 ). Stress-rupture tests have shown that the weakest part of the pin is at the hot end of the fuel column. There the stress-rupture strength is only 60% of the strength of an unirradiated tube, because of corrosion with fission products and other influences. Taking into account the loss of cross section due to corrosion, the stress-rupture strength of pin sections agree with that of specimens from material irradiation experiments. The ductility is above 0.2% in the stress rupture test and above 0.5% in the tensile test. (orig.) [de

  7. Component external leakage and rupture frequency estimates

    International Nuclear Information System (INIS)

    Eide, S.A.; Khericha, S.T.; Calley, M.B.; Johnson, D.A.; Marteeny, M.L.

    1991-11-01

    In order to perform detailed internal flooding risk analyses of nuclear power plants, external leakage and rupture frequencies are needed for various types of components - piping, valves, pumps, flanges, and others. However, there appears to be no up-to-date, comprehensive source for such frequency estimates. This report attempts to fill that void. Based on a comprehensive search of Licensee Event Reports (LERs) contained in Nuclear Power Experience (NPE), and estimates of component populations and exposure times, component external leakage and rupture frequencies were generated. The remainder of this report covers the specifies of the NPE search for external leakage and rupture events, analysis of the data, a comparison with frequency estimates from other sources, and a discussion of the results

  8. 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

  9. 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.

  10. Dynamic rupture simulation of the 2017 Mw 7.8 Kaikoura (New Zealand) earthquake: Is spontaneous multi-fault rupture expected?

    Science.gov (United States)

    Ando, R.; Kaneko, Y.

    2017-12-01

    The coseismic rupture of the 2016 Kaikoura earthquake propagated over the distance of 150 km along the NE-SW striking fault system in the northern South Island of New Zealand. The analysis of In-SAR, GPS and field observations (Hamling et al., 2017) revealed that the most of the rupture occurred along the previously mapped active faults, involving more than seven major fault segments. These fault segments, mostly dipping to northwest, are distributed in a quite complex manner, manifested by fault branching and step-over structures. Back-projection rupture imaging shows that the rupture appears to jump between three sub-parallel fault segments in sequence from the south to north (Kaiser et al., 2017). The rupture seems to be terminated on the Needles fault in Cook Strait. One of the main questions is whether this multi-fault rupture can be naturally explained with the physical basis. In order to understand the conditions responsible for the complex rupture process, we conduct fully dynamic rupture simulations that account for 3-D non-planar fault geometry embedded in an elastic half-space. The fault geometry is constrained by previous In-SAR observations and geological inferences. The regional stress field is constrained by the result of stress tensor inversion based on focal mechanisms (Balfour et al., 2005). The fault is governed by a relatively simple, slip-weakening friction law. For simplicity, the frictional parameters are uniformly distributed as there is no direct estimate of them except for a shallow portion of the Kekerengu fault (Kaneko et al., 2017). Our simulations show that the rupture can indeed propagate through the complex fault system once it is nucleated at the southernmost segment. The simulated slip distribution is quite heterogeneous, reflecting the nature of non-planar fault geometry, fault branching and step-over structures. We find that optimally oriented faults exhibit larger slip, which is consistent with the slip model of Hamling et al

  11. Modeling and analysis of thermal damping in heat exchanger tube bundles

    Energy Technology Data Exchange (ETDEWEB)

    Khushnood, Shahab, E-mail: seeshahab@yahoo.co [University of Engineering and Technology, Taxila (Pakistan); Khan, Zaffar Muhammad, E-mail: mafzmlk@hotmail.co [National University of Sciences and Technology, Rawalpindi (Pakistan); Malik, Muhammad Afzaal [National University of Sciences and Technology, Rawalpindi (Pakistan); Iqbal, Qamar, E-mail: qamarch@yahoo.co [University of Engineering and Technology, Taxila (Pakistan); Bashir, Sajid; Khan, Muddasar [University of Engineering and Technology, Taxila (Pakistan); Koreshi, Zafarullah, E-mail: zaffark@yahoo.co [Air University, Islamabad (Pakistan); Khan, Mahmood Anwar [National University of Sciences and Technology, Rawalpindi (Pakistan); Malik, Tahir Nadeem [University of Engineering and Technology, Taxila (Pakistan); Qureshi, Arshad Hussain [University of Engineering and Technology, Lahore (Pakistan)

    2010-07-15

    Most structures and equipment used in nuclear power plant and process plant, such as reactor internals, fuel rods, steam generator tubes bundles, and process heat exchanger tube bundles, are subjected to flow-induced vibrations (FIV). Costly plant shutdowns have been the source of motivation for continuing studies on cross-flow-induced vibration in these structures. Damping has been the target of various research attempts related to FIV in tube bundles. A recent research attempt has shown the usefulness of a phenomenon termed as 'thermal damping'. The current paper focuses on the modeling and analysis of thermal damping in tube bundles subjected to cross-flow. It is expected that the present attempt will help in establishing improved design guidelines with respect to damping in tube bundles.

  12. Influence of composition on precipitation behavior and stress rupture properties in INCONEL RTM740 series superalloys

    Science.gov (United States)

    Casias, Andrea M.

    Increasing demands for energy efficiency and reduction in CO2 emissions have led to the development of advanced ultra-supercritical (AUSC) boilers. These boilers operate at temperatures of 760 °C and pressures of 35 MPa, providing efficiencies close to 50 pct. However, austenitic stainless steels typically used in boiler applications do not have sufficient creep or oxidation resistance. For this reason, nickel (Ni)-based superalloys, such as IN740, have been identified as potential materials for AUSC boiler tube components. However, IN740 is susceptible to heat-affected-zone liquation cracking in the base metal of heavy section weldments. To improve weldability, IN740H was developed. However, IN740H has lower stress rupture ductility compared to IN740. For this reason, two IN740H modifications have been produced by lowering carbon content and increasing boron content. In this study, IN740, IN740H, and the two modified IN740H alloys (modified 1 and 2) were produced with equiaxed grain sizes of 90 ìm (alloys IN740, IN740H, and IN740H modified 1 alloys) and 112 µm (IN740H modified 2 alloy). An aging study was performed at 800 °C on all alloys for 1, 3, 10, and 30 hours to assess precipitation behavior. Stress rupture tests were performed at 760 °C with the goal of attaining stress levels that would yield rupture at 1000 hours. The percent reduction in area was measured after failure as a measure of creep ductility. Light optical, scanning electron, and transmission electron microscopy were used in conjunction with X-ray diffraction to examine precipitation behavior of annealed, aged, and stress rupture tested samples. The amount and type of precipitation that occurred during aging prior to stress rupture testing or in-situ during stress rupture testing influenced damage development, stress rupture life, and ductility. In terms of stress rupture life, IN740H modified 2 performed the best followed by IN740H modified 1 and IN740, which performed similarly, and IN740

  13. Long-term results after repair of ruptured and non-ruptured abdominal aortic aneurysm

    Directory of Open Access Journals (Sweden)

    Kuzmanović Ilija B.

    2004-01-01

    peripheral arteries and other vascular reconstructive procedures were the factors that significantly reduced long-term survival of patients operated immediately due to rupture. DISCUSSION This comprehensive study has searched for more factors than others had done before. The applied discriminative analysis numerically evaluated the influence of any risk factor of mortality. These factors were divided in three groups as follows: preoperative, operative and postoperative ones. Preoperative factors were sex, age, diabetes mellitus, arterial hypertension, obesity, COPD, and naturally, the indication for operative treatment of ruptured or non-ruptured abdominal aneurysm. Among all these factors, only obesity significantly reduced long-term survival of electively operated patients. It may be said that immediately operated patients who survived the first 30 postoperative days had quite good long-term survival. Operative factors such as type of operative procedure and vascular graft had no influence on long-term survival of patients in both groups. Postoperative risk factors were early postoperative complications, graft infection, symptomatic cerebrovascular disease, carotid endarterectomy, myocardial revascularization, ventral hernias, "other" non vascular operations, malignancy, mental disorders, peripheral aneurysms and occlusive vascular disease, and other vascular operations either due to aneurysm or peripheral occlusive disease. Early postoperative complications (even graft infection had no significant effect on long-term survival. Ventral hernias and peripheral aneurysms were factors that significantly decreased long-term survival of patients operated for rupture of the abdominal aneurysm. CONCLUSION It is interesting that endarterectomy, myocardial revascularization or malignancy after repair of the abdominal aneurysm (ruptured or non-ruptured had no effect on long-term survival.

  14. Common and uncommon CT findings in rupture and impending rupture of abdominal aortic aneurysms

    International Nuclear Information System (INIS)

    Ahmed, M.Z.; Ling, L.; Ettles, D.F.

    2013-01-01

    The rapid imaging evaluation and diagnosis of rupture and impending rupture of an abdominal aortic aneurysm (AAA) is imperative. This article describes the imaging findings of rupture, impending rupture, and other abdominal aortic abnormalities. It is important not to overlook AAA as the consequences can be life threatening. All patients who had open or endovascular repair of AAA rupture over 6 years (2008–2012) were identified from our departmental database. The computed tomography (CT) images of 99 patients were reviewed for relevant findings. The mean age of the patients was 65 years and 85% were male

  15. Calculation of relative tube/tube support plate displacements in steam generators under accident condition loads using non-linear dynamic analysis methodologies

    International Nuclear Information System (INIS)

    Smith, R.E.; Waisman, R.; Hu, M.H.; Frick, T.M.

    1995-01-01

    A non-linear analysis has been performed to determine relative motions between tubes and tube support plates (TSP) during a steam line break (SLB) event for steam generators. The SLB event results in blowdown of steam and water out of the steam generator. The fluid blowdown generates pressure drops across the TSPS, resulting in out-of-plane motion. The SLB induced pressure loads are calculated with a computer program that uses a drift-flux modeling of the two-phase flow. In order to determine the relative tube/TSP motions, a nonlinear dynamic time-history analysis is performed using a structural model that considers all of the significant component members relative to the tube support system. The dynamic response of the structure to the pressure loads is calculated using a special purpose computer program. This program links the various substructures at common degrees of freedom into a combined mass and stiffness matrix. The program accounts for structural non-linearities, including potential tube and TSP interaction at any given tube position. The program also accounts for structural damping as part of the dynamic response. Incorporating all of the above effects, the equations of motion are solved to give TSP displacements at the reduced set of DOF. Using the displacement results from the dynamic analysis, plate stresses are then calculated using the detailed component models. Displacements form the dynamic analysis are imposed as boundary conditions at the DOF locations, and the finite element program then solves for the overall distorted geometry. Calculations are also performed to assure that assumptions regarding elastic response of the various structural members and support points are valid

  16. The roentgenographic findings of achilles tendon rupture

    Energy Technology Data Exchange (ETDEWEB)

    Seouk, Kang Hyo; Keun, Rho Yong [Shilla General Hospital, Seoul (Korea, Republic of)

    1999-03-01

    To evaluate the diagnostic value of a lateral view of the ankles in Achilles tendon rupture. We performed a retrospective analysis of the roentgenographic findings of 15 patients with surgically proven Achilles tendon rupture. Four groups of 15 patients(normal, ankle sprain, medial lateral malleolar fracture, and calcaneal fracture) were analysed as reference groups. Plain radiographs were reviewed with regard to Kager's triangle, Arner's sign, Toygar's angle, ill defined radiolucent shadow through the Achilles tendon, sharpness of the anterior margin of Achilles tendon, and meniscoid smooth margin of the posterior skin surface of the ankle. Kager's triangle was deformed and disappeared after rupture of the Achilles tendon in nine patients(60%) with operative verification of the rupture, six patients(40%) had a positive Arner's sign, while none had a diminished Toygars angle. In 13 patients(87%) with a ruptured Achilles tendon, the thickness of this was nonuniform compared with the reference group. The anterior margin of the Achilles tendon became serrated and indistinct in 14 patients(93%) in whom this was ruptured. An abnormal ill defined radiolucent shadow through the Achilles tendon was noted in nine patient(60%), and nonparallelism between the anterior margin of the Achilles tendon and posterior skin surface of the ankle was detected in 11 patients(73%). The posterior skin surface of the ankle had a nodular surface margin in 13 patients(87%). A deformed Kager's triangle and Achilles tendon, and an abnormal ill defined radiolucent shadow through the Achilles tendon in a lateral view of the ankles are important findings for the diagnesis of in diagnosing achilles tendon rupture.

  17. The roentgenographic findings of achilles tendon rupture

    International Nuclear Information System (INIS)

    Seouk, Kang Hyo; Keun, Rho Yong

    1999-01-01

    To evaluate the diagnostic value of a lateral view of the ankles in Achilles tendon rupture. We performed a retrospective analysis of the roentgenographic findings of 15 patients with surgically proven Achilles tendon rupture. Four groups of 15 patients(normal, ankle sprain, medial lateral malleolar fracture, and calcaneal fracture) were analysed as reference groups. Plain radiographs were reviewed with regard to Kager's triangle, Arner's sign, Toygar's angle, ill defined radiolucent shadow through the Achilles tendon, sharpness of the anterior margin of Achilles tendon, and meniscoid smooth margin of the posterior skin surface of the ankle. Kager's triangle was deformed and disappeared after rupture of the Achilles tendon in nine patients(60%) with operative verification of the rupture, six patients(40%) had a positive Arner's sign, while none had a diminished Toygars angle. In 13 patients(87%) with a ruptured Achilles tendon, the thickness of this was nonuniform compared with the reference group. The anterior margin of the Achilles tendon became serrated and indistinct in 14 patients(93%) in whom this was ruptured. An abnormal ill defined radiolucent shadow through the Achilles tendon was noted in nine patient(60%), and nonparallelism between the anterior margin of the Achilles tendon and posterior skin surface of the ankle was detected in 11 patients(73%). The posterior skin surface of the ankle had a nodular surface margin in 13 patients(87%). A deformed Kager's triangle and Achilles tendon, and an abnormal ill defined radiolucent shadow through the Achilles tendon in a lateral view of the ankles are important findings for the diagnesis of in diagnosing achilles tendon rupture

  18. The diagnosis of breast implant rupture

    DEFF Research Database (Denmark)

    Hölmich, Lisbet R; Vejborg, Ilse; Conrad, Carsten

    2005-01-01

    participated in either one or two study MRI examinations, aiming at determining the prevalence and incidence of silent implant rupture, respectively, and who subsequently underwent explantation. Implant rupture status was determined by four independent readers and a consensus diagnosis of either rupture...... were in fact ruptured at surgery. Thirty-four of the 43 intact implants were described as intact at surgery. When categorising possible ruptures as ruptures, there were one false positive and nine false negative rupture diagnoses at MRI yielding an accuracy of 92%, a sensitivity of 89...

  19. Radiological analysis of subarachnoid hemorrhage from ruptured intracranial aneurysms

    International Nuclear Information System (INIS)

    Lee, Jong Doo; Suh, Jung Ho; Kim, Dong Ik

    1988-01-01

    The CT findings of 98 patients with subarachnoid hemorrhage due to aneurysmal rupture were analyzed and compared with cerebral angiography for the purpose of preangiographic prediction of aneurysmal location as well as evaluation of the CT features corresponding to the vasospasm or ischemic neurologic dysfunctions. The results were as follows: 1.Aneurysms could be identified on initial cerebral angiography in 82 out of 98 patients with subarachnoid hemorrhage and anterior communicating artery aneurysms were most common (42 cases), followed by MCA, posterior communicating artery, ICA, basilar artery in order of frequency. 2.The CT findings of those patients were hemorrhage in subarachnoid space (69%), localized hematoma (47%), ventricular dilatation (31%), enhancing nodule (23%), cisternal enhancement (20%), cerebral infarction (15%), ventricular hemorrhage (14%), and epidural hemorrhage (3%). 3.Localized hematoma was more prevalent in anterior communicating artery aneurysm rupture (54%), and less frequently in MCA, posterior communicating artery and ICA aneurysms. 4.Most of aneurysmal sac could be identified as enhancing nodule on CT when the real size were over 1 cm. 5.The size of ruptured aneurysm could be predicted in many patients with ACA and MCA aneurysm according to the CT features such as hemorrhagic patterns, location of hematomas or enhancing nodules. 6.Localized hematoma or blood clots and cerebral infarction are considered to be the CT features corresponding to the angiographic vasospasm

  20. One-leg hop kinematics 20 years following anterior cruciate ligament rupture: Data revisited using functional data analysis.

    Science.gov (United States)

    Hébert-Losier, Kim; Pini, Alessia; Vantini, Simone; Strandberg, Johan; Abramowicz, Konrad; Schelin, Lina; Häger, Charlotte K

    2015-12-01

    Despite interventions, anterior cruciate ligament ruptures can cause long-term deficits. To assist in identifying and treating deficiencies, 3D-motion analysis is used for objectivizing data. Conventional statistics are commonly employed to analyze kinematics, reducing continuous data series to discrete variables. Conversely, functional data analysis considers the entire data series. Here, we employ functional data analysis to examine and compare the entire time-domain of knee-kinematic curves from one-leg hops between and within three groups. All subjects (n=95) were part of a long-term follow-up study involving anterior cruciate ligament ruptures treated ~20 years ago conservatively with physiotherapy only or with reconstructive surgery and physiotherapy, and matched knee-healthy controls. Between-group differences (injured leg, treated groups; non-dominant leg, controls) were identified during the take-off and landing phases, and in the sagittal (flexion/extension) rather than coronal (abduction/adduction) and transverse (internal/external) planes. Overall, surgical and control groups demonstrated comparable knee-kinematic curves. However, compared to controls, the physiotherapy-only group exhibited less flexion during the take-off (0-55% of the normalized phase) and landing (44-73%) phase. Between-leg differences were absent in controls and the surgically treated group, but observed during the flight (4-22%, injured leg>flexion) and the landing (57-85%, injured legFunctional data analysis identified specific functional knee-joint deviations from controls persisting 20 years post anterior cruciate ligament rupture, especially when treated conservatively. This approach is suggested as a means for comprehensively analyzing complex movements, adding to previous analyses. Copyright © 2015 Elsevier Ltd. All rights reserved.

  1. Splenic rupture following idiopathic rupture of the urinary bladder presenting as acute abdomen

    Directory of Open Access Journals (Sweden)

    Jurisic D

    2007-01-01

    Full Text Available Idiopathic rupture of the urinary bladder is an uncommon condition and represents less than 1% of all bladder rupture cases. In most of the cases the main etiological factor was heavy alcohol ingestion. A combined injury of the spleen and bladder is a very rare condition that is almost often associated with trauma and foreign bodies. In this paper we present the extremely rare clinical course of acute abdomen caused by a combined spontaneous intraperitoneal injury; spontaneous rupture of the urinary bladder and spleen. According to our opinion, spontaneous bladder rupture caused by bladder distension due to alcohol ingestion led to urinary ascites and abdominal distension. Finally, repeated minor abdominal blunt trauma during everyday life, to a moderately distended abdomen caused a spontaneous splenic rupture in the patient with abnormal coagulation studies.

  2. Vortex dynamics in ruptured and unruptured intracranial aneurysms

    Science.gov (United States)

    Trylesinski, Gabriel

    the current hypothesized biological triggers of pathological remodeling of the artery walls. Having a good natural ratio of statuses in our IA cohort (55 unruptured vs. 19 ruptured), we were able to test the statistical significance of our predictor to fortify our findings. We also performed a distribution analysis of our cohort with respect to the number of WKV to strengthen the encouraging statistical analysis result; both analyses provided a clear good separation of the status of the aneurysms based on our predictor. Lastly, we constructed a receiver operating characteristic (ROC) curve to analyze the power different thresholds of WKV had in splitting the data in a binary way (unruptured/ruptured). The number of WKV was efficaciously able to stratify the rupture status, identifying 84.21 % of the ruptured aneurysms (with 25.45 % of false positives, i.e. unruptured IAs tagged as ruptured) when using a threshold value of 2. Our novel work undertaken to study the vortex structures in IAs brought to light interesting characteristics of the flow in the aneurysmal sac. We found that there are several distinct categories in which the aneurysm vortex topologies can be put in without relationship to the aneurysm rupture status. This first finding was in contradiction with available already-published results. Nonetheless, ruptured IAs had a statistically significant larger amount of WKV as opposed to unruptured aneurysms. This new predictor we propose to the community could very well clear a new path among the currently controversial WSS-based parameters. Although it needs to be improved to be more resilient, the first results obtained by the WKV-based parameter are promising when applied to a large dataset of 74 IAs patient-specific transient CFD simulations.

  3. Spontaneous rupture of ovarian cystadenocarcinoma: pre- and post-rupture computed tomography evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Salvadori, Priscila Silveira; Atzingen, Augusto Castelli von; D' Ippolito, Giuseppe [Universidade Federal de Sao Paulo (EPM/UNIFESP), Sao Paulo, SP (Brazil). Escola Paulista de Medicina; Bomfim, Lucas Novais [Universidade Tiradentes (UNIT), Maceio, AL, (Brazil)

    2015-09-15

    Epithelial ovarian tumors are the most common malignant ovarian neoplasms and, in most cases, eventual rupture of such tumors is associated with a surgical procedure. The authors report the case of a 54-year-old woman who presented with spontaneous rupture of ovarian cystadenocarcinoma documented by computed tomography, both before and after the event. In such cases, a post-rupture staging tends to be less favorable, compromising the prognosis. (author)

  4. Frequency, predisposing factors and fetomaternal outcome in uterine rupture

    International Nuclear Information System (INIS)

    Malik, H.S.

    2006-01-01

    To determine the frequency and to analyze the predisposing factors, maternal and fetal outcome of uterine rupture. All cases of ruptured uterus, who were either admitted with or who developed this complication in the hospital, were included in the study. Demographic data, details regarding the most probable predisposing factor, type of rupture, the management and maternal and fetal outcome were taken into consideration for analysis.During three years, total number of deliveries was 18668, and there were 103 cases of uterine rupture (0.55%).Out of these, only 13 (12.62%) patients were booked. Most of the patients presented between the ages of 26-30 years (42.71%). Majority of ruptures occurred in para 2-4 (44.66%). Fifty five cases (53.39%) had a previous caesarean section scar. In 68 (66.01%) cases, the tear was located in lower uterine segment. In 93 (90.29%) cases, anterior uterine wall was involved. Rupture was complete in 79 (76.69%)cases. Repair of uterus was done in 79 (76.69%) cases. Hysterectomy was performed in 24 (23.30%) cases. There were 8 (7.76% or 77.66/1000) maternal deaths and 85 (81.73% or 825 / 1000) perinatal deaths.This study confirms high frequency of such serious preventable obstetrical problem which can lead to high fetomaternal mortality. Rupture of caesarean section scar was the most common cause of uterine rupture found in this series. (author)

  5. Mechanical reliability analysis of tubes intended for hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-02-15

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

  6. Vibro-impact responses of a tube with tube--baffle interaction

    International Nuclear Information System (INIS)

    Shin, Y.S.; Sass, D.E.; Jendrzejczyk, J.A.

    1978-01-01

    The relatively small, inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the vibrational characteristics and the response of the tube. Numerical studies were made to predict the vibro-impact response of a tube with tube-baffle interaction. The finite element method has been employed with a non-linear elastic contact spring-dashpot to model the effect of the relative approach between the tube and the baffle plate. The coupled equations of motion are directly integrated with a proportional system damping represented by a linear combination of mass and stiffness. Lumped mass approach with explicit time integration scheme was found to be a suitable choice for tube-baffle impacting analysis. Fourier analyses indicate that the higher mode contributions to the tube response are significant for strong tube-baffle impacting. The contact damping forces are negligible compared with the contact spring forces. The numerical analysis results are in reasonably good agreement with those of the experiments

  7. An analysis of uterine rupture at the Nnamdi Azikiwe University Teaching Hospital Nnewi, Southeast Nigeria.

    Science.gov (United States)

    Mbamara, S U; Obiechina, Nja; Eleje, G U

    2012-01-01

    Uterine rupture is a preventable condition which has persistently remained in our environment. The aim of this study therefore is to ascertain the incidence of uterine rupture, examine the predisposing factors and maternal and fetal outcome of patients managed of uterine rupture in a tertiary hospital. This descriptive case series was conducted at the department of Obstetrics and Gynaecology, Nnamdi Azikiwe, University Teaching Hospital Nnewi from March 2004 to February 2009. The incidence of uterine rupture was 6.2 per 1000 deliveries. The commonest age range of occurrence was 30-34 years. Uterine rupture occurred predominantly among women of low parity. Previous caesarean section with concurrent use of oxytocics was the commonest risk factor documented.The maternal and perinatal mortality ratio was 94 per 100,000 deliveries and 6 per 1000 births respectively. Surgery was the main stay of treatment and the commonest procedure carried out was uterine repair only. Rupture of the gravid uterus is still a significant cause of maternal mortality and morbidity in our environment. The causes are commonly preventable. The provision of maternal care by skilled personnel, proper antenatal care, update training programmes for health care providers and appropriate legislation on maternal care will significantly reduce the incidence of uterine rupture and improve its prognosis.

  8. Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albright, D.C.; Bari, R.A.

    1976-04-01

    In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident.

  9. Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Albright, D.C.; Bari, R.A.

    1976-04-01

    In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident

  10. Application of probabilistic fracture mechanics to optimize the maintenance of PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pitner, P.; Riffard, T.

    1993-09-01

    This paper describes the COMPROMIS code developed by Electricite de France (EDF) to optimize the tube bundle maintenance of steam generators (SG). The model, based on probabilistic fracture mechanics, makes it possible to quantify the influence of in-service inspections and maintenance work on the risk of an SG tube rupture, taking all significant parameters into account as random variables (initial defect size distribution, reliability of nondestructive detection and sizing, crack initiation and propagation, critical sizes, leak before risk of break, etc). (authors). 14 figs., 4 tabs., 12 refs

  11. Treatment strategy for ruptured abdominal aortic aneurysms.

    Science.gov (United States)

    Davidovic, L

    2014-07-01

    Rupture is the most serious and lethal complication of the abdominal aortic aneurysm. Despite all improvements during the past 50 years, ruptured abdominal aortic aneurysms are still associated with very high mortality. Namely, including patients who die before reaching the hospital, the mortality rate due to abdominal aortic aneurysm rupture is 90%. On the other hand, during the last twenty years, the number of abdominal aortic aneurysms significantly increased. One of the reasons is the fact that in majority of countries the general population is older nowadays. Due to this, the number of degenerative AAA is increasing. This is also the case for patients with abdominal aortic aneurysm rupture. Age must not be the reason of a treatment refusal. Optimal therapeutic option ought to be found. The following article is based on literature analysis including current guidelines but also on my Clinics significant experience. Furthermore, this article show cases options for vascular medicine in undeveloped countries that can not apply endovascular procedures at a sufficient level and to a sufficient extent. At this moment the following is evident. Thirty-day-mortality after repair of ruptured abdominal aortic aneurysms is significantly lower in high-volume hospitals. Due to different reasons all ruptured abdominal aortic aneurysms are not suitable for EVAR. Open repair of ruptured abdominal aortic aneurysm should be performed by experienced open vascular surgeons. This could also be said for the treatment of endovascular complications that require open surgical conversion. There is no ideal procedure for the treatment of AAA. Each has its own advantages and disadvantages, its own limits and complications, as well as indications and contraindications. Future reductions in mortality of ruptured abdominal aortic aneurysms will depend on implementation of population-based screening; on strategies to prevent postoperative organ injury and also on new medical technology

  12. A Rare Case of Simultaneous Acute Bilateral Quadriceps Tendon Rupture and Unilateral Achilles Tendon Rupture

    Directory of Open Access Journals (Sweden)

    Wei Yee Leong

    2013-07-01

    Full Text Available Introduction: There have been multiple reported cases of bilateral quadriceps tendon ruptures (QTR in the literature. These injuries frequently associated with delayed diagnosis, which results in delayed surgical treatment. In very unusual cases, bilateral QTRs can be associated with other simultaneous tendon ruptures. Case Report: We present a rare case of bilateral QTR with a simultaneous Achilles Tendon Rupture involving a 31 years old Caucasian man who is a semi-professional body builder taking anabolic steroids. To date bilateral QTR with additional TA rupture has only been reported once in the literature and to our knowledge this is the first reported case of bilateral QTR and simultaneous TA rupture in a young, fit and healthy individual. Conclusion: The diagnosis of bilateral QTR alone can sometimes be challenging and the possibility of even further tendon injuries should be carefully assessed. A delay in diagnosis could result in delay in treatment and potentially worse outcome for the patient. Keywords: Quadriceps tendon rupture; Achilles tendon rupture; Bilateral.

  13. Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

    International Nuclear Information System (INIS)

    Strong, B.R. Jr.; Baschiere, R.J.

    1978-01-01

    The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the 'force equivalent area' (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques. (Auth.)

  14. Morphology parameters for intracranial aneurysm rupture risk assessment.

    Science.gov (United States)

    Dhar, Sujan; Tremmel, Markus; Mocco, J; Kim, Minsuok; Yamamoto, Junichi; Siddiqui, Adnan H; Hopkins, L Nelson; Meng, Hui

    2008-08-01

    The aim of this study is to identify image-based morphological parameters that correlate with human intracranial aneurysm (IA) rupture. For 45 patients with terminal or sidewall saccular IAs (25 unruptured, 20 ruptured), three-dimensional geometries were evaluated for a range of morphological parameters. In addition to five previously studied parameters (aspect ratio, aneurysm size, ellipticity index, nonsphericity index, and undulation index), we defined three novel parameters incorporating the parent vessel geometry (vessel angle, aneurysm [inclination] angle, and [aneurysm-to-vessel] size ratio) and explored their correlation with aneurysm rupture. Parameters were analyzed with a two-tailed independent Student's t test for significance; significant parameters (P 41; 95% confidence interval, 1.03-1.92) and undulation index (odds ratio, 1.51; 95% confidence interval, 1.08-2.11) had the strongest independent correlation with ruptured IA. From the receiver operating characteristic analysis, size ratio and aneurysm angle had the highest area under the curve values of 0.83 and 0.85, respectively. Size ratio and aneurysm angle are promising new morphological metrics for IA rupture risk assessment. Because these parameters account for vessel geometry, they may bridge the gap between morphological studies and more qualitative location-based studies.

  15. Failure evaluation on a high-strength alloy SA213-T91 super heater tube of a power generation

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, J.; Purbolaksono, J.; Beng, L.C.; Ahmad, A. [University of Tenaga Nas, Kajang (Malaysia). Dept. of Mechanical Engineering

    2010-07-01

    This article presents failure investigation on a high-strength alloy SA213-T91 superheater tube. This failure is the first occurrence involving the material in Kapar Power Station Malaysia. The investigation includes visual inspections, hardness measurements, and microscopic examinations. The failed super-heater tube shows a wide open rupture with thin and blunt edges. Hardness readings on all the as-received tubes are used for estimating the operating metal temperature of the super-heater tubes. Microstructures of the failed tube show numerous creep cavities consisting of individual pores and chain of pores which form micro-and macro-cracks. The findings confirmed that the super-heater tube is failed by short-term overheating. Higher temperatures of the flue gas due to the inconsistent feeding of pulverized fuels into the burner is identified to cause overheating of the failed tube.

  16. Shared and Distinct Rupture Discriminants of Small and Large Intracranial Aneurysms.

    Science.gov (United States)

    Varble, Nicole; Tutino, Vincent M; Yu, Jihnhee; Sonig, Ashish; Siddiqui, Adnan H; Davies, Jason M; Meng, Hui

    2018-04-01

    Many ruptured intracranial aneurysms (IAs) are small. Clinical presentations suggest that small and large IAs could have different phenotypes. It is unknown if small and large IAs have different characteristics that discriminate rupture. We analyzed morphological, hemodynamic, and clinical parameters of 413 retrospectively collected IAs (training cohort; 102 ruptured IAs). Hierarchal cluster analysis was performed to determine a size cutoff to dichotomize the IA population into small and large IAs. We applied multivariate logistic regression to build rupture discrimination models for small IAs, large IAs, and an aggregation of all IAs. We validated the ability of these 3 models to predict rupture status in a second, independently collected cohort of 129 IAs (testing cohort; 14 ruptured IAs). Hierarchal cluster analysis in the training cohort confirmed that small and large IAs are best separated at 5 mm based on morphological and hemodynamic features (area under the curve=0.81). For small IAs (IAs (area under the curve=0.84; 95% confidence interval, 0.78-0.88), whereas for large IAs (≥5 mm), the model included undulation index, low wall shear stress, previous subarachnoid hemorrhage, and IA location (area under the curve=0.87; 95% confidence interval, 0.82-0.93). The model for the aggregated training cohort retained all the parameters in the size-dichotomized models. Results in the testing cohort showed that the size-dichotomized rupture discrimination model had higher sensitivity (64% versus 29%) and accuracy (77% versus 74%), marginally higher area under the curve (0.75; 95% confidence interval, 0.61-0.88 versus 0.67; 95% confidence interval, 0.52-0.82), and similar specificity (78% versus 80%) compared with the aggregate-based model. Small (IAs have different hemodynamic and clinical, but not morphological, rupture discriminants. Size-dichotomized rupture discrimination models performed better than the aggregate model. © 2018 American Heart Association, Inc.

  17. Flow Analysis of Isobutane (R-600A) Inside AN Adiabatic Capillary Tube

    Science.gov (United States)

    Alok, Praveen; Sahu, Debjyoti

    2018-02-01

    Capillary tubes are simple narrow tubes but the phase change which occurs inside the capillary tubes is complex to analyze. In the present investigation, an attempt is made to analyze the flow of Isobutane (R-600a) inside the coiled capillary tubes for different load conditions by Homogeneous Equilibrium Model. The Length and diameter of the capillary tube not only depend on the pressure and temperature of the condenser and evaporator but also on the cooling load. The present paper investigates the change in dimensions of the coil capillary tube with respect to the change in cooling load on the system for the constant condenser and evaporator conditions. ANSYS CFX (Central Florida Expressway) software is used to study the flow characteristics of the refrigerant. Appropriate helical coil is selected for this analysis.

  18. Uterine rupture: a retrospective analysis of causes, complications ...

    African Journals Online (AJOL)

    We conducted a retrospective review of case notes (from 2003 to 2009) to determine the incidence, causes, complications and foetal/maternal outcome among women with a diagnosis of ... Out of 72,570 deliveries 163 cases of ruptured uterus were recorded in seven years, making an incidence of 2.25 per 1000 births.

  19. Coincident steam generator tube rupture and stuck-open safety relief valve carryover tests: MB-2 steam generator transient response test program

    International Nuclear Information System (INIS)

    Garbett, K.; Mendler, O.J.; Gardner, G.C.; Garnsey, R.; Young, M.Y.

    1987-03-01

    In PWR steam generator tube rupture (SGTR) faults, a direct pathway for the release of radioactive fission products can exist if there is a coincident stuck-open safety relief valve (SORV) or if the safety relief valve is cycled. In addition to the release of fission products from the bulk steam generator water by moisture carryover, there exists the possibility that some primary coolant may be released without having first mixed with the bulk water - a process called primary coolant bypassing. The MB-2 Phase II test program was designed specifically to identify the processes for droplet carryover during SGTR faults and to provide data of sufficient accuracy for use in developing physical models and computer codes to describe activity release. The test program consisted of sixteen separate tests designed to cover a range of steady-state and transient fault conditions. These included a full SGTR/SORV transient simulation, two SGTR overfill tests, ten steady-state SGTR tests at water levels ranging from very low levels in the bundle up to those when the dryer was flooded, and three moisture carryover tests without SGTR. In these tests the influence of break location and the effect of bypassing the dryer were also studied. In a final test the behavior with respect to aerosol particles in a dry steam generator, appropriate to a severe accident fault, was investigated

  20. Hepatic rupture in preeclampsia

    International Nuclear Information System (INIS)

    Winer-Muram, H.T.; Muram, D.; Salazar, J.; Massie, J.D.

    1985-01-01

    The diagnosis of hepatic rupture in patients with pregnancy-induced hypertension (preeclampsia and eclampsia) is rarely made preoperatively. Diagnostic imaging can be utilized in some patients to confirm the preoperative diagnosis. Since hematoma formation precedes hepatic rupture, then, when diagnostic modalities such as sonography and computed tomography identify patients with hematomas, these patients are at risk of rupture, and should be hospitalized until the hematomas resolve

  1. Failure analysis of burst tested fuel tube samples

    International Nuclear Information System (INIS)

    Padmaprabu, C.; Ramana Rao, S.V.; Srivatsava, R.K.

    2005-01-01

    The Total Circumferential Elongation (TCE) is an important parameter for evaluation of ductility of the Zircaloy-4 fuel tubes for the PHWR reactors. The TCE values of the fuel tubes were obtained using the burst testing technique. In some lots there is a variation in the values of the TCE. To investigate the reasons for such a large variation in the TCE, samples were selected at appropriate intervals and sectioned at the fractured portion. The surface morphology of the fractured surfaces was examined under Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectrometer (EDS). The morphologies show segregation of elements at specific locations. Energy dispersive spectra was obtained from those segregated particles. According to the magnitude of TCE value the samples were classified into low, intermediate and high ductility. Low ductility samples were found to contain large amount of segregations along the thickness direction of the tube. This forms a brittle region and a path for the easy crack growth along thickness direction. In the case of intermediate samples the segregation occurred in fewer locations compared to low ductile samples and also confined to the circumferential direction of the outside surface of the tube. Due to this, probability of crack formation at the surface of the tube could be high. But crack growth would be slower in the ductile matrix along the thickness direction resulting in the enhancement of TCE value compared to the low ductile sample. In the high ductile samples, the segregations were very scarce and found to be isolated and embedded in the ductile matrix. The mode of failure in these types of samples was found to be purely ductile. Cracks were found to originate solely from the micro voids in the material. As the probability of crack formation and its propagation is low, very high TCE values were observed in these samples. Microstructural observations of fractured surfaces and EDAX analysis was able to identify the

  2. On the mechanism of explosive eruption of mount erebus volcano: the dynamics of the rupture structure in a cavitating layer

    International Nuclear Information System (INIS)

    Bol'shakova, E S; Kedrinskiy, V K

    2016-01-01

    This paper presents the results of an experimental simulation of rupture development in heavily cavitating magma melt flow in volcanic conduits and its effect on the structure of explosive volcanic eruptions. The dynamics of the state of a layer of distilled water (similar in the density of cavitation nuclei to magma melt) under shock-wave loading was studied. The experiments were performed using electromagnetic hydrodynamic shock tubes (EM HST) with maximum capacitor bank energy of up to 100 J and 5 kJ. It was found that the topology of the rupture formed on the membrane surface did not change during its development. Empirical estimates were obtained for the proportion of the capacitor bank energy expended in the development of the rupture and the characteristic time of its existence. The study revealed a number of fundamentally new physical effects in the cavity dynamics in a cavitating medium: a cavitation “boundary layer” is formed on the surface of the quasi-empty rupture, which is transformed into a cluster of high energy density upon closure of the flow. (paper)

  3. Leak on a steam generator tube: in-depth analysis

    International Nuclear Information System (INIS)

    Berger, J.; Deotto, G.; Mathon, C.; Madurel, A.; Pitner, P.; Gay, N.; Guivarch, M.

    2015-01-01

    A circumferential through crack was observed on a steam generator tube of the unit 2 of the Fessenheim plant. Destructive tests showed that the crack was due to cycle fatigue combined with the presence of inter-granular corrosion zones. An in-depth analysis based on simulations shows that the combination of 5 elements caused the crack. First, a specific position of the anti-vibration bar near this tube, secondly, a local presence of fouling, these 2 first elements led to an increase of the tube vibratory level. Thirdly, the 600 MA alloy used is known to be susceptible to corrosion. Fourthly, the trapping of chemical species on the secondary circuit side due to the presence of interstices on the crosspiece and fifthly, the presence of spots where inter-granular corrosion developed. (A.C.)

  4. Thermal analysis on x-ray tube for exhaust process

    Science.gov (United States)

    Kumar, Rakesh; Rao Ratnala, Srinivas; Veeresh Kumar, G. B.; Shivakumar Gouda, P. S.

    2018-02-01

    It is great importance in the use of X-rays for medical purposes that the dose given to both the patient and the operator is carefully controlled. There are many types of the X- ray tubes used for different applications based on their capacity and power supplied. In present thesis maxi ray 165 tube is analysed for thermal exhaust processes with ±5% accuracy. Exhaust process is usually done to remove all the air particles and to degasify the insert under high vacuum at 2e-05Torr. The tube glass is made up of Pyrex material, 95%Tungsten and 5%rhenium is used as target material for which the melting point temperature is 3350°C. Various materials are used for various parts; during the operation of X- ray tube these waste gases are released due to high temperature which in turn disturbs the flow of electrons. Thus, before using the X-ray tube for practical applications it has to undergo exhaust processes. Initially we build MX 165 model to carry out thermal analysis, and then we simulate the bearing temperature profiles with FE model to match with test results with ±5%accuracy. At last implement the critical protocols required for manufacturing processes like MF Heating, E-beam, Seasoning and FT.

  5. Fluidelastic instability analysis of steam generator U-tubes at antivibration bar-inactive modes

    International Nuclear Information System (INIS)

    Lee, S.K.; Jo, J.C.

    1995-01-01

    This paper presents the results of thermal-hydraulic and fluidelastic U-tube instability analyses performed for the vertical type pressurized water reactor (PWR) steam generator model being employed at Kori units 2, 3 and 4, and Yonggwang units 1 and 2 in Korea. The thermal-hydraulic analysis for providing the detailed three-dimensional two-phase flow field in the secondary side of the steam generator was accomplished using the ATHOS3 steam generator thermal-hydraulic analysis code. The UTVA2 code designed for calculating both the free vibration responses and fluidelastic stability ratio of a specific U-tube under consideration was used to assess the potential for fluidelastic instability of the steam generator U-tubes at various conditions of antivibration bar (AVB)-inactive modes. The results of the fluidelastic instability analysis were discussed in comparison with those obtained for the steam generator U-tubes at AVB-active mode

  6. Stress analysis and fatigue life prediction for a U-bend steam generator tube

    International Nuclear Information System (INIS)

    Cheng Weili; Finnie, I.

    1996-01-01

    An analysis is carried out to determine the stresses in a steam generator tube that failed by fatigue. Using data available for the failed tube and for failures in two similar steam generators, the magnitudes of the alternating and mean stresses produced during operation are estimated. The cause for the early fatigue failure is shown to be the high mean stress caused by denting of the tube in the location where it passed through the tube sheet. (orig.)

  7. 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.

  8. 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.

  9. 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.

  10. Searching for evidence of a preferred rupture direction in small earthquakes at Parkfield

    Science.gov (United States)

    Kane, D. L.; Shearer, P. M.; Allmann, B.; Vernon, F. L.

    2009-12-01

    Theoretical modeling of strike-slip ruptures along a bimaterial interface suggests that the interface will have a preferred rupture direction and will produce asymmetric ground motion (Shi and Ben-Zion, 2006). This could have widespread implications for earthquake source physics and for hazard analysis on mature faults because larger ground motions would be expected in the direction of rupture propagation. Studies have shown that many large global earthquakes exhibit unilateral rupture, but a consistently preferred rupture direction along faults has not been observed. Some researchers have argued that the bimaterial interface model does not apply to natural faults, noting that the rupture of the M 6 2004 Parkfield earthquake propagated in the opposite direction from previous M 6 earthquakes along that section of the San Andreas Fault (Harris and Day, 2005). We analyze earthquake spectra from the Parkfield area to look for evidence of consistent rupture directivity along the San Andreas Fault. We separate the earthquakes into spatially defined clusters and quantify the differences in high-frequency energy among earthquakes recorded at each station. Propagation path effects are minimized in this analysis because we compare earthquakes located within a small volume and recorded by the same stations. By considering a number of potential end-member models, we seek to determine if a preferred rupture direction is present among small earthquakes at Parkfield.

  11. Analysis of forming limit in tube hydroforming

    International Nuclear Information System (INIS)

    Kim, Chan Il; Yang, Seung Hang; Kim, Young Suk

    2013-01-01

    The automotive industry has shown increasing interest in tube hydroforming. Despite many automobile structural parts being produced from cylindrical tubes, failures frequently occur during tube hydroforming under improper forming conditions. These problems include wrinkling, buckling, folding back, and bursting. We perform analytical studies to determine forming limits in tube hydroforming and demonstrate how these forming limits are influenced by the loading path. Theoretical results for the forming limits of wrinkling and bursting are compared with experimental results for an aluminum tube.

  12. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  13. Phase Identification and Internal Stress Analysis of Steamside Oxides on Plant Exposed Superheater Tubes

    DEFF Research Database (Denmark)

    Pantleon, Karen; Montgomery, Melanie

    2012-01-01

    During long-term, high-temperature exposure of superheater tubes in thermal power plants, various oxides are formed on the inner side (steamside) of the tubes, and oxide spallation is a serious problem for the power plant industry. Most often, oxidation in a steam atmosphere is investigated...... in laboratory experiments just mimicking the actual conditions in the power plant for simplified samples. On real plant-exposed superheater tubes, the steamside oxides are solely investigated microscopically. The feasibility of X-ray diffraction for the characterization of steamside oxidation on real plant......-exposed superheater tubes was proven in the current work; the challenges for depth-resolved phase analysis and phase-specific residual stress analysis at the inner side of the tubes with concave surface curvature are discussed. Essential differences between the steamside oxides formed on two different steels...

  14. Vibro-impact responses of a tube with tube--baffle interaction. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Y S; Sass, D E; Jendrzejczyk, J A

    1978-01-01

    The relatively small, inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the vibrational characteristics and the response of the tube. Numerical studies were made to predict the vibro-impact response of a tube with tube-baffle interaction. The finite element method has been employed with a non-linear elastic contact spring-dashpot to model the effect of the relative approach between the tube and the baffle plate. The coupled equations of motion are directly integrated with a proportional system damping represented by a linear combination of mass and stiffness. Lumped mass approach with explicit time integration scheme was found to be a suitable choice for tube-baffle impacting analysis. Fourier analyses indicate that the higher mode contributions to the tube response are significant for strong tube-baffle impacting. The contact damping forces are negligible compared with the contact spring forces. The numerical analysis results are in reasonably good agreement with those of the experiments.

  15. Endometriosis-related spontaneous diaphragmatic rupture.

    Science.gov (United States)

    Triponez, Frédéric; Alifano, Marco; Bobbio, Antonio; Regnard, Jean-François

    2010-10-01

    Non-traumatic, spontaneous diaphragmatic rupture is a rare event whose pathophysiology is not known. We report the case of endometriosis-related spontaneous rupture of the right diaphragm with intrathoracic herniation of the liver, gallbladder and colon. We hypothesize that the invasiveness of endometriotic tissue caused diaphragm fragility, which finally lead to its complete rupture without traumatic event. The treatment consisted of a classical management of diaphragmatic rupture, with excision of the endometriotic nodule followed by medical ovarian suppression for six months.

  16. Inelastic analysis of finite length and depth cracked tubes

    International Nuclear Information System (INIS)

    Reich, M.; Gardner, D.; Prachuktam, S.; Chang, T.Y.

    1977-01-01

    Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdowns. This paper concerns itself with the prediction of the failure pressures for typical PWR steam generator tubes with longitudinal finite length and finite depth cracks. Only local plastic overload failure is considered since the material is non-notch sensitive. Non-linear finite element analyses are carried out to determine the burst pressures of steam generator tubes containing longitudinal cracks located on the outer surface of the tubes. The non-linearities considered herein include elastic-plastic material behavior and large deformations. A non-proprietary general purpose non-linear finite element program, NFAP was adopted for the analysis. Due to the asymmetric nature of the cracks, two-dimensional, as well as three-dimensional finite element analyses, were performed. The two-dimensional element and its formulations are similar to those of NONSAP. The three-dimensional isoparametric element with elastic-plastic material characteristics together with the large deformation formulations used in NFAP are described in the Report BNL-20684. The numerical accuracy of the program was investigated and checked with known solutions of benchmark problems. In addition to the three-dimensional element which was specifically inserted into NFAP for this problem, other features such as direct pressure inputs for isoparametric elements, automatic load increment adjustments for convergent non-linear solutions, and automatic bandwidth reduction schemes are incorporated into the program thus allowing for a more economical evaluation of three-dimensional inelastic analysis. In summary the analysis clearly shows that for short cracks axial effects play a significant role. For long cracks, they are not important since two-dimensional conditions predominate and failure is governed by circumferential or hoop stress conditions

  17. Failure Analysis and Magnetic Evaluation of Tertiary Superheater Tube Used in Gas-Fired Boiler

    Science.gov (United States)

    Mohapatra, J. N.; Patil, Sujay; Sah, Rameshwar; Krishna, P. C.; Eswarappa, B.

    2018-02-01

    Failure analysis was carried out on a prematurely failed tertiary superheater tube used in gas-fired boiler. The analysis includes a comparative study of visual examination, chemical composition, hardness and microstructure at failed region, adjacent and far to failure as well as on fresh tube. The chemistry was found matching to the standard specification, whereas the hardness was low in failed tube compared to the fish mouth opening region and the fresh tube. Microscopic examination of failed sample revealed the presence of spheroidal carbides of Cr and Mo predominantly along the grain boundaries. The primary cause of failure is found to be localized heating. Magnetic hysteresis loop (MHL) measurements were carried out to correlate the magnetic parameters with microstructure and mechanical properties to establish a possible non-destructive evaluation (NDE) for health monitoring of the tubes. The coercivity of the MHL showed a very good correlation with microstructure and mechanical properties deterioration enabling a possible NDE technique for the health monitoring of the tubes.

  18. Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

    International Nuclear Information System (INIS)

    Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

    1999-12-01

    In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

  19. Time-Domain Analysis of Coupled Carbon Nano tube Interconnects

    International Nuclear Information System (INIS)

    Fathi, D.

    2014-01-01

    This paper describes a new method for the analysis of coupling effects including the crosstalk effects between two driven coupled single-walled carbon nano tubes (SWCNTs) and the intertalk effects between two neighboring shells in a multi walled carbon nano tube (MWCNT), based on transmission line circuit modeling. Using rigorous calculations, a new parametric transfer function has been obtained for the analysis of the impact of aggressor line on the victim line, which depends on the various coupling parameters such as the mutual inductance, the coupling capacitance, and the tunneling resistance. The influences of various parameters such as the contact resistance and the switching factor on the time behavior of coupling effects between the two coupled CNTs and an important effect named “crosstalk-induced delay” are studied and analyzed

  20. New considerations on variability of creep rupture data and life prediction

    International Nuclear Information System (INIS)

    Kim, Seon Jin; Jeong, Won Taek; Kong, Yu Sik

    2009-01-01

    This paper deals with the variability analysis of short term creep rupture test data based on the previous creep rupture tests and the possibility of the creep life prediction. From creep tests performed by constant uniaxial stresses at 600, 650 and 700 .deg. C elevated temperature, in order to investigate the variability of short-term creep rupture data, the creep curves were analyzed for normalized creep strain divided by initial strain. There are some variability in thee creep rupture data. And, the difference between general creep curves and normalized creep curves were obtained. The effects of the creep rupture time and state steady creep rate on the Weibull distribution parameters were investigated. There were good relation between normal Weibull parameters and normalized Weibull parameters. Finally, the predicted creep life were compared with the Monkman-Grant model.

  1. New Considerations on Variability of Creep Rupture Data and Life Prediction

    International Nuclear Information System (INIS)

    Jung, Won Taek; Kong, Yu Sik; Kim, Seon Jin

    2009-01-01

    This paper deals with the variability analysis of short term creep rupture test data based on the previous creep rupture tests and the possibility of the creep life prediction. From creep tests performed by constant uniaxial stresses at 600, 650 and 700 .deg. C elevated temperature, in order to investigate the variability of short-term creep rupture data, the creep curves were analyzed for normalized creep strain divided by initial strain. There are some variability in the creep rupture data. And, the difference between general creep curves and normalized creep curves were obtained. The effects of the creep rupture time (RT) and steady state creep rate (SSCR) on the Weibull distribution parameters were investigated. There were good relation between normal Weibull parameters and normalized Weibull parameters. Finally, the predicted creep life were compared with the Monkman-Grant model

  2. The Numerical and Experimental Analysis of Ballizing Process of Steel Tubes

    Directory of Open Access Journals (Sweden)

    Dyl T.

    2017-06-01

    Full Text Available This paper presents chosen results of experimental and numerical research of ballizing process of the steel tubes. Ballizing process is a method of burnishing technology of an internal diameter by precisely forcing a ball through a slightly undersized pre-machined tubes. Ballizing process is a fast, low-cost process for sizing and finishing tubes. It consists of pressing a slightly oversized ball through an unfinished tube to quickly bring the hole to desired size. The ball is typically made from a very hard material such as tungsten carbide or bearing steel. Ballizing process is by cold surface plastic forming of the surface structure, thereby leaving a layer of harder material and reducing its roughness. After theoretical and experimental analysis it was determined that the smaller the diameter of the balls, the bigger intensity of stress and strain and strain rate. The paper presents influence of ballizing process on the strain and stress state and on the surface roughness reduction rate of the steel tubes.

  3. Tube manufacturing trials by different routes in 9CrW-ODS martensitic steels

    International Nuclear Information System (INIS)

    Ukai, S.; Narita, T.; Alamo, A.; Parmentier, P.

    2004-01-01

    In the collaboration work between JNC and CEA-Saclay, JNC and CEA independently manufactured ODS martensitic cladding tubes by their own fabrication routes. Manufacturing started from the same hollow shape mother tubes with a composition of 9Cr-2W-0.1Ti-0.24Y 2 O 3 . The HPTR cold-rolling process was used by both JNC and CEA, but the applied fabrication routes included different cross-section reduction ratios, number of passes and intermediate heat treatments. The manufactured claddings exhibited an isotropic grain structure and equivalent tensile strength in the longitudinal and transverse directions. Even though different cross-section reduction ratios and intermediate annealing treatments were used, both cladding tubes manufactured by JNC and CEA showed similar levels of tensile and internal creep rupture strength

  4. Experimental investigation on the effect of the tube vibration on the aerosol retention during SGTR meltdown sequences

    International Nuclear Information System (INIS)

    Tardaguila, R. D.; Herranz, L. E.

    2013-01-01

    In PWRs Steam Generator Tube Rupture (SGTR) severe accident sequences scenario, with containment bypass, may become a significant contribution to the NPP risk. Since last two decades the EU-SGTR, ARTIST 1 and 2 and the on-going ARTIST-extension programs have investigated the potential attenuation of the source term in these accidental sequences. Thanks to them, it has been identified key factors that could influence on the source term attenuation as the tube vibration. This paper presents the results of the Phenomenon Test (PT) campaign, focused on the vibration influence on the mass retention on the break stage of a SG and the characterization of the tubes vibration.

  5. Retrospective Review of Pectoralis Major Ruptures in Rodeo Steer Wrestlers

    Directory of Open Access Journals (Sweden)

    Breda H. F. Lau

    2013-01-01

    Full Text Available Background. Pectoralis major tendon ruptures have been reported in the literature as occupational injuries, accidental injuries, and sporting activities. Few cases have been reported with respect to rodeo activities. Purpose. To describe a series of PM tendon ruptures in professional steer wrestlers. Study Design. Case series, level of evidence, 4. Methods. A retrospective analysis of PM ruptures in a steer wrestling cohort was performed. Injury data between 1992 and 2008 were reviewed using medical records from the University of Calgary Sport Medicine Center. Results. Nine cases of pectoralis major ruptures in professional steer wrestlers were identified. Injuries occurred during the throwing phase of the steer or while breaking a fall. All athletes reported unexpected or abnormal behavior of the steer that contributed to the mechanism of injury. Seven cases were surgically repaired, while two cases opted for nonsurgical intervention. Eight cases reported successful return to competition following the injury. Conclusion. Steer wrestlers represent a unique cohort of PM rupture case studies. Steer wrestling is a demanding sport that involves throwing maneuvers that may predispose the muscle to rupture. All cases demonstrated good functional outcomes regardless of surgical or non-surgical treatment.

  6. PREMATURE RUPTURE OF THE MEMBRANES*

    African Journals Online (AJOL)

    In patients presenting with premature rupture of the membranes there are two factors which influence the foetal morbidity and mortality. These factors are prema- turity and intra-uterine infection. The purpose of this analysis was to elucidate which factor carried the greater risk to the foetus. Recently there has been a spate of.

  7. Numerical analysis of creep brittle rupture by the finite element method

    International Nuclear Information System (INIS)

    Goncalves, O.J.A.; Owen, D.R.J.

    1983-01-01

    In this work an implicit algorithm is proposed for the numerical analysis of creep brittle rupture problems by the finite element method. This kind of structural failure, typical in components operating at high temperatures for long periods of time, is modelled using either a three dimensional generalization of the Kachanov-Rabotnov equations due to Leckie and Hayhurst or the Monkman-Grant fracture criterion together with the Linear Life Fraction Rule. The finite element equations are derived by the displacement method and isoparametric elements are used for the spatial discretization. Geometric nonlinear effects (large displacements) are accounted for by an updated Lagrangian formulation. Attention is also focussed on the solution of the highly stiff differential equations that govern damage growth. Finally the numerical results of a three-dimensional analysis of a pressurized thin cylinder containing oxidised pits in its external wall are discussed. (orig.)

  8. Stresses, fatigue and fracture analysis in the tube sheets

    International Nuclear Information System (INIS)

    Billon, F.

    1986-05-01

    The purpose of the present work is to study the behaviour of the nuclear PWR steam generator tube sheet. But the methods developed in this field can easily be generalized in order to study tube sheets from any other type of heat exchangers. The aim of the stress analysis of these sheets is to verify their correct design, to quantify the risk of fatigue damage in the areas submitted to a high stress concentration and through the fracture mechanic, to make sure there is no risk of fast fracture resulting from initiated or pre-existing defects. This analysis necessarily relates to the calculation of stresses in all parts of the multidrilled area, mainly around the holes where they are concentrated. However the tube sheets are so complexe structures that their direct modelization cannot be envisaged within the context of the finite element method. We then must refer to the concept of equivalent medium in order to calculate the nominal stresses. Then using the stresses multiple fonctions appropriate to the net geometry, we can calculate the actual stresses concentrated around the holes. The method depends on the behaviour of the elementary volume which represents the behaviour of the multidrilled medium. This approach must allow to correctly take account of the ''thermal skin effect'', which is a phenomenon particular to the tube sheets with thermal loads. It must as well be generalized in order to analyse the irregular ligaments which affect the periodical stresses distribution and locally overconcentrate them [fr

  9. Laser interferometer system for the measurement of creep in pressurized tubes

    International Nuclear Information System (INIS)

    Kirchner, T.L.

    1976-07-01

    A laser interferometer measurement system was developed to measure the length, diameter, and radius of various pressurized tube specimens. The machine measures and records profilometric data of the pressurized tubes prior to insertion in the reactor and then again after a predetermined fluence has been reached to determine the amount of creep which has occurred. This data provides a statistical basis for the description of steady-state in-reactor creep and creep rupture behavior of the reference fuel cladding and structural materials for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR). In addition, this data will be used to determine the relative in-reactor creep and creep rupture behavior of candidate alloys for advanced cladding and structural materials. The laser interferometer system, referred to as the Biaxial Creep Measurement Machine (BCMM), was built to meet or exceed design criteria such as: automatic measurement of the five biaxial creep specimens varying in size; complete automation of the machine using a mini-computer; complete specimen loading, unloading, and data processing in less than five minutes; storage of data on magnetic cassette tapes; quick-look data readout and error checking during each run to determine proper machine operation; and remote operation in a radioactive environment

  10. A Case Report of Ruptured Spontaneous Heterotopic Pregnancy

    Directory of Open Access Journals (Sweden)

    F Seidoshohadaei

    2008-04-01

    Full Text Available ABSTRACT: Introduction & Objective: Heterotopic pregnancy refers to the simultaneous occurrence of pregnancy intrauterine and outside of uterine corpus. It is most often manifested in women who have undergone artificial reproductive technology (ART but rarely occurs spontaneously. Heterotopic pregnancy still remains as a diagnostic and therapeutic challenge to practitioners. In this situation physicians should have high suspicion for diagnosis and intrauterine pregnancy protection. This study reported a case of ruptured spontaneous heterotopic pregnancy. Case: A 32 year-old woman with abdominal pain, nausea, vomiting and hypovolumic shock in 1386 referred to emergency department in Sanandaj hospital. She reported one previous cesarean section. On examination, the patient's abdomen was distended. She had generalized tenderness and rebound tenderness in abdomen. The ultrasonographic examination revealed large amount of fluid in pelvic and abdominal cavity with a large hematoma in right adnex but there was intrauterine pregnancy at 7 weeks with normal fetal heart activity. She underwent laparotomy for heterotopic pregnancy and ruptured tube with tubal pregnancy removed. Intrauterine pregnancy continued without problem and led to birth of a healthy female neonate. Conclusion: Physicians should be quite cautious of heterotopic pregnancy in woman at reproductive age. Any abnormality on physical examination or ultrasonography of a patient with intrauterine pregnancy and abdominal pain should heighten the clinician's suspicion for heterotopic pregnancy

  11. Elastic-plastic analysis of tube expansion in tubesheets

    International Nuclear Information System (INIS)

    Kasraie, B.; O'Donnell, W.J.; Porowski, J.S.; Selz, A.

    1983-01-01

    Conditions for expansion of tubes in tubesheets are often determined by the test. The tightness of the joint and pull out force are used as criteria for evaluation of the results. For closely spaced tubes, it is also necessary to control development of the plastic regions in the ligaments surrounding the tube being expanded. High local strains may occur and excessive distortion may result if the expansion of the tube is continued beyond the admissible limits. Elastic-plastic finite element analyses are performed herein in order to establish conditions for rolling of the tubes in tubesheets of low ligament efficiency. Such penetration patterns are often required in the design of tubular reactors for catalytic processes. The model considered includes individual tube expansion in tubesheets with triangular penetration patterns. The effect of prior expansion of the neighboring tubes is also evaluated. Gap elements are used to model the initial clearance of the tube in the hole. Development of the plastic zones and distortion of the ligaments is monitored during radial expansion of the tube diameter. The residual stresses between the tube and the hole surface and the history of gap closing after removal of the expansion tool are determined. The effect of axial extension of the tube on the tube thinning is determined. Tube thinning is often used as a measure of tube expansion in manufacturing processes. For the analyzed ligament efficiency, reliable joints are obtained for a thinning range within 2% to 3%

  12. Two and dimensional heat analysis inside a high pressure electrical discharge tube

    International Nuclear Information System (INIS)

    Aghanajafi, C.; Dehghani, A. R.; Fallah Abbasi, M.

    2005-01-01

    This article represents the heat transfer analysis for a horizontal high pressure mercury steam tube. To get a more realistic numerical simulation, heat radiation at different wavelength width bands, has been used besides convection and conduction heat transfer. The analysis for different gases with different pressure in two and three dimensional cases has been investigated and the results compared with empirical and semi empirical values. The effect of the environmental temperature on the arc tube temperature is also studied

  13. Rupture of Achilles Tendon : Usefulness of Ultrasonography

    International Nuclear Information System (INIS)

    Kim, Nam Hyeon; Ki, Won Woo; Yoon, Kwon Ha; Kim, Song Mun; Shin, Myeong Jin; Kwon, Soon Tae

    1996-01-01

    To differentiate a complete rupture of Achilles tendon from an incomplete one which is important because its treatment is quite different. And it is necessary to know the exact site of the rupture preoperatively. Fifteen cases of fourteen patients which were diagnosed as Achilles tendon rupture by ultrasonography and surgery were reviewed. We compared sonographic rupture site with surgical findings. Ultrasonographic criteria for differentiation of complete and incomplete rupture was defined as follows : the discreteness, which means the proximal intervening hypoechogenicity to the interface echogenicity of distal margin of ruptured tendon : the slant sign, which represents the interface of ruptured distal margin which was seen over the 3/4 of the thickness of the tendon without intervening low echogeneicity : the invagination sign, which means the echogenic invagination from Kager triangle into posterior aspect of Achilles tendon over the half thickness of the tendon. The sites of complete tendon rupture were exactly corresponded to surgical finding in four cases of ten complete ruptures. And the discrepancy between sonographic and surgical findings in the site of complete rupture was 1.2 ± 0.4 cm in six cases. Three of ten complete ruptures showed the discreteness sign, all of ten showed the slant sign and two of ten showed the invagination sign. It is helpful to differentiate a complete from incomplete rupture of the Achilles tendon and to localize the site of the complete rupture with the ultrasonographic evaluation

  14. Morphological and Hemodynamic Discriminators for Rupture Status in Posterior Communicating Artery Aneurysms.

    Directory of Open Access Journals (Sweden)

    Nan Lv

    Full Text Available The conflicting findings of previous morphological and hemodynamic studies on intracranial aneurysm rupture may be caused by the relatively small sample sizes and the variation in location of the patient-specific aneurysm models. We aimed to determine the discriminators for aneurysm rupture status by focusing on only posterior communicating artery (PCoA aneurysms.In 129 PCoA aneurysms (85 ruptured, 44 unruptured, clinical, morphological and hemodynamic characteristics were compared between the ruptured and unruptured cases. Multivariate logistic regression analysis was performed to determine the discriminators for rupture status of PCoA aneurysms.While univariate analyses showed that the size of aneurysm dome, aspect ratio (AR, size ratio (SR, dome-to-neck ratio (DN, inflow angle (IA, normalized wall shear stress (NWSS and percentage of low wall shear stress area (LSA were significantly associated with PCoA aneurysm rupture status. With multivariate analyses, significance was only retained for higher IA (OR = 1.539, p < 0.001 and LSA (OR = 1.393, p = 0.041.Hemodynamics and morphology were related to rupture status of intracranial aneurysms. Higher IA and LSA were identified as discriminators for rupture status of PCoA aneurysms.

  15. Influence of fault steps on rupture termination of strike-slip earthquake faults

    Science.gov (United States)

    Li, Zhengfang; Zhou, Bengang

    2018-03-01

    A statistical analysis was completed on the rupture data of 29 historical strike-slip earthquakes across the world. The purpose of this study is to examine the effects of fault steps on the rupture termination of these events. The results show good correlations between the type and length of steps with the seismic rupture and a poor correlation between the step number and seismic rupture. For different magnitude intervals, the smallest widths of the fault steps (Lt) that can terminate the rupture propagation are variable: Lt = 3 km for Ms 6.5 6.9, Lt = 4 km for Ms 7.0 7.5, Lt = 6 km for Ms 7.5 8.0, and Lt = 8 km for Ms 8.0 8.5. The dilational fault step is easier to rupture through than the compression fault step. The smallest widths of the fault step for the rupture arrest can be used as an indicator to judge the scale of the rupture termination of seismic faults. This is helpful for research on fault segmentation, as well as estimating the magnitude of potential earthquakes, and is thus of significance for the assessment of seismic risks.

  16. Rupture of the Pitáycachi Fault in the 1887 Mw 7.5 Sonora, Mexico earthquake (southern Basin-and-Range Province): Rupture kinematics and epicenter inferred from rupture branching patterns

    Science.gov (United States)

    Suter, Max

    2015-01-01

    During the 3 May 1887 Mw 7.5 Sonora earthquake (surface rupture end-to-end length: 101.8 km), an array of three north-south striking Basin-and-Range Province faults (from north to south Pitáycachi, Teras, and Otates) slipped sequentially along the western margin of the Sierra Madre Occidental Plateau. This detailed field survey of the 1887 earthquake rupture zone along the Pitáycachi fault includes mapping the rupture scarp and measurements of surface deformation. The surface rupture has an endpoint-to-endpoint length of ≥41.0 km, dips 70°W, and is characterized by normal left-lateral extension. The maximum surface offset is 487 cm and the mean offset 260 cm. The rupture trace shows a complex pattern of second-order segmentation. However, this segmentation is not expressed in the 1887 along-rupture surface offset profile, which indicates that the secondary segments are linked at depth into a single coherent fault surface. The Pitáycachi surface rupture shows a well-developed bipolar branching pattern suggesting that the rupture originated in its central part, where the polarity of the rupture bifurcations changes. Most likely the rupture first propagated bilaterally along the Pitáycachi fault. The southern rupture front likely jumped across a step over to the Teras fault and from there across a major relay zone to the Otates fault. Branching probably resulted from the lateral propagation of the rupture after breaching the seismogenic part of the crust, given that the much shorter ruptures of the Otates and Teras segments did not develop branches.

  17. Biomechanical rupture risk assessment of abdominal aortic aneurysms based on a novel probabilistic rupture risk index.

    Science.gov (United States)

    Polzer, Stanislav; Gasser, T Christian

    2015-12-06

    A rupture risk assessment is critical to the clinical treatment of abdominal aortic aneurysm (AAA) patients. The biomechanical AAA rupture risk assessment quantitatively integrates many known AAA rupture risk factors but the variability of risk predictions due to model input uncertainties remains a challenging limitation. This study derives a probabilistic rupture risk index (PRRI). Specifically, the uncertainties in AAA wall thickness and wall strength were considered, and wall stress was predicted with a state-of-the-art deterministic biomechanical model. The discriminative power of PRRI was tested in a diameter-matched cohort of ruptured (n = 7) and intact (n = 7) AAAs and compared to alternative risk assessment methods. Computed PRRI at 1.5 mean arterial pressure was significantly (p = 0.041) higher in ruptured AAAs (20.21(s.d. 14.15%)) than in intact AAAs (3.71(s.d. 5.77)%). PRRI showed a high sensitivity and specificity (discriminative power of 0.837) to discriminate between ruptured and intact AAA cases. The underlying statistical representation of stochastic data of wall thickness, wall strength and peak wall stress had only negligible effects on PRRI computations. Uncertainties in AAA wall stress predictions, the wide range of reported wall strength and the stochastic nature of failure motivate a probabilistic rupture risk assessment. Advanced AAA biomechanical modelling paired with a probabilistic rupture index definition as known from engineering risk assessment seems to be superior to a purely deterministic approach. © 2015 The Author(s).

  18. Vibration impact acoustic emission technique for identification and analysis of defects in carbon steel tubes: Part B Cluster analysis

    Energy Technology Data Exchange (ETDEWEB)

    Halim, Zakiah Abd [Universiti Teknikal Malaysia Melaka (Malaysia); Jamaludin, Nordin; Junaidi, Syarif [Faculty of Engineering and Built, Universiti Kebangsaan Malaysia, Bangi (Malaysia); Yahya, Syed Yusainee Syed [Universiti Teknologi MARA, Shah Alam (Malaysia)

    2015-04-15

    Current steel tubes inspection techniques are invasive, and the interpretation and evaluation of inspection results are manually done by skilled personnel. Part A of this work details the methodology involved in the newly developed non-invasive, non-destructive tube inspection technique based on the integration of vibration impact (VI) and acoustic emission (AE) systems known as the vibration impact acoustic emission (VIAE) technique. AE signals have been introduced into a series of ASTM A179 seamless steel tubes using the impact hammer. Specifically, a good steel tube as the reference tube and four steel tubes with through-hole artificial defect at different locations were used in this study. The AEs propagation was captured using a high frequency sensor of AE systems. The present study explores the cluster analysis approach based on autoregressive (AR) coefficients to automatically interpret the AE signals. The results from the cluster analysis were graphically illustrated using a dendrogram that demonstrated the arrangement of the natural clusters of AE signals. The AR algorithm appears to be the more effective method in classifying the AE signals into natural groups. This approach has successfully classified AE signals for quick and confident interpretation of defects in carbon steel tubes.

  19. Vibration impact acoustic emission technique for identification and analysis of defects in carbon steel tubes: Part B Cluster analysis

    International Nuclear Information System (INIS)

    Halim, Zakiah Abd; Jamaludin, Nordin; Junaidi, Syarif; Yahya, Syed Yusainee Syed

    2015-01-01

    Current steel tubes inspection techniques are invasive, and the interpretation and evaluation of inspection results are manually done by skilled personnel. Part A of this work details the methodology involved in the newly developed non-invasive, non-destructive tube inspection technique based on the integration of vibration impact (VI) and acoustic emission (AE) systems known as the vibration impact acoustic emission (VIAE) technique. AE signals have been introduced into a series of ASTM A179 seamless steel tubes using the impact hammer. Specifically, a good steel tube as the reference tube and four steel tubes with through-hole artificial defect at different locations were used in this study. The AEs propagation was captured using a high frequency sensor of AE systems. The present study explores the cluster analysis approach based on autoregressive (AR) coefficients to automatically interpret the AE signals. The results from the cluster analysis were graphically illustrated using a dendrogram that demonstrated the arrangement of the natural clusters of AE signals. The AR algorithm appears to be the more effective method in classifying the AE signals into natural groups. This approach has successfully classified AE signals for quick and confident interpretation of defects in carbon steel tubes.

  20. SCC analysis of Alloy 600 tubes from a retired steam generator

    Science.gov (United States)

    Hwang, Seong Sik; Kim, Hong Pyo

    2013-09-01

    Steam generators (SG) equipped with Alloy 600 tubes of a Korean nuclear power plants were replaced with a new one having Alloy 690 tubes in 1998 after 20 years of operation. To set up a guide line for an examination of the other SG tubes, a metallographic examination of the defected tubes was carried out. A destructive analysis on 71 tubes was addressed, and a relation among the stress corrosion crack (SCC) defect location, defect depth, and location of the sludge pile was obtained. Tubes extracted from the retired SG were transferred to a hot laboratory. Detailed nondestructive analysis examinations were taken again at the laboratory, and the tubes were then destructively examined. The types and sizes of the cracks were characterized. The location and depth of the SCC were evaluated in terms of the location and height of the sludge. Most axial cracks were in the sludge pile, whereas the circumferential ones were around the top of the tube sheet (TTS) or below the TTS. Average defect depth of the axial cracks was deeper than that of the circumferential ones. Axial cracks at tube support plate (TSP) seem to be related with corrosion/sludge in crevice like at the TTS region. Circumferential cracks at TSP seem to be caused by tube denting at the upper part of the TSP. Tubes not having clear ECT signals for quantifying an ECT data-base. Tubes having no ECT signal. Tubes with a large ECT signal. Tubes with various types and sizes of flaws (primary water stress corrosion cracking (PWSCC), outside diameter stress corrosion cracking (ODSCC), Pit). Tubes with distinct PWSCC or ODSCC. Tubes were extracted from the RSG based on the field ECT with the criteria, and transferred to a hot laboratory at the Korea Atomic Energy Research Institute (KAERI) for destructive examination. A comprehensive ECT inspection was performed again at the hot laboratory to confirm the location of the cracks obtained from a field inspection. These exact locations of the defects were marked on the

  1. Root cause analysis of SG tube leakage at Fessenheim unit 2 in 2008

    International Nuclear Information System (INIS)

    Berger, J.; Deotto, G.; Mathon, C.; Madurel, A.; Pitner, P.; Gay, N.; Guivarch, M.

    2015-01-01

    In February 2008, a primary-to-secondary leak caused an unscheduled shutdown at Fessenheim Unit 2 NPP. A circumferential crack was observed just above the top support plate of Row 12 Column 62 U-bend tube on Steam Generator (SG) number 3, which has been attributed to high cycle fatigue. This tube was pulled out in 2011, just before the SG replacement at the third decenal outage, in order to perform exhaustive metallurgical investigations. The destructive examinations revealed that the circumferential crack (70 degrees of extension) was due to high cycle fatigue, with several external initiation areas associated with the presence of small piles of Intergranular Attack (IGA) (600 MA tube) and with very low stress intensity factors ΔK (close to the non-propagating threshold region). This paper complements the metallurgical investigations by carrying out numerical analyses (thermal-hydraulic computation, fluid-elastic instability evaluation, tube vibratory response analysis and fatigue evaluation). The first objective of the study is to attempt to clarify the effect of IGA and the role of several competing factors that could be involved in the tube vibration induced fatigue failure. From these results, a root cause analysis of the R12C62 tube fatigue failure is then provided. It appears that a combination of various factors led to the failure of the tube

  2. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

    1996-01-01

    The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

  3. Morphological parameters associated with ruptured posterior communicating aneurysms.

    Science.gov (United States)

    Ho, Allen; Lin, Ning; Charoenvimolphan, Nareerat; Stanley, Mary; Frerichs, Kai U; Day, Arthur L; Du, Rose

    2014-01-01

    The rupture risk of unruptured intracranial aneurysms is known to be dependent on the size of the aneurysm. However, the association of morphological characteristics with ruptured aneurysms has not been established in a systematic and location specific manner for the most common aneurysm locations. We evaluated posterior communicating artery (PCoA) aneurysms for morphological parameters associated with aneurysm rupture in that location. CT angiograms were evaluated to generate 3-D models of the aneurysms and surrounding vasculature. Univariate and multivariate analyses were performed to evaluate morphological parameters including aneurysm volume, aspect ratio, size ratio, distance to ICA bifurcation, aneurysm angle, vessel angles, flow angles, and vessel-to-vessel angles. From 2005-2012, 148 PCoA aneurysms were treated in a single institution. Preoperative CTAs from 63 patients (40 ruptured, 23 unruptured) were available and analyzed. Multivariate logistic regression revealed that smaller volume (p = 0.011), larger aneurysm neck diameter (0.048), and shorter ICA bifurcation to aneurysm distance (p = 0.005) were the most strongly associated with aneurysm rupture after adjusting for all other clinical and morphological variables. Multivariate subgroup analysis for patients with visualized PCoA demonstrated that larger neck diameter (p = 0.018) and shorter ICA bifurcation to aneurysm distance (p = 0.011) were significantly associated with rupture. Intracerebral hemorrhage was associated with smaller volume, larger maximum height, and smaller aneurysm angle, in addition to lateral projection, male sex, and lack of hypertension. We found that shorter ICA bifurcation to aneurysm distance is significantly associated with PCoA aneurysm rupture. This is a new physically intuitive parameter that can be measured easily and therefore be readily applied in clinical practice to aid in the evaluation of patients with PCoA aneurysms.

  4. Tube to tube excursive instability - sensitivities and transients

    International Nuclear Information System (INIS)

    Brown, M.; Layland, M.W.

    1980-01-01

    A simple basic analysis of excursive instability in a boiler tube shows how it depends upon operating conditions and physical properties. A detailed mathematical model of an AGR boiler is used to conduct a steady state parameter sensitivity survey. It is possible from this basis to anticipate the effects of changes in operating conditions and changes in design parameters upon tube to tube stability. Dynamic responses of tubes operating near the stability threshold are examined using a mathematical model. Simulated excursions are triggered by imparting small abrupt pressure changes on the boiler inlet pressure. The influences of the magnitude of the pressure change, waterside friction factor and gas side coupling between tubes are examined. (author)

  5. Estimation of time to rupture in a fire using 6FIRE, a lumped parameter UF6 cylinder transient heat transfer/stress analysis model

    International Nuclear Information System (INIS)

    Williams, W.R.; Anderson, J.C.

    1995-01-01

    The transportation of UF 6 is subject to regulations requiring the evaluation of packaging under a sequence of hypothetical accident conditions including exposure to a 30-min 800 degree C (1475 degree F) fire [10 CFR 71.73(c)(3)]. An issue of continuing interest is whether bare cylinders can withstand such a fire without rupturing. To address this issue, a lumped parameter heat transfer/stress analysis model (6FIRE) has been developed to simulate heating to the point of rupture of a cylinder containing UF 6 when it is exposed to a fire. The model is described, then estimates of time to rupture are presented for various cylinder types, fire temperatures, and fill conditions. An assessment of the quantity of UF 6 released from containment after rupture is also presented. Further documentation of the model is referenced

  6. Experimental studies on the deformation and rupture of thin metal plates subject to underwater shock wave loading

    Directory of Open Access Journals (Sweden)

    Chen Pengwan

    2015-01-01

    Full Text Available In this paper, the dynamic deformation and rupture of thin metal plates subject to underwater shock wave loading are studied by using high-speed 3D digital image correlation (3D-DIC. An equivalent device consist of a gas gun and a water anvil tube was used to supplying an exponentially decaying pressure in lieu of explosive detonation which acted on the panel specimen. The thin metal plate is clamped on the end of the shock tube by a flange. The deformation and rupture process of the metal plates subject to underwater shock waves are recorded by two high-speed cameras. The shape, displacement fields and strain fields of the metal plates under dynamic loading are obtained by using VIC-3D digital image correlation software. The strain gauges also were used to monitor the structural response on the selected position for comparison. The DIC data and the strain gauges results show a high level of correlation, and 3D-DIC is proven to be an effective method to measure 3D full-field dynamic response of structures under underwater impact loading. The effects of pre-notches on the failure modes of thin circular plate were also discussed.

  7. DEVELOPMENT OF COILED TUBING STRESS ANALYSIS

    Directory of Open Access Journals (Sweden)

    Davorin Matanović

    1998-12-01

    Full Text Available The use of coiled tubing is increasing rapidly with drilling of horizontal wells. To satisfy all requirements (larger mechanical stresses, larger fluid capacities the production of larger sizes and better material qualities was developed. Stresses due to axial forces and pressures that coiled tubing is subjected are close to its performance limits. So it is really important to know and understand the behaviour of coiled tubing to avoid its break, burst or collapse in the well.

  8. Numerical Analysis for Heat Transfer Characteristics of Elliptic Fin-Tube Heat Exchanger with Various Shapes

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Hwan; Yoon, Jun Kyu [Gachon Univ., Seongnam (Korea, Republic of)

    2013-04-15

    In this study, the characteristics of the heat transfer coefficient and pressure drop were numerically analyzed according to the axis ratio (A R), pitch, location of vortex generator, and bump phase of the tube surface about an elliptical fin-tube heat exchanger. The boundary condition for CAD analysis was decided as a tube surface temperature of 348 K and inlet air velocity of 1.5 m/s. RCM 7th turbulent model was chosen as the numerical analysis for the sensitivity level. The analysis results indicated that the A R and transverse pitch decreased whereas the heat transfer coefficient increased. On the other hand, there was little difference in the longitudinal pitch. Furthermore, the heat transfer rate was more favorable when the vortex generator was located in front of the tube. Also, the bump phase of the tube surface indicated that the pressure drop and heat transfer were more favorable with the circle type than with the serrated type.

  9. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    International Nuclear Information System (INIS)

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J.

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR

  10. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    Energy Technology Data Exchange (ETDEWEB)

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

  11. Morphological and Hemodynamic Discriminators for Rupture Status in Posterior Communicating Artery Aneurysms.

    Science.gov (United States)

    Lv, Nan; Wang, Chi; Karmonik, Christof; Fang, Yibin; Xu, Jinyu; Yu, Ying; Cao, Wei; Liu, Jianmin; Huang, Qinghai

    2016-01-01

    The conflicting findings of previous morphological and hemodynamic studies on intracranial aneurysm rupture may be caused by the relatively small sample sizes and the variation in location of the patient-specific aneurysm models. We aimed to determine the discriminators for aneurysm rupture status by focusing on only posterior communicating artery (PCoA) aneurysms. In 129 PCoA aneurysms (85 ruptured, 44 unruptured), clinical, morphological and hemodynamic characteristics were compared between the ruptured and unruptured cases. Multivariate logistic regression analysis was performed to determine the discriminators for rupture status of PCoA aneurysms. While univariate analyses showed that the size of aneurysm dome, aspect ratio (AR), size ratio (SR), dome-to-neck ratio (DN), inflow angle (IA), normalized wall shear stress (NWSS) and percentage of low wall shear stress area (LSA) were significantly associated with PCoA aneurysm rupture status. With multivariate analyses, significance was only retained for higher IA (OR = 1.539, p PCoA aneurysms.

  12. Dynamic transient analysis of rupture disks by the finite-element method

    International Nuclear Information System (INIS)

    Hsieh, B.J.

    1975-02-01

    A finite element method utilizing the principle of virtual work in convected coordinates is used to analyze the axisymmetric dynamic transient response of rupture disks. This method can treat non-linearities arising both from inelastic material properties and large displacements/rotations provided that the convected strains are small. This report contains extensive calculations using a variety of rupture disk geometries and attempts to relate the static buckling of such disks to their dynamic response characteristics. A majority of the calculations treat the response of 18 inch disks typical of those currently considered for use in the Clinch River Breeder Reactor intermediate heat transport system

  13. High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments

    International Nuclear Information System (INIS)

    Chung, H.M.; Garde, A.M.; Kassner, T.F.

    1977-01-01

    The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350 0 C

  14. Predicting creep rupture from early strain data

    International Nuclear Information System (INIS)

    Holmstroem, Stefan; Auerkari, Pertti

    2009-01-01

    To extend creep life modelling from classical rupture modelling, a robust and effective parametric strain model has been developed. The model can reproduce with good accuracy all parts of the creep curve, economically utilising the available rupture models. The resulting combined model can also be used to predict rupture from the available strain data, and to further improve the rupture models. The methodology can utilise unfailed specimen data for life assessment at lower stress levels than what is possible from rupture data alone. Master curves for creep strain and rupture have been produced for oxygen-free phosphorus-doped (OFP) copper with a maximum testing time of 51,000 h. Values of time to specific strain at given stress (40-165 MPa) and temperature (125-350 deg. C) were fitted to the models in the strain range of 0.1-38%. With typical inhomogeneous multi-batch creep data, the combined strain and rupture modelling involves the steps of investigation of the data quality, extraction of elastic and creep strain response, rupture modelling, data set balancing and creep strain modelling. Finally, the master curves for strain and rupture are tested and validated for overall fitting efficiency. With the Wilshire equation as the basis for the rupture model, the strain model applies classical parametric principles with an Arrhenius type of thermal activation and a power law type of stress dependence for the strain rate. The strain model also assumes that the processes of primary and secondary creep can be reasonably correlated. The rupture model represents a clear improvement over previous models in the range of the test data. The creep strain information from interrupted and running tests were assessed together with the rupture data investigating the possibility of rupture model improvement towards lower stress levels by inverse utilisation of the combined rupture based strain model. The developed creep strain model together with the improved rupture model is

  15. Hydrogen attack evaluation of boiler tube using ultrasonic wave

    International Nuclear Information System (INIS)

    Won, Soon Ho; Hyun, Yang Ki; Lee, Jong O; Cho, Kyung Shik; Lee, Jae Do

    2001-01-01

    The presence of hydrogen in industrial plants is a source of damage. Hydrogen attack is one such form of degradation and often causing large tube ruptures that necessitate an immediate shutdown. Hydrogen attack may reduce the fracture toughness as well as the strength of steels. This reduction is caused partially by the presence of cavities and microcracks at the grain boundaries. In the past several techniques have been used with limited results. This paper describes the application of an ultrasonic velocity, attenuation and backscatter techniques for detecting the presence of hydrogen damage in utility boiler tubes. Ultrasonic tests showed a decrease in wave velocity and an increase in attenuation. Such results demonstrate the potential for ultrasonic nondestructive testing to quantify damage. Based on this study, recommendations are that both velocity and attenuation be used to detect hydrogen attack in steels.

  16. Pierce gain analysis for a sheet beam in a rippled waveguide traveling-wave tube

    International Nuclear Information System (INIS)

    Carlsten, Bruce E.

    2001-01-01

    A Pierce-type mode analysis is presented for a planar electron beam in a rippled planar waveguide. This analysis describes the gain of a traveling-wave tube consisting of that geometry. The dispersion relation is given by the determinant of a matrix based on the coupling of different free-space modes through the boundary conditions. For the case of high-frequency, low-power amplifiers, the dispersion relation reduces to a simple cubic expression for the Compton regime, leading to three roots analogous to the Pierce solution of a standard traveling-wave tube. The analysis shows that this type of traveling-wave tube is capable of very high gain at extremely high frequencies

  17. Development and application of an efficient method for performing modal analysis of steam generator tubes in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Huinam [Dept of Mechanical and Aerospace Engineering, Sunchon National University, Sunchon, 540-742 (Korea, Republic of); Boo, Myung-Hwan [Korea Hydro and Nuclear Power Company, Yuseong-Gu, Daejeon 305-343 (Korea, Republic of); Park, Chi-Yong [KEPCO Research Institute, Yuseong-Gu, Daejeon 305-380 (Korea, Republic of); Ryu, Ki-Wahn, E-mail: kwryu@chonbuk.ac.k [Department of Aerospace Engineering, Chonbuk National University, 664-14, Deogjin-Dong, Jeonju 561-756 (Korea, Republic of)

    2010-10-15

    A typical pressurized water reactor (PWR) steam generator has approximately 10,000 tubes. These tubes have different geometries, supporting conditions, and different material properties due to the non-uniform temperature distribution throughout the steam generator. Even though some tubes may have the same geometry and boundary conditions, the non-uniform distribution of coolant densities adjacent to the tubes causes them to have different added mass effects and dynamic characteristics. Therefore, for a reliable design of the steam generator, a separate modal analysis for each tube is necessary to perform the FIV (flow-induced vibration) analysis. However, the modal analysis of a tube including the finite element modeling is cumbersome and takes lots of time. And when a commercial finite element code is used, interfacing the modal analysis result, such as natural frequencies and mode shapes, with the FIV analysis procedure requires an additional significant amount of time and can possibly incur inadvertent error due to the complexity of data processing. It is therefore impossible to perform the complete FIV analysis for ten thousands of tubes when designing or maintaining a steam generator although it is necessary. Rather, to verify the safe design against the FIV, only a couple of tubes are chosen based on engineering judgment or past experience. In this paper, a computer program, PIAT-MODE, was developed which is able to perform modal analysis of all tubes of a PWR steam generator in a very efficient way. The geometries and boundary conditions of every tube were incorporated into PIAT-MODE using appropriate mathematical formulae. Material property data including the added mass effect was also included in the program. Once a specific tube is selected, the program automatically constructs the finite element model and generates the modal data very quickly. Therefore, modal analysis can be performed for every single tube in a straight way. When PIAT-MODE is coupled

  18. Tube cystostomy for management of obstructive urolithiasis in ruminants

    Directory of Open Access Journals (Sweden)

    P. Tamilmahan

    2014-04-01

    Full Text Available Aim: The aim of this study was to evaluate the simple tube cystostomy procedure for management of urethral obstruction cases in ruminants. Materials and Methods: Tube cystostomy was used to treat a total of 58 ruminants, which included 35 buffalo calves and 23 goats. Diagnosis of the disease was made with the history of anuria, clinical signs, and physical examinations. Physical parameters like heart rate, respiratory rate, rectal temperature dehydration status of animals by skin tenting test, and intraoperative findings were compared. Results: Young ruminants were most commonly affected and the mean age was 4-5 months in both species. Only male were considered for the study in which buffalo calves were not castrated but in goat's 73.91% animal were castrated and 34.7% not castrated. Rupture of bladder was more common in buffalo calves as compared to goats. The confirmed cases of obstructive urolithiasis were selected for tube cystostomy with Foley's catheter. Postoperatively all cases were administered with broad spectrum antibiotic, anti-inflammatory agent, and caliculolytic agents like ammonium chloride. Postoperative complications recorded only in 10 animals and remaining 48 animals had an uneventful recovery. Conclusion: Tube cystostomy is a simple and effective procedure particularly in intact urinary bladder, which can be adopted at field level.

  19. Scramjet test flow reconstruction for a large-scale expansion tube, Part 2: axisymmetric CFD analysis

    Science.gov (United States)

    Gildfind, D. E.; Jacobs, P. A.; Morgan, R. G.; Chan, W. Y. K.; Gollan, R. J.

    2017-11-01

    This paper presents the second part of a study aiming to accurately characterise a Mach 10 scramjet test flow generated using a large free-piston-driven expansion tube. Part 1 described the experimental set-up, the quasi-one-dimensional simulation of the full facility, and the hybrid analysis technique used to compute the nozzle exit test flow properties. The second stage of the hybrid analysis applies the computed 1-D shock tube flow history as an inflow to a high-fidelity two-dimensional-axisymmetric analysis of the acceleration tube. The acceleration tube exit flow history is then applied as an inflow to a further refined axisymmetric nozzle model, providing the final nozzle exit test flow properties and thereby completing the analysis. This paper presents the results of the axisymmetric analyses. These simulations are shown to closely reproduce experimentally measured shock speeds and acceleration tube static pressure histories, as well as nozzle centreline static and impact pressure histories. The hybrid scheme less successfully predicts the diameter of the core test flow; however, this property is readily measured through experimental pitot surveys. In combination, the full test flow history can be accurately determined.

  20. Measurement and analysis of pressure tube elongation in the Douglas Point reactor

    International Nuclear Information System (INIS)

    Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

    1980-02-01

    Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

  1. Thermodynamic analysis of a pulse tube engine

    International Nuclear Information System (INIS)

    Moldenhauer, Stefan; Thess, André; Holtmann, Christoph; Fernández-Aballí, Carlos

    2013-01-01

    Highlights: ► Numerical model of the pulse tube engine process. ► Proof that the heat transfer in the pulse tube is out of phase with the gas velocity. ► Proof that a free piston operation is possible. ► Clarifying the thermodynamic working principle of the pulse tube engine. ► Studying the influence of design parameters on the engine performance. - Abstract: The pulse tube engine is an innovative simple heat engine based on the pulse tube process used in cryogenic cooling applications. The working principle involves the conversion of applied heat energy into mechanical power, thereby enabling it to be used for electrical power generation. Furthermore, this device offers an opportunity for its wide use in energy harvesting and waste heat recovery. A numerical model has been developed to study the thermodynamic cycle and thereby help to design an experimental engine. Using the object-oriented modeling language Modelica, the engine was divided into components on which the conservation equations for mass, momentum and energy were applied. These components were linked via exchanged mass and enthalpy. The resulting differential equations for the thermodynamic properties were integrated numerically. The model was validated using the measured performance of a pulse tube engine. The transient behavior of the pulse tube engine’s underlying thermodynamic properties could be evaluated and studied under different operating conditions. The model was used to explore the pulse tube engine process and investigate the influence of design parameters.

  2. [Development of Achilles tendon rupture in skiing].

    Science.gov (United States)

    Suckert, K; Benedetto, K P; Vogel, A

    1983-06-01

    This is an analysis of decline of rupture of the Achilles tendon in skiing while there is a steady increase of skiing injuries. Three groups, equipped with three different types of ski boots were observed once on a plane slope on the other hand on a bump track. The simultaneous size of angle of knee and ankle was measured by telemetry. The high plastic ski boot, which obstructs the ankle forward and lateral is apart from the rise of heel in the boot, the safety binding and the new skiing style the main reason for decline of rupture of the Achilles tendon in skiing.

  3. WWER steam generator tube structural and leakage integrity

    International Nuclear Information System (INIS)

    Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

    1998-01-01

    The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

  4. Lower head failure analysis

    International Nuclear Information System (INIS)

    Rempe, J.L.; Thinnes, G.L.; Allison, C.M.; Cronenberg, A.W.

    1991-01-01

    The US Nuclear Regulatory Commission is sponsoring a lower vessel head research program to investigate plausible modes of reactor vessel failure in order to determine (a) which modes have the greatest likelihood of occurrence during a severe accident and (b) the range of core debris and accident conditions that lead to these failures. This paper presents the methodology and preliminary results of an investigation of reactor designs and thermodynamic conditions using analytic closed-form approximations to assess the important governing parameters in non-dimensional form. Preliminary results illustrate the importance of vessel and tube geometrical parameters, material properties, and external boundary conditions on predicting vessel failure. Thermal analyses indicate that steady-state temperature distributions will occur in the vessel within several hours, although the exact time is dependent upon vessel thickness. In-vessel tube failure is governed by the tube-to-debris mass ratio within the lower head, where most penetrations are predicted to fail if surrounded by molten debris. Melt penetration distance is dependent upon the effective flow diameter of the tube. Molten debris is predicted to penetrate through tubes with a larger effective flow diameter, such as a boiling water reactor (BWR) drain nozzle. Ex-vessel tube failure for depressurized reactor vessels is predicted to be more likely for a BWR drain nozzle penetration because of its larger effective diameter. At high pressures (between ∼0.1 MPa and ∼12 MPa) ex-vessel tube rupture becomes a dominant failure mechanism, although tube ejection dominates control rod guide tube failure at lower temperatures. However, tube ejection and tube rupture predictions are sensitive to the vessel and tube radial gap size and material coefficients of thermal expansion

  5. Some aspects of metallurgical assessment of boiler tubes-Basic principles and case studies

    International Nuclear Information System (INIS)

    Chaudhuri, Satyabrata

    2006-01-01

    Microstructural changes in boiler tubes during prolong operation at high temperature and pressure decrease load bearing capacity limiting their useful lives. When the load bearing capacity falls below a critical level depending on operating parameters and tube geometry, failure occurs. In order to avoid such failures mainly from the view point of economy and safety, this paper describes some basic principles behind remaining life assessment of service exposed components and also a few case studies related to failure of a reheater tube of 1.25Cr-0.5Mo steel, a carbon steel tube and final superheater tubes of 2.25Cr-1Mo steel and remaining creep life assessment of service exposed but unfailed platen superheater and reheater tubes of 2.25Cr-1Mo steel. Sticking of fly ash particles causing reduction in effective tube wall thickness is responsible for failure of reheater tubes. Decarburised metal containing intergranular cracks at the inner surface of the carbon steel tube exhibiting a brittle window fracture is an indicative of hydrogen embrittlement responsible for this failure. In contrast, final superheater tube showed that the failure took place due to short-term overheating. The influence of prolong service revealed that unfailed reheater tubes exhibit higher tensile properties than that of platen superheater tubes. In contrast both the tubes at 50 MPa meet the minimum creep rupture properties when compared with NRIM data. The remaining creep life of platen superheater tube as estimated at 50 MPa and 570 deg. C (1058 o F) is more than 10 years and that of reheater tube at 50 MPa and 580 deg. C (1076 o F) is 9 years

  6. Poly Implant Prothèse (PIP) incidence of rupture: a retrospective MR analysis in 64 patients.

    Science.gov (United States)

    Scotto di Santolo, Mariella; Cusati, Bianca; Ragozzino, Alfonso; Dell'Aprovitola, Nicoletta; Acquaviva, Alessandra; Altiero, Michele; Accurso, Antonello; Riccardi, Albina; Imbriaco, Massimo

    2014-12-01

    The purpose of this retrospective study was to describe the magnetic resonance imaging (MRI) features of Poly Implant Prothèse (PIP) hydrogel implants in a group of 64 patients and to assess the incidence of rupture, compared to other clinical trials. In this double-center study, we retrospectively reviewed the data sets of 64 consecutive patients (mean age, 43±9 years, age range, 27-65 years), who underwent breast MRI examinations, between January 2008 and October 2013, with suspected implant rupture on the basis of clinical assessment or after conventional imaging examination (either mammography or ultrasound). All patients had undergone breast operation with bilateral textured cohesive gel PIP implant insertion for aesthetic reasons. The mean time after operation was 8 years (range, 6-14 years). No patients reported history of direct trauma to their implants. At the time of clinical examination, 41 patients were asymptomatic, 16 complained of breast tenderness and 7 had clinical evidence of rupture. Normal findings were observed in 15 patients. In 26 patients there were signs of mild collapse, with associated not significant peri-capsular fluid collections and no evidence of implant rupture; in 23 patients there was suggestion of implant rupture, according to breast MRI leading to an indication for surgery. In particular, 14 patients showed intra-capsular rupture, with associated evidence of the linguine sign in all cases; the keyhole sign and the droplet signs were observed in 6 cases. In 9 patients there was evidence of extra-capsular rupture, with presence of axillary collections (siliconomas) in 7 cases and peri-prosthetic and mediastinal cavity siliconomas, in 5 cases. The results of this double center retrospective study, confirm the higher incidence (36%) of prosthesis rupture observed with the PIP implants, compared to other breast implants.

  7. Application of DFM in human reliability analysis

    International Nuclear Information System (INIS)

    Yu Shaojie; Zhao Jun; Tong Jiejuan

    2011-01-01

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

  8. Time-resolved observation of thermally activated rupture of a capillary-condensed water nanobridge

    International Nuclear Information System (INIS)

    Bak, Wan; Sung, Baekman; Kim, Jongwoo; Kwon, Soyoung; Kim, Bongsu; Jhe, Wonho

    2015-01-01

    The capillary-condensed liquid bridge is one of the most ubiquitous forms of liquid in nature and contributes significantly to adhesion and friction of biological molecules as well as microscopic objects. Despite its important role in nanoscience and technology, the rupture process of the bridge is not well understood and needs more experimental works. Here, we report real-time observation of rupture of a capillary-condensed water nanobridge in ambient condition. During slow and stepwise stretch of the nanobridge, we measured the activation time for rupture, or the latency time required for the bridge breakup. By statistical analysis of the time-resolved distribution of activation time, we show that rupture is a thermally activated stochastic process and follows the Poisson statistics. In particular, from the Arrhenius law that the rupture rate satisfies, we estimate the position-dependent activation energies for the capillary-bridge rupture

  9. Development of a computer model to predict aortic rupture due to impact loading.

    Science.gov (United States)

    Shah, C S; Yang, K H; Hardy, W; Wang, H K; King, A I

    2001-11-01

    Aortic injuries during blunt thoracic impacts can lead to life threatening hemorrhagic shock and potential exsanguination. Experimental approaches designed to study the mechanism of aortic rupture such as the testing of cadavers is not only expensive and time consuming, but has also been relatively unsuccessful. The objective of this study was to develop a computer model and to use it to predict modes of loading that are most likely to produce aortic ruptures. Previously, a 3D finite element model of the human thorax was developed and validated against data obtained from lateral pendulum tests. The model included a detailed description of the heart, lungs, rib cage, sternum, spine, diaphragm, major blood vessels and intercostal muscles. However, the aorta was modeled as a hollow tube using shell elements with no fluid within, and its material properties were assumed to be linear and isotropic. In this study fluid elements representing blood have been incorporated into the model in order to simulate pressure changes inside the aorta due to impact. The current model was globally validated against experimental data published in the literature for both frontal and lateral pendulum impact tests. Simulations of the validated model for thoracic impacts from a number of directions indicate that the ligamentum arteriosum, subclavian artery, parietal pleura and pressure changes within the aorta are factors that could influence aortic rupture. The model suggests that a right-sided impact to the chest is potentially more hazardous with respect to aortic rupture than any other impact direction simulated in this study. The aortic isthmus was the most likely site of aortic rupture regardless of impact direction. The reader is cautioned that this model could only be validated on a global scale. Validation of the kinematics and dynamics of the aorta at the local level could not be done due to a lack of experimental data. It is hoped that this model will be used to design

  10. MRI of tibialis anterior tendon rupture

    International Nuclear Information System (INIS)

    Gallo, Robert A.; DeMeo, Patrick J.; Kolman, Brett H.; Daffner, Richard H.; Sciulli, Robert L.; Roberts, Catherine C.

    2004-01-01

    Ruptures of the tibialis anterior tendon are rare. We present the clinical histories and MRI findings of three recent male patients with tibialis anterior tendon rupture aged 58-67 years, all of whom presented with pain over the dorsum of the ankle. Two of the three patients presented with complete rupture showing discontinuity of the tendon, thickening of the retracted portion of the tendon, and excess fluid in the tendon sheath. One patient demonstrated a partial tear showing an attenuated tendon with increased surrounding fluid. Although rupture of the tibialis anterior tendon is a rarely reported entity, MRI is a useful modality in the definitive detection and characterization of tibialis anterior tendon ruptures. (orig.)

  11. Size ratio correlates with intracranial aneurysm rupture status: a prospective study.

    Science.gov (United States)

    Rahman, Maryam; Smietana, Janel; Hauck, Erik; Hoh, Brian; Hopkins, Nick; Siddiqui, Adnan; Levy, Elad I; Meng, Hui; Mocco, J

    2010-05-01

    significantly smaller SRs (2.57 + or - 0.24 mm) compared with the ruptured group (4.08 + or - 0.54 mm; PIA maximum size and SR). Using stepwise selection, only SR remained in the final predictive model (OR, 2.12; 95% CI, 1.09 to 4.13). SR, the ratio between aneurysm size and parent artery diameter, can be easily calculated from 2-dimensional angiograms and correlates with IA rupture status on presentation in a blinded analysis. SR should be further studied in a large prospective observational cohort to predict true IA risk of rupture.

  12. Improved circulating microparticle analysis in acid-citrate dextrose (ACD) anticoagulant tube.

    Science.gov (United States)

    György, Bence; Pálóczi, Krisztina; Kovács, Alexandra; Barabás, Eszter; Bekő, Gabriella; Várnai, Katalin; Pállinger, Éva; Szabó-Taylor, Katalin; Szabó, Tamás G; Kiss, Attila A; Falus, András; Buzás, Edit I

    2014-02-01

    Recently extracellular vesicles (exosomes, microparticles also referred to as microvesicles and apoptotic bodies) have attracted substantial interest as potential biomarkers and therapeutic vehicles. However, analysis of microparticles in biological fluids is confounded by many factors such as the activation of cells in the blood collection tube that leads to in vitro vesiculation. In this study we aimed at identifying an anticoagulant that prevents in vitro vesiculation in blood plasma samples. We compared the levels of platelet microparticles and non-platelet-derived microparticles in platelet-free plasma samples of healthy donors. Platelet-free plasma samples were isolated using different anticoagulant tubes, and were analyzed by flow cytometry and Zymuphen assay. The extent of in vitro vesiculation was compared in citrate and acid-citrate-dextrose (ACD) tubes. Agitation and storage of blood samples at 37 °C for 1 hour induced a strong release of both platelet microparticles and non-platelet-derived microparticles. Strikingly, in vitro vesiculation related to blood sample handling and storage was prevented in samples in ACD tubes. Importantly, microparticle levels elevated in vivo remained detectable in ACD tubes. We propose the general use of the ACD tube instead of other conventional anticoagulant tubes for the assessment of plasma microparticles since it gives a more realistic picture of the in vivo levels of circulating microparticles and does not interfere with downstream protein or RNA analyses. Copyright © 2013 Elsevier Ltd. All rights reserved.

  13. Analysis of reverse flow in inverted U-tubes of steam generator under natural circulation condition

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Ruolei; Liu Jinggong; Qin Shiwei

    2008-01-01

    In this paper, we report on the analysis of reverse flow in inverted U-tubes of a steam generator under natural circulation condition. The mechanism of reverse flow in inverted U-tubes of the steam generator with natural circulation is graphically analyzed by using the full-range characteristic curve of parallel U-tubes. The mathematical model and numerical calculation method for analyzing the reverse flow in inverted U-tubes of the steam generator with natural circulation have been developed. The reverse flow in an inverted U-tube steam generator of a simulated pressurized water reactor with natural circulation in analyzed. Through the calculation, the mass flow rates of normal and reverse flows in individual U-tubes are obtained. The predicted sharp drop of the fluid temperature in the inlet plenum of the steam generator due to reverse flow agrees very well with the experimental data. This indicates that the developed mathematical model and solution method can be used to correctly predict the reverse flow in the inverted U-tubes of the steam generator with natural circulation. The obtained results also show that in the analysis of natural circulation flow in the primary circuit, the reverse flow in the inverted U-tubes of the steam generator must be taken into account. (author)

  14. Rupture, waves and earthquakes.

    Science.gov (United States)

    Uenishi, Koji

    2017-01-01

    Normally, an earthquake is considered as a phenomenon of wave energy radiation by rupture (fracture) of solid Earth. However, the physics of dynamic process around seismic sources, which may play a crucial role in the occurrence of earthquakes and generation of strong waves, has not been fully understood yet. Instead, much of former investigation in seismology evaluated earthquake characteristics in terms of kinematics that does not directly treat such dynamic aspects and usually excludes the influence of high-frequency wave components over 1 Hz. There are countless valuable research outcomes obtained through this kinematics-based approach, but "extraordinary" phenomena that are difficult to be explained by this conventional description have been found, for instance, on the occasion of the 1995 Hyogo-ken Nanbu, Japan, earthquake, and more detailed study on rupture and wave dynamics, namely, possible mechanical characteristics of (1) rupture development around seismic sources, (2) earthquake-induced structural failures and (3) wave interaction that connects rupture (1) and failures (2), would be indispensable.

  15. Flooding experiments with steam and water in a large diameter vertical tube

    International Nuclear Information System (INIS)

    Williams, S.N.; Solom, M.; Draznin, O.; Choutapalli, I.; Vierow, K.

    2009-01-01

    An experimental study on flooding in a large diameter tube is being conducted. In a countercurrent, two-phase flow system, flooding can be defined as the onset of flow reversal of the liquid component which results in cocurrent flow. Flooding can be perceived as a limit to two-phase countercurrent flow, meaning that pairs of liquid and gas flow rates exist that define the envelope for stable countercurrent flow for a given system. Flooding in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA. Analysis of hypothetical severe accidents with current simplified flooding models show that these models represent the largest uncertainty in steam generator tube creep rupture. During a hypothetical station blackout scenario without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. Experiments have been conducted in a 3-inch (76.2 mm) diameter tube with subcooled water and superheated steam as the working fluids at atmospheric pressure. Water flows down the inside of the tube as an annulus while the steam flows upward in the middle. Water flow rates vary from 3.5 to 12 GPM (0.00022 to 0.00076 m 3 /s) and the water inlet temperature is about 70degC. The steam inlet temperature is about 110degC. It was found that a larger steam flow rate was needed to achieve flooding for a lower water flow rate and for a higher water flow rate. This unique data for flooding in steam-water systems in large diameter tubes will reduce uncertainty in flooding models currently utilized in reactor safety codes. (author)

  16. Automated analysis technique developed for detection of ODSCC on the tubes of OPR1000 steam generator

    International Nuclear Information System (INIS)

    Kim, In Chul; Nam, Min Woo

    2013-01-01

    A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.

  17. Intrinsic, Transitional, and Extrinsic Morphological Factors Associated With Rupture of Intracranial Aneurysms.

    Science.gov (United States)

    Ho, Allen L; Lin, Ning; Frerichs, Kai U; Du, Rose

    2015-09-01

    As diagnosis and treatment of unruptured intracranial aneurysms continues to increase, management principles remain largely based on size. This is despite mounting evidence that aneurysm location and other morphologic variables could play a role in predicting overall risk of rupture. Morphological parameters can be divided into 3 main groups, those that are intrinsic to the aneurysm, those that are extrinsic to the aneurysm, and those that involve both the aneurysm and surrounding vasculature (transitional). We present an evaluation of intrinsic, transitional, and extrinsic factors and their association with ruptured aneurysms. Using preoperative computed tomographic angiography, we generated 3-dimensional models of aneurysms and their surrounding vasculature with Slicer software. Using univariate and multivariate analyses, we examined the association of intrinsic, transitional, and extrinsic aspects of aneurysm morphology with rupture. Between 2005 and 2013, 227 cerebral aneurysms in 4 locations were evaluated/treated at a single institution, and computed tomographic angiographies of 218 patients (97 unruptured and 130 ruptured) were analyzed. Ruptured aneurysms analyzed were associated with clinical factors of absence of multiple aneurysms and history of no prior rupture, and morphologic factors of greater aspect ratio. On multivariate analysis, aneurysm rupture remained associated with history of no prior rupture, greater flow angle, greater daughter-daughter vessel angle, and smaller parent-daughter vessel angle. By studying the morphology of aneurysms and their surrounding vasculature, we identified several parameters associated with ruptured aneurysms that include intrinsic, transitional, and extrinsic factors of cerebral aneurysms and their surrounding vasculature.

  18. The Siesta Habit is Associated with a Decreased Risk of Rupture of Intracranial Aneurysms

    Directory of Open Access Journals (Sweden)

    Huibin Kang

    2017-09-01

    Full Text Available BackgroundPrevious studies have examined an association between the siesta habit and hypertension, as well as coronary heart disease. However, the relationship between a siesta and the risk of rupture of an intracranial aneurysm (IA has not yet been established. We aimed to investigate the effects of a siesta on the risk of rupture of IAs.MethodsWe prospectively enrolled consecutive patients diagnosed with IAs at our hospital between January 2016 and December 2016. Univariate and multivariate logistic regression analysis were performed to identify independent risk factors associated with IA rupture.ResultsWe studied 581 consecutive patients with 514 unruptured and 120 ruptured aneurysms. Univariate analysis demonstrated that hypertension, hyperlipidemia, diabetes mellitus, cigarette smoking, location, size, as well as shape and aspect ratio were associated with the risk of rupture of IAs. Multivariate analysis identified hypertension [odds ratio (OR 1.68, 95% confidence interval (CI 1.03–2.73], hyperlipidemia (OR 0.25, 95% CI 0.08–0.72, current cigarette smoking ≥20 cigarettes/day (d (OR 3.48, 95% CI 1.63–7.47, siesta (siesta time <1 h, OR 0.49, 95% CI 0.24–0.98 and siesta time ≥1 h, OR 0.32, 95% CI 0.19–0.57, location of largest aneurysm on the anterior communicating and internal carotid-posterior communicating artery (PCOM (anterior communicating artery OR 16.27, 95% CI 7.40–35.79 and PCOM OR 11.21, 95% CI 5.15–24.43, and size of aneurysm ≥7 mm (OR 2.19, 95% CI 1.21–3.97 as independent strong risk factors associated with risk of aneurysm rupture.ConclusionIn the present study, we found that a habitual siesta is a new predictive factor to assess the risk of rupture of an IA. We found the siesta habit may reduce the risk of aneurysm rupture. We also found that hypertension, hyperlipidemia, cigarette smoking, location, and size of aneurysm were associated with the risk of rupture of IAs.

  19. Slow rupture of frictional interfaces

    OpenAIRE

    Sinai, Yohai Bar; Brener, Efim A.; Bouchbinder, Eran

    2011-01-01

    The failure of frictional interfaces and the spatiotemporal structures that accompany it are central to a wide range of geophysical, physical and engineering systems. Recent geophysical and laboratory observations indicated that interfacial failure can be mediated by slow slip rupture phenomena which are distinct from ordinary, earthquake-like, fast rupture. These discoveries have influenced the way we think about frictional motion, yet the nature and properties of slow rupture are not comple...

  20. Depth-Sizing Technique for Crack Indications in Steam Generator Tubing

    International Nuclear Information System (INIS)

    Cho, Chan Hee; Lee, Hee Jeong; Kim, Hong Deok

    2009-01-01

    The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program

  1. Management of Labor and Delivery After Fetoscopic Repair of an Open Neural Tube Defect.

    Science.gov (United States)

    Kohn, Jaden R; Rao, Vibha; Sellner, Allison A; Sharhan, Dina; Espinoza, Jimmy; Shamshirsaz, Alireza A; Whitehead, William E; Belfort, Michael A; Sanz Cortes, Magdalena

    2018-06-01

    To report labor, delivery, and neonatal outcomes in a cohort of women delivering neonates who had undergone fetoscopic neural tube defect repair. We conducted a retrospective cohort study from April 2014 to January 2018. All patients met Management of Myelomeningocele Study eligibility criteria. We included patients with completed second-trimester fetoscopic neural tube defect repair (laparotomy, uterine exteriorization, and minimally invasive access through two or three uterine ports) followed by standardized management of labor and delivery at our institution. Outcomes included rates of vaginal delivery, term delivery, and intrapartum cesarean delivery as well as obstetric and neonatal outcomes after oxytocin. Complications of interest included preterm prelabor rupture of membranes, chorioamnionitis, uterine dehiscence or rupture, 5-minute Apgar score less than 7, and neonatal acidosis (umbilical artery pH less than 7.15). Thirty-four patients had fetoscopic repair, followed by 17 vaginal deliveries (50%, 95% CI 32-68%). Median gestational age was 38 1/7 weeks at vaginal delivery (range 26 0/7-40 2/7 weeks of gestation) and 37 1/7 weeks of gestation at cesarean delivery (range 25 5/7-40 5/7 weeks of gestation); 62% of deliveries occurred at term. Eight patients had prelabor cesarean delivery: three nonurgent and five urgent (for nonreassuring fetal heart tracings). Twenty-six patients labored; six were induced and 20 labored spontaneously. Of the latter, five were augmented. Of 26 laboring patients, 17 delivered vaginally and nine underwent urgent cesarean delivery (35%, 95% CI 17-56%; seven nonreassuring fetal heart tracings and two breech). There were no cases of uterine rupture or dehiscence. Most (94%, 95% CI 80-99%) had normal 5-minute Apgar scores; one neonate (3%, 95% CI 0-15%) had acidosis but normal Apgar scores. Our data regarding trial of labor, use of low-dose oxytocin, and vaginal delivery after prenatal fetoscopic neural tube defect repair are

  2. Erythromycin for Promoting the Postpyloric Placement of Feeding Tubes: A Systematic Review and Meta-Analysis

    Directory of Open Access Journals (Sweden)

    Qing-Jun Jiang

    2018-01-01

    Full Text Available Background. Critically ill patients can benefit from enteral nutrition with postpyloric feeding tubes, but the low success rate limits its wide use. Erythromycin could elevate the success rate of tube insertion, but its clinical efficiency still remains controversial. Methods. Included studies must be RCTs which assessed the success rate of postpyloric feeding tube insertion using erythromycin. Results. 284 patients were enrolled in six studies. Meta-analysis showed that erythromycin significantly increases the rate of successful postpyloric feeding tube placement (RR 1.45, 95% CI (1.12, 1.86 and did not increase the risk of adverse effects (RR 2.15, 95% CI (0.20, 22.82. Subgroup analysis showed that unweighted feeding tubes (RR 1.47, 95% CI (1.03, 2.11 could significantly increase the success rate. Country of study, intravenous route of erythromycin, and year of participant enrollment did not influence these results. Conclusions. Erythromycin significantly increases the success rate of postpyloric feeding tube placement. This suggests that erythromycin can be used as an auxiliary method to improve the success rate of bedside insertion.

  3. Aortic ruptures in seat belt wearers.

    Science.gov (United States)

    Arajärvi, E; Santavirta, S; Tolonen, J

    1989-09-01

    Several investigations have indicated that rupture of the thoracic aorta is one of the leading causes of immediate death in victims of road traffic accidents. In Finland in 1983, 92% of front-seat passengers were seat belt wearers on highways and 82% in build-up areas. The mechanisms of rupture of the aorta have been intensively investigated, but the relationship between seat belt wearing and injury mechanisms leading to aortic rupture is still largely unknown. This study comprises 4169 fatally injured victims investigated by the Boards of Traffic Accident Investigation of Insurance Companies during the period 1972 to 1985. Chest injuries were recorded as the main cause of death in 1121 (26.9%) victims, 207 (5.0%) of those victims having worn a seat belt. Aortic ruptures were found at autopsy in 98 victims and the exact information of the location of the aortic tears was available in 68. For a control group, we analyzed 72 randomly chosen unbelted victims who had a fatal aortic rupture in similar accidents. The location of the aortic rupture in unbelted victims was more often in the ascending aorta, especially in drivers, whereas in seat belt wearers the distal descending aorta was statistically more often ruptured, especially in right-front passengers (p less than 0.05). The steering wheel predominated statistically as the part of the car estimated to have caused the injury in unbelted victims (37/72), and some interior part of the car was the most common cause of fatal thoracic impacts in seat belt wearers (48/68) (p less than 0.001). The mechanism of rupture of the aorta in the classic site just distal to the subclavian artery seems to be rapid deceleration, although complex body movements are also responsible in side impact collisions. The main mechanism leading to rupture of the ascending aorta seems to be severe blow to the bony thorax. This also often causes associated thoracic injuries, such as heart rupture and sternal fracture. Injuries in the ascending

  4. Morphological and clinical risk factors for posterior communicating artery aneurysm rupture.

    Science.gov (United States)

    Matsukawa, Hidetoshi; Fujii, Motoharu; Akaike, Gensuke; Uemura, Akihiro; Takahashi, Osamu; Niimi, Yasunari; Shinoda, Masaki

    2014-01-01

    Recent studies have shown that posterior circulation aneurysms, specifically posterior communicating artery (PCoA) aneurysms, are more likely to rupture than other aneurysms. To date, few studies have investigated the factors contributing to PCoA aneurysm rupture. The authors aimed to identify morphological and clinical characteristics predisposing to PCoA aneurysm rupture. The authors retrospectively reviewed 134 consecutive patients with PCoA aneurysms managed at their facility between July 2003 and December 2012. The authors divided patients into groups of those with aneurysmal rupture (n = 39) and without aneurysmal rupture (n = 95) and compared morphological and clinical characteristics. Morphological characteristics were mainly evaluated by 3D CT angiography and included diameter of arteries (anterior cerebral artery, middle cerebral artery, and internal carotid artery), size of the aneurysm, dome-to-neck ratio, neck direction of the aneurysmal dome around the PCoA (medial, lateral, superior, inferior, and posterior), aneurysm bleb formation, whether the PCoA was fetal type, and the existence of other intracranial unruptured aneurysm(s). Patients with ruptured PCoA aneurysms were significantly younger (a higher proportion were PCoA aneurysms showed a lateral direction of the aneurysmal dome around the PCoA, had bleb formation, and the aneurysm was > 7 mm in diameter and/or the dome-to-neck ratio was > 2.0. Multivariate logistic regression analysis showed age PCoA (OR 6.7, p = 0.0001), and bleb formation (OR 11, p PCoA aneurysm rupture. The present results demonstrated that lateral projection of a PCoA aneurysm may be related to rupture.

  5. The development and application of overheating failure model of FBR steam generator tubes. 3

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi

    2002-03-01

    The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: 1. To evaluate the structural integrity of tube material, the strength standard for 2. 25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200degC) creep data. This standard has been validated with the tube rupture simulation test data. 2. The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. 3. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. 4. The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. 5. The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system. (author)

  6. Thermal-hydraulic analysis of Ignalina NPP compartments response to group distribution header rupture using RALOC4 code

    International Nuclear Information System (INIS)

    Urbonavicius, E.

    2000-01-01

    The Accident Localisation System (ALS) of Ignalina NPP is a containment of pressure suppression type designed to protect the environment from the dangerous impact of the radioactivity. The failure of ALS could lead to contamination of the environment and prescribed public radiation doses could be exceeded. The purpose of the presented analysis is to perform long term thermal-hydraulic analysis of compartments response to Group Distribution Header rupture and verify if design pressure values are not exceeded. (authors)

  7. Simulation and analysis of the thermal and deformation behaviour of `as-received` and `hydrided` pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

    Energy Technology Data Exchange (ETDEWEB)

    Muir, W C; Bayoumi, M H [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 {mu}g/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs.

  8. Simulation and analysis of the thermal and deformation behaviour of 'as-received' and 'hydrided' pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

    International Nuclear Information System (INIS)

    Muir, W.C.; Bayoumi, M.H.

    1995-01-01

    It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 μg/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs

  9. Ruptured cornual pregnancy

    International Nuclear Information System (INIS)

    Hussain, M.; Yasmeen, H.; Noorani, K.

    2003-01-01

    A case of ruptured cornual pregnancy is presented here. The patient presented with history of 30 weeks gestational amenorrhoea and pain in the lower abdomen and epigastrium for the last seven days. Ultrasound revealed a 29 weeks abdominal pregnancy with blood in the pelvic cavity. On laparotomy; there was a ruptured right cornual pregnancy, treated cornual resection and uterine repair. An alive male baby of one kg weight was delivered from the resected cornua of the uterus. (author)

  10. Fracture analysis of tube boiler for physical explosion accident.

    Science.gov (United States)

    Kim, Eui Soo

    2017-09-01

    Material and failure analysis techniques are key tools for determining causation in case of explosive and bursting accident result from material and process defect of product in the field of forensic science. The boiler rupture generated by defect of the welding division, corrosion, overheating and degradation of the material have devastating power. If weak division of boiler burner is fractured by internal pressure, saturated vapor and water is vaporized suddenly. At that time, volume of the saturated vapor and water increases up to thousands of volume. This failure of boiler burner can lead to a fatal disaster. In order to prevent an explosion and of the boiler, it is critical to introduce a systematic investigation and prevention measures in advance. In this research, the cause of boiler failure is investigated through forensic engineering method. Specifically, the failure mechanism will be identified by fractography using scanning electron microscopes (SEM) and Optical Microscopes (OM) and mechanical characterizations. This paper presents a failure analysis of household welding joints for the water tank of a household boiler burner. Visual inspection was performed to find out the characteristics of the fracture of the as-received material. Also, the micro-structural changes such as grain growth and carbide coarsening were examined by optical microscope. Detailed studies of fracture surfaces were made to find out the crack propagation on the weld joint of a boiler burner. It was concluded that the rupture may be caused by overheating induced by insufficient water on the boiler, and it could be accelerated by the metal temperature increase. Copyright © 2017 Elsevier B.V. All rights reserved.

  11. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  12. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  13. YouTube and 'psychiatry'.

    Science.gov (United States)

    Gordon, Robert; Miller, John; Collins, Noel

    2015-12-01

    YouTube is a video-sharing website that is increasingly used to share and disseminate health-related information, particularly among younger people. There are reports that social media sites, such as YouTube, are being used to communicate an anti-psychiatry message but this has never been confirmed in any published analysis of YouTube clip content. This descriptive study revealed that the representation of 'psychiatry' during summer 2012 was predominantly negative. A subsequent smaller re-analysis suggests that the negative portrayal of 'psychiatry' on YouTube is a stable phenomenon. The significance of this and how it could be addressed are discussed.

  14. An analysis of uterine rupture at the Nnamdi Azikiwe University ...

    African Journals Online (AJOL)

    Materials and Methods: This descriptive case series was conducted at the department of Obstetrics and Gynaecology, Nnamdi Azikiwe, University Teaching Hospital Nnewi from March 2004 to February 2009. Results: The incidence of uterine rupture was 6.2 per 1000 deliveries. The commonest age range of occurrence ...

  15. Ruptured corpus luteal cyst: Prediction of clinical outcomes with CT

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Seok; Moon, Min Hoan; Woo, Hyun Sik; Sung, Chang Kyu; Jeon, Hye Won; Lee, Taek Sang [SMG-SNU Boramae Medical Center, Seoul National University College of Medicine, Seoul (Korea, Republic of)

    2017-08-01

    To evaluate the determinant pretreatment CT findings that can predict surgical intervention for patients suffering from corpus luteal cyst rupture with hemoperitoneum. From January 2009 to December 2014, a total of 106 female patients (mean age, 26.1 years; range, 17–44 years) who visited the emergency room of our institute for acute abdominal pain and were subsequently diagnosed with ruptured corpus luteal cyst with hemoperitoneum were included in the retrospective study. The analysis of CT findings included cyst size, cyst shape, sentinel clot sign, ring of fire sign, hemoperitoneum depth, active bleeding in portal phase and attenuation of hemoperitoneum. The comparison of CT findings between the surgery and conservative management groups was performed with the Mann-Whitney U test or chi-square test. Logistic regression analysis was used to determine significant CT findings in predicting surgical intervention for a ruptured cyst. Comparative analysis revealed that the presence of active bleeding and the hemoperitoneum depth were significantly different between the surgery and conservative management groups and were confirmed as significant CT findings for predicting surgery, with adjusted odds ratio (ORs) of 3.773 and 1.318, respectively (p < 0.01). On the receiver-operating characteristic curve analysis for hemoperitoneum depth, the optimal cut-off value was 5.8 cm with 73.7% sensitivity and 58.6% specificity (Az = 0.711, p = 0.004). In cases with a hemoperitoneum depth > 5.8 cm and concurrent active bleeding, the OR for surgery increased to 5.786. The presence of active bleeding and the hemoperitoneum depth on a pretreatment CT scan can be predictive warning signs of surgery for a patient with a ruptured corpus luteal cyst with hemoperitoneum.

  16. Categorising YouTube

    OpenAIRE

    Simonsen, Thomas Mosebo

    2011-01-01

    This article provides a genre analytical approach to creating a typology of the User Generated Content (UGC) of YouTube. The article investigates the construction of navigation processes on the YouTube website. It suggests a pragmatic genre approach that is expanded through a focus on YouTube’s technological affordances. Through an analysis of the different pragmatic contexts of YouTube, it is argued that a taxonomic understanding of YouTube must be analysed in regards to the vacillation of a...

  17. Global variations of large megathrust earthquake rupture characteristics

    Science.gov (United States)

    Kanamori, Hiroo

    2018-01-01

    Despite the surge of great earthquakes along subduction zones over the last decade and advances in observations and analysis techniques, it remains unclear whether earthquake complexity is primarily controlled by persistent fault properties or by dynamics of the failure process. We introduce the radiated energy enhancement factor (REEF), given by the ratio of an event’s directly measured radiated energy to the calculated minimum radiated energy for a source with the same seismic moment and duration, to quantify the rupture complexity. The REEF measurements for 119 large [moment magnitude (Mw) 7.0 to 9.2] megathrust earthquakes distributed globally show marked systematic regional patterns, suggesting that the rupture complexity is strongly influenced by persistent geological factors. We characterize this as the existence of smooth and rough rupture patches with varying interpatch separation, along with failure dynamics producing triggering interactions that augment the regional influences on large events. We present an improved asperity scenario incorporating both effects and categorize global subduction zones and great earthquakes based on their REEF values and slip patterns. Giant earthquakes rupturing over several hundred kilometers can occur in regions with low-REEF patches and small interpatch spacing, such as for the 1960 Chile, 1964 Alaska, and 2011 Tohoku earthquakes, or in regions with high-REEF patches and large interpatch spacing as in the case for the 2004 Sumatra and 1906 Ecuador-Colombia earthquakes. Thus, combining seismic magnitude Mw and REEF, we provide a quantitative framework to better represent the span of rupture characteristics of great earthquakes and to understand global seismicity. PMID:29750186

  18. Triple Achilles Tendon Rupture: Case Report.

    Science.gov (United States)

    Saxena, Amol; Hofer, Deann

    We present a case report with 1-year follow-up data of a 57-year-old male soccer referee who had sustained an acute triple Achilles tendon rupture injury during a game. His triple Achilles tendon rupture consisted of a rupture of the proximal watershed region, a rupture of the main body (mid-watershed area), and an avulsion-type rupture of insertional calcific tendinosis. The patient was treated surgically with primary repair of the tendon, including tenodesis with anchors. Postoperative treatment included non-weightbearing for 4 weeks and protected weightbearing until 10 weeks postoperative, followed by formal physical therapy, which incorporated an "antigravity" treadmill. The patient was able to return to full activity after 26 weeks, including running and refereeing, without limitations. Copyright © 2017 The American College of Foot and Ankle Surgeons. Published by Elsevier Inc. All rights reserved.

  19. Rupture disc opening property for using pipe rupture test in JAERI

    International Nuclear Information System (INIS)

    Kato, Rokuro

    1983-03-01

    In the Mechanical Strength and Structure Lab of JAERI there are being performed pipe break tests which are a postulated instantaneous guillotine break of the primary coolant piping in nuclear power plants. The test being performed are pipe whip tests and jet discharging tests. The bursting of the rupture disc is initiated by an electrical arc and is concluded by the internal pressure. Because the time characteristics during the opening of the rupture disc affects the dynamic thrust force of the pipe, it is necessary to measure these time characteristics. However, it is difficult to measure the conditions during this continuous opening because at the same time of the opening the high temperature and high pressure water is flashing. Therefore, the rupture disc opening was postulated on the measuring of the effective opening characteristics with electric contraction terminals which were attached to the inner surface of the test pipe downstream of the rupture disc and were extended toward the pipe centerline in a ring whose area is about 60 % of the area of the pipe flow sectional area. The measurement voltage was recorded when the data recorder was started in sequence with the electrical arc release from a trigger signal. As a result, it is evident that under high temperature and high pressure water the effective opening time is delayed by a few milliseconds. (author)

  20. Diagnosis of Complex Pulley Ruptures Using Ultrasound in Cadaver Models.

    Science.gov (United States)

    Schöffl, Isabelle; Hugel, Arnica; Schöffl, Volker; Rascher, Wolfgang; Jüngert, Jörg

    2017-03-01

    Pulley ruptures are common in climbing athletes. The purposes of this study were to determine the specific positioning of each pulley with regards to the joint, and to evaluate the ultrasound diagnostics of various pulley rupture combinations. For this, 34 cadaver fingers were analyzed via ultrasound, the results of which were compared to anatomic measurements. Different pulley ruptures were then simulated and evaluated using ultrasound in standardized dynamic forced flexion. Visualization of the A2 and A4 pulleys was achieved 100% of the time, while the A3 pulley was visible in 74% of cases. Similarly, injuries to the A2 and A4 pulleys were readily observable, while A3 pulley injuries were more challenging to identify (sensitivity of 0.2 for singular A3 pulley, 0.5 for A2/A4 pulley and 0.33 for A3/A4 pulley ruptures). Receiver operating characteristic analysis was used to evaluate the optimal tendon-bone distance for pulley rupture diagnosis, a threshold which was determined to be 1.9 mm for A2 pulley ruptures and 1.85 for A4 pulley ruptures. This study was the first to carry out a cadaver ultrasound examination of a wide variety of pulley ruptures. Ultrasound is a highly accurate tool for visualizing the A2 and A4 pulleys in a cadaver model. This method of pathology diagnosis was determined to be suitable for injuries to the A2 and A4 pulleys, but inadequate for A3 pulley injuries. Copyright © 2016 World Federation for Ultrasound in Medicine & Biology. Published by Elsevier Inc. All rights reserved.

  1. Influence of microstructure modification on the circumferential creep of Zr–Nb–Sn–Fe cladding tubes

    International Nuclear Information System (INIS)

    Jeong, Gu Beom; Kim, In Won; Hong, Sun Ig

    2016-01-01

    Out-of-reactor, non-irradiated thermal creep performances and lives of annealed and stress-relieved Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were studied and compared. The creep rates of annealed Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were appreciably slower than those of stress-relieved annealed counterpart. The stress exponent increased slightly from 5.1 to 6.1 in the stress-relieved cladding to 5.3–6.3 in the annealed cladding. The creep activation energy of the annealed Zr-1.02Nb-0.69Sn-0.12Fe alloy (300–330 kJ/mol) was larger compared to that of the stress-relieved alloy (210–260 kJ/mol). The creep activation energy of annealed alloy is close to that of self-diffusion in α-Zr (336 kJ/mol). The smaller activation energy in the stress-relieved alloy is attributed to the increasing contribution of faster diffusion path such as grain boundaries and dislocations. The presence of dislocation arrays with higher dislocation density and smaller grain size in the stress-relived alloy was confirmed by TEM analysis. The creep rupture time increased dramatically in the annealed Zr–1Nb- 0.7Sn-0.1Fe alloy compared to that of stress-relieved alloy, supporting the decrease of creep rate by annealing. The creep life of Zr-1.02Nb-0.69Sn-0.12Fe claddings can be extended through microstructure modification by annealing at intermediate temperatures in which dislocation creep dominates. - Highlights: • Effect of microstructure modification on creep in Zr–Nb–Sn–Fe tubes was studied. • Creep activation energy in annealed tubes was larger than in stress-relieved tubes. • Lower dislocation density in lager grains was observed after creep in annealed tubes. • Larson–Miller parameter of annealed tube was larger than that of stress-relieved one. • Creep life of tubes was extended through microstructure modification by annealing.

  2. Splenic rupture masquerading ruptured ectopic pregnancy | Kigbu ...

    African Journals Online (AJOL)

    The classical triad of presentation of delayed menses, irregular vaginal bleeding and abdominal pain may not be encountered at all! Overwhelming features of abdominal pain, amenorrhea, pallor, abdominal tenderness, shifting dullness with positive pregnancy test gave a clinical diagnosis of ruptured ectopic pregnancy.

  3. Labor Dystocia and the Risk of Uterine Rupture in Women with Prior Cesarean.

    Science.gov (United States)

    Vachon-Marceau, Chantale; Demers, Suzanne; Goyet, Martine; Gauthier, Robert; Roberge, Stéphanie; Chaillet, Nils; Laroche, Jasmin; Bujold, Emmanuel

    2016-05-01

    Objective The objective of this study was to evaluate the association between labor dystocia and uterine rupture. Methods We performed a secondary analysis of a multicenter case-control study that included women with single, prior, low-transverse cesarean section who experienced complete uterine rupture during a trial of labor (TOL). For each case, three women who underwent a TOL without uterine rupture were selected as controls. Data were collected on cervical dilatations from admission to delivery. We evaluated the relationship between uterine rupture and labor dystocia according to several criteria, including the World Health Organization's (WHO's) partogram. Results Data were available for 90 cases and 260 controls. Compared with the controls, uterine rupture was associated with less cervical dilatation on admission, slower cervical dilatation in the first stage of labor and longer second stage of labor (all with p dystocia is a significant risk factor for uterine rupture. Labor progression should be assessed regularly in women with prior cesarean. Thieme Medical Publishers 333 Seventh Avenue, New York, NY 10001, USA.

  4. MRI findings of achilles tendon rupture

    International Nuclear Information System (INIS)

    Zhang Xuezhe

    2009-01-01

    Objective: To evaluate the MRI findings of achilles tendon rupture. Methods: The MRI data of 7 patients with achilles tendon rupture were retrospectively analysed. All 7 patients were male with the age ranging from 34 to 71 years. Routine MR scanning was performed in axial and sagittal planes, including T 1 WI, T 2 WI and a fat suppression MRI (SPIR). Results: Among 7 patients, complete achilles tendon rupture was seen in 6 cases, partial achilles tendon rupture 1 case. The site of tendon disruption were 2.6-11.0 cm( mean 5.4 cm) proximal to the insertion in the calcaneus. The MRI findings of a partial or complete rupture of the achilles tendon included enlarged and thickened achilles tendon (7 cases), wavy lax achilles tendon (2 cases), discontinuity of some or all of its fibers and intratendinous regions of increased signal intensity (7 cases). In the cases of complete tendon rupture, the size of the tendinous gap varied from 3.0-8.0 mm, which was filled with blood and appeared as edema of increase signal intensity on T 2 WI and SPIR. In all 7 patients, MR scanning showed medium signal intensity (7 cases) on T 1 WI, or medium signal intensity (1 cases), medium-high signal intensity (3 cases ), high signal intensity (3 cases) on T 2 WI, and medium-high signal intensity (2 cases), high signal intensity (5 cases) on fat suppression MRI. The preachilles fat pad showed obscure in 6 cases of complete achilles tendon rupture. Conclusion: MRI is an excellent method for revealing achilles tendon rupture and confirming the diagnosis. (authors)

  5. Fatigue and rupture codified rules comparison

    International Nuclear Information System (INIS)

    Faidy, C.

    2004-01-01

    The European Directive on Pressure Equipment requests risk studies and in particular to assure no risk of fatigue and rupture in operation. The answers to these questions are different in the different existing design codes (EN Standards, ASME III and VIII or RCC-M or CODAP-CODETI codes) and corresponding in operation codes (ASME or RSE-M). Design safety factors, material properties, fabrication, refinement in the analysis methods, monitoring in operation, hydro-proof test level... Around these Codes, different rules are under development. A16 in France, R6 in UK or FITNET at the EC level. This paper is concerned by a comparison between 2 different Codes to analyze the risk of fatigue or rupture of pressure equipments and mainly a comparison between RCC-M Code and EN 13445 standard for pressure vessel. Recommendations for future work will be proposed. (authors)

  6. Science on TeacherTube: A Mixed Methods Analysis of Teacher Produced Video

    Science.gov (United States)

    Chmiel, Margaret (Marjee)

    Increased bandwidth, inexpensive video cameras and easy-to-use video editing software have made social media sites featuring user generated video (UGV) an increasingly popular vehicle for online communication. As such, UGV have come to play a role in education, both formal and informal, but there has been little research on this topic in scholarly literature. In this mixed-methods study, a content and discourse analysis are used to describe the most successful UGV in the science channel of an education-focused site called TeacherTube. The analysis finds that state achievement tests, and their focus on vocabulary and recall-level knowledge, drive much of the content found on TeacherTube.

  7. A statistical method for draft tube pressure pulsation analysis

    International Nuclear Information System (INIS)

    Doerfler, P K; Ruchonnet, N

    2012-01-01

    Draft tube pressure pulsation (DTPP) in Francis turbines is composed of various components originating from different physical phenomena. These components may be separated because they differ by their spatial relationships and by their propagation mechanism. The first step for such an analysis was to distinguish between so-called synchronous and asynchronous pulsations; only approximately periodic phenomena could be described in this manner. However, less regular pulsations are always present, and these become important when turbines have to operate in the far off-design range, in particular at very low load. The statistical method described here permits to separate the stochastic (random) component from the two traditional 'regular' components. It works in connection with the standard technique of model testing with several pressure signals measured in draft tube cone. The difference between the individual signals and the averaged pressure signal, together with the coherence between the individual pressure signals is used for analysis. An example reveals that a generalized, non-periodic version of the asynchronous pulsation is important at low load.

  8. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

  9. Transient megaoesophagus and oesophagitis following diaphragmatic rupture repair in a cat.

    Science.gov (United States)

    Joseph, Rotem; Kuzi, Sharon; Lavy, Eran; Aroch, Itamar

    2008-07-01

    A 6-month-old domestic shorthair female cat was presented with suspected diaphragmatic hernia (DH) that was later confirmed by thoracic radiography. The cat underwent exploratory celiotomy with a diaphragmatic rupture (DR) repair and recovered. Six days later, it was represented with vomiting and anorexia. Megaoesophagus (MO) and gastric dilatation were diagnosed by contrast radiography. A second celiotomy revealed no abnormalities and gastropexy was performed. Endoscopy demonstrated MO, oesophagitis and gastro-oesophageal reflux. MO persisted for several weeks and was an unexpected complication as no association between DR (or DH) and MO has never been described in the veterinary literature. The cat was treated medically with aggressive prokinetic and antacid therapy along with prolonged temporary oesophageal diversion (percutaneous endoscopic gastrostomy tube) with an excellent outcome.

  10. Case report: ruptured aortic aneurysm into oesophagus - treatment by covered stent

    International Nuclear Information System (INIS)

    Young, N.; Tan, I.; Costa, R.

    2002-01-01

    Full text: A case report of a 70 year old man acutely presenting with a large descending thoracic aortic aneurysm, ruptured into the adjacent oesophagus. He was treated with an Aneurex covered stent (Medtronics). This patient has a long history of ischaemic heart disease, hypertension and previously treated abdominal aortic aneurysm. After diagnosis by CT, the bleeding was emergency temponaded by insertion of a Sengstaken-Blackmore tube into the oesophagus and inflated to arterial pressure. After stabilisation in ICU, it was decided that open surgery would probably prove fatal, and insertion of a covered thoracic aortic stent was performed in theatre with a femoral artery cut-down. Post-operatively the bleeding from the aortic aneurysm ceased. However, the patient has ongoing problems with large, open oesophageal ulcer with chronic ooze requiring intermittent transfusions, chronic infection with MRSA, requiring long term antibiotics, feeding via feeding gastrostomy tubes. The patient is stable 12 months after presentation. Treatment of this otherwise fatal problem by covered stent has permitted survival benefit. However, there are significant, unresolved issues of oesophageal ulcer and ongoing MRSA infection. Copyright (2002) Blackwell Science Pty Ltd

  11. A LOCA analysis for AHWR caused by ECCS header rupture

    International Nuclear Information System (INIS)

    Chatterjee, B.; Gawai, Amol; Gupta, S.K.; Kushwaha, H.S.

    2000-01-01

    Loss of coolant accident (LOCA) analyses for the proposed 750 MWth Advanced Heavy Water Reactor (AHWR), initiated by the rupture of 8 inch NB ECCS header has been carried out. This paper narrates the description of AHWR and associated ECCS, postulated scenario with which the analyses is carried out, results, discussion and conclusion

  12. Plastic collapse and energy absorption of circular filled tubes under quasi-static loads by computational analysis

    Energy Technology Data Exchange (ETDEWEB)

    Beng, Yeo Kiam; Tzeng, Woo Wen [Universiti Malaysia Sabah, Sabah (Malaysia)

    2017-02-15

    This study presents the finite element analysis of plastic collapse and energy absorption of polyurethane-filled aluminium circular tubes under quasi-static transverse loading. Increasing focuses were given to impact damage of structures where energy absorbed during impact could be controlled to avoid total structure collapse of energy absorbers and devices designed to dissipate energy. ABAQUS finite element analysis application was utilized for modelling and simulating the polyurethane-filled aluminium tubes, different set of diameterto- thickness ratios and span lengths, subjected to transverse three-point-bending load. Different sets of polyurethane-filled aluminium tubes subjected to the transverse loading were modelled and simulated. The failure modes and mechanisms of filled tubes and its capabilities as energy absorbers to further improve and strengthening of empty tube were also identified. The results showed that plastic deformation response was affected by the geometric constraints and parameters of the specimens. The diameter-to-thickness ratio and span lengths had shown to play crucial role in optimizing the PU-filled tube as energy absorber.

  13. 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.)

  14. Liver Hydatid Cyst with Transdiaphragmatic Rupture and Lung Hydatid Cyst Ruptured into Bronchi and Pleural Space

    International Nuclear Information System (INIS)

    Arıbaş, Bilgin Kadri; Dingil, Gürbüz; Köroğlu, Mert; Üngül, Ümit; Zaralı, Aliye Ceylan

    2011-01-01

    The aim of this case study is to present effectiveness of percutaneous drainage as a treatment option of ruptured lung and liver hydatid cysts. A 65-year-old male patient was admitted with complicated liver and lung hydatid cysts. A liver hydatid cyst had ruptured transdiaphragmatically, and a lung hydatid cyst had ruptured both into bronchi and pleural space. The patient could not undergo surgery because of decreased respiratory function. Both cysts were drained percutaneously using oral albendazole. Povidone–iodine was used to treat the liver cyst after closure of the diaphragmatic rupture. The drainage was considered successful, and the patient had no recurrence of signs and symptoms. Clinical, laboratory, and radiologic recovery was observed during 2.5 months of catheterization. The patient was asymptomatic after catheter drainage. No recurrence was detected during 86 months of follow-up. For inoperable patients with ruptured liver and lung hydatid cysts, percutaneous drainage with oral albendazole is an alternative treatment option to surgery. The percutaneous approach can be life-saving in such cases.

  15. Wavelet time-frequency analysis of accelerating and decelerating flows in a tube bank

    International Nuclear Information System (INIS)

    Indrusiak, M.L.S.; Goulart, J.V.; Olinto, C.R.; Moeller, S.V.

    2005-01-01

    In the present work, the steady approximation for accelerating and decelerating flows through tube banks is discussed. With this purpose, the experimental study of velocity and pressure fluctuations of transient turbulent cross-flow in a tube bank with square arrangement and a pitch-to-diameter ratio of 1.26 is performed. The Reynolds number at steady-state flow, computed with the tube diameter and the flow velocity in the narrow gap between the tubes, is 8 x 10 4 . Air is the working fluid. The accelerating and decelerating transients are obtained by means of start and stop of the centrifugal blower. Wavelet and wavelet packet multiresolution analysis were applied to decompose the signal in frequency intervals, using Daubechies 20 wavelet and scale functions, thus allowing the analysis of phenomena in a time-frequency domain. The continuous wavelet transform was also applied, using the Morlet function. The signals in the steady state, which presented a bistable behavior, were separated in two modes and analyzed with usual statistic tools. The results were compared with the steady-state assumption, demonstrating the ability of wavelets for analyzing time varying signals

  16. Finite Element Modeling of Dieless Tube Drawing of Strain Rate Sensitive Material with Coupled Thermo-Mechanical Analysis

    Science.gov (United States)

    Furushima, Tsuyoshi; Sakai, Takashi; Manabe, Ken-ichi

    2004-06-01

    Dieless drawing is a unique deformation process without conventional dies, which can achieve a great reduction of wire and tube metals in single pass by means of local heating and cooling approach. In this study, for microtube forming, the dieless drawing process applying superplastic behavior was analyzed by finite element method (FEM) in order to clarify the effect of dieless tube drawing conditions such as tensile speed, moving speed of heating and cooling system, and material properties on deformation behavior of the tube. In the calculation, the material properties were dealt in a special subroutine, whose constitutive equation was defined as σ = Kɛnɛ˙m, and was linked to the solver. A coupled thermo-mechanical analysis was performed for the dieless tube drawing using the FEM. In the thermal analysis of dieless tube drawing, heat transfer was introduced to calculate the heat flux between heating coil and tube surface, and heat conduction in a tube. The influence of dieless tube drawing conditions on deformation behavior was clarified. As a result, for the strain rate sensitive material, the maximum reduction of area and the minimum outer diameter in single pass attain to 90.9% and 2.56mm, respectively. From the result, it is concluded that the dieless tube drawing is essential to produce an extrafine microtube by reason of keeping cylindrical tube diameter ratio constant with extremely high reduction.

  17. Application of the visual system analyzer (ViSA): simulation of the steam generator tube rupture event at Ulchin unit 4

    International Nuclear Information System (INIS)

    Lee, S.W.; Kim, K.D.; Hwang, M.K.; Jeong, J.J.

    2004-01-01

    Korea Atomic Energy Research Institute (KAERI) has developed the Visual System Analyzer (ViSA) based on the best-estimate (B-E) codes, MARS and RETRAN-3D. The key features of ViSA are: (1) The use of the same input and the same level of accuracy as the original codes is guaranteed (2) Users can design their own plant mimic by a drag-and-drop from the provided indicators (3) The on-line interactive control enables users to simulate the operator's actions (4) The nodalization window is designed to display the transient temperature and void distributions. ViSA is composed of two parts; the B-E code with plant input and the Graphic User Interface (GUI) that includes the plant mimic and an interactive control function, etc. The calculation results of the B-E code are transferred to a user via the GUI and a user can apply the operator action to the B-E code using an interactive control function. Therefore, it is not necessary to prepare complex control input data to simulate the various manual operations which may occur during the plant transient. In this study, the Steam Generator Tube Rupture (SGTR) Accident, which occurred at Ulchin Unit 4 in April 2002, has been simulated using ViSA and the simulation results are compared with the measured plant data. The RETRAN-3D plant input data used in this simulation is a genetic input deck prepared for the simulation from a normal operation condition to a Small-Break LOCA. From the results of the SGTR simulation, we found that the GUI functions of ViSA and the input data for Ulchin Unit 4 have enough effectiveness and soundness. (author)

  18. [Simultaneous Traumatic Rupture of Patellar Ligament and Contralateral Rupture of Quadriceps Femoris Muscle].

    Science.gov (United States)

    Hladký, V; Havlas, V

    2017-01-01

    Our paper presents a unique case of a 64-year-old patient after a fall, treated with oral antidiabetic drugs for type II diabetes mellitus. Following a series of examinations, a bilateral injury was diagnosed - patellar ligament tear on the right side and rupture of quadriceps femoris muscle on the left side. It is a rare injury, complicated by simultaneous involvement of both knee joints. The used therapy consisted of a bilateral surgery followed by gradual verticalisation, first with the support of a walking frame and later with the use of forearm crutches. During the final examination, the patient demonstrated full flexion at both knees, while an extension deficit of approx. 5 degrees was still present on the left side. The right knee X-ray showed a proper position of the patella after the removal of temporary tension band wire. Although the clinical results of operative treatment of both the patellar ligament rupture and rupture of quadriceps femoris muscle are in most cases good, early operative treatment, proper technique and post-operative rehabilitation are a prerequisite for success. Key words: knee injuries, patellar ligament, quadriceps muscle, rupture.

  19. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifting Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, S. L.

    2009-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle s Kevlar-49 (DuPont) fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed nonconservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23 percent lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  20. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifing Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, Leigh

    2007-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle's Kevlar-49 fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed non-conservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic-plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23% lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  1. Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.

    1983-01-01

    System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

  2. The U-tube sampling methodology and real-time analysis of geofluids

    International Nuclear Information System (INIS)

    Freifeld, Barry; Perkins, Ernie; Underschultz, James; Boreham, Chris

    2009-01-01

    The U-tube geochemical sampling methodology, an extension of the porous cup technique proposed by Wood (1973), provides minimally contaminated aliquots of multiphase fluids from deep reservoirs and allows for accurate determination of dissolved gas composition. The initial deployment of the U-tube during the Frio Brine Pilot CO 2 storage experiment, Liberty County, Texas, obtained representative samples of brine and supercritical CO 2 from a depth of 1.5 km. A quadrupole mass spectrometer provided real-time analysis of dissolved gas composition. Since the initial demonstration, the U-tube has been deployed for (1) sampling of fluids down gradient of the proposed Yucca Mountain High-Level Waste Repository, Armagosa Valley, Nevada (2) acquiring fluid samples beneath permafrost in Nunuvut Territory, Canada, and (3) at a CO 2 storage demonstration project within a depleted gas reservoir, Otway Basin, Victoria, Australia. The addition of in-line high-pressure pH and EC sensors allows for continuous monitoring of fluid during sample collection. Difficulties have arisen during U-tube sampling, such as blockage of sample lines from naturally occurring waxes or from freezing conditions; however, workarounds such as solvent flushing or heating have been used to address these problems. The U-tube methodology has proven to be robust, and with careful consideration of the constraints and limitations, can provide high quality geochemical samples.

  3. Steam generator tube integrity program: Annual report, August 1995--September 1996. Volume 2

    International Nuclear Information System (INIS)

    Diercks, D.R.; Bakhtiari, S.; Kasza, K.E.; Kupperman, D.S.; Majumdar, S.; Park, J.Y.; Shack, W.J.

    1998-02-01

    This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of the program in August 1995 through September 1996. The program is divided into five tasks: (1) assessment of inspection reliability, (2) research on ISI (inservice-inspection) technology, (3) research on degradation modes and integrity, (4) tube removals from steam generators, and (5) program management. Under Task 1, progress is reported on the preparation of facilities and evaluation of nondestructive evaluation techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate failure pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Results are reported in Task 2 on closed-form solutions and finite-element electromagnetic modeling of EC probe responses for various probe designs and flaw characteristics. In Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe-accident conditions. Crack behavior and stability are also being modeled to provide guidance for test facility design, develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the acquisition of tubes and tube sections from retired steam generators for use in the other research tasks. Progress on the acquisition of tubes from the Salem and McGuire 1 nuclear plants is reported

  4. Gender, smoking, body size, and aneurysm geometry influence the biomechanical rupture risk of abdominal aortic aneurysms as estimated by finite element analysis.

    Science.gov (United States)

    Lindquist Liljeqvist, Moritz; Hultgren, Rebecka; Siika, Antti; Gasser, T Christian; Roy, Joy

    2017-04-01

    Finite element analysis (FEA) has been suggested to be superior to maximal diameter measurements in predicting rupture of abdominal aortic aneurysms (AAAs). Our objective was to investigate to what extent previously described rupture risk factors were associated with FEA-estimated rupture risk. One hundred forty-six patients with an asymptomatic AAA of a 40- to 60-mm diameter were retrospectively identified and consecutively included. The patients' computed tomography angiograms were analyzed by FEA without (neutral) and with (specific) input of patient-specific mean arterial pressure (MAP), gender, family history, and age. The maximal wall stress/wall strength ratio was described as a rupture risk equivalent diameter (RRED), which translated this ratio into an average aneurysm diameter of corresponding rupture risk. In multivariate linear regression, RRED neutral increased with female gender (3.7 mm; 95% confidence interval [CI], 0.13-7.3) and correlated with patient height (0.27 mm/cm; 95% CI, 0.11-0.43) and body surface area (BSA, 16 mm/m 2 ; 95% CI, 8.3-24) and inversely with body mass index (BMI, -0.40 mm/kg m -2 ; 95% CI, -0.75 to -0.054) in a wall stress-dependent manner. Wall stress-adjusted RRED neutral was raised if the patient was currently smoking (1.1 mm; 95% CI, 0.21-1.9). Age, MAP, family history, and patient weight were unrelated to RRED neutral . In specific FEA, RRED specific increased with female gender, MAP, family history positive for AAA, height, and BSA, whereas it was inversely related to BMI. All results were independent of aneurysm diameter. Peak wall stress and RRED correlated with aneurysm diameter and lumen volume. Female gender, current smoking, increased patient height and BSA, and low BMI were found to increase the mechanical rupture risk of AAAs. Previously described rupture risk factors may in part be explained by patient characteristic-dependent variations in aneurysm biomechanics. Copyright © 2016 Society for Vascular

  5. Untreated silicone breast implant rupture

    DEFF Research Database (Denmark)

    Hölmich, Lisbet R; Vejborg, Ilse M; Conrad, Carsten

    2004-01-01

    Implant rupture is a well-known complication of breast implant surgery that can pass unnoticed by both patient and physician. To date, no prospective study has addressed the possible health implications of silicone breast implant rupture. The aim of the present study was to evaluate whether untre...

  6. Traumatic rupture of an intracranial dermoid cyst

    Directory of Open Access Journals (Sweden)

    Raksha Ramlakhan, BMedSc, MBBCh

    2015-01-01

    Full Text Available Intracranial dermoid cysts are congenital tumors of ectodermal origin. Rupture of these cysts can occur spontaneously, but rupture in association with trauma is reported infrequently. The diagnosis of rupture is made by the presence of lipid (cholesterol droplets in the subarachnoid spaces and ventricles. Nonenhanced CT of the head demonstrates multiple foci of low attenuation that correspond with hyperintense signal on T1-weighted MRI. We present a case of an adult patient with rupture of an intracranial dermoid cyst, precipitated by minor trauma.

  7. Rapid Estimates of Rupture Extent for Large Earthquakes Using Aftershocks

    Science.gov (United States)

    Polet, J.; Thio, H. K.; Kremer, M.

    2009-12-01

    The spatial distribution of aftershocks is closely linked to the rupture extent of the mainshock that preceded them and a rapid analysis of aftershock patterns therefore has potential for use in near real-time estimates of earthquake impact. The correlation between aftershocks and slip distribution has frequently been used to estimate the fault dimensions of large historic earthquakes for which no, or insufficient, waveform data is available. With the advent of earthquake inversions that use seismic waveforms and geodetic data to constrain the slip distribution, the study of aftershocks has recently been largely focused on enhancing our understanding of the underlying mechanisms in a broader earthquake mechanics/dynamics framework. However, in a near real-time earthquake monitoring environment, in which aftershocks of large earthquakes are routinely detected and located, these data may also be effective in determining a fast estimate of the mainshock rupture area, which would aid in the rapid assessment of the impact of the earthquake. We have analyzed a considerable number of large recent earthquakes and their aftershock sequences and have developed an effective algorithm that determines the rupture extent of a mainshock from its aftershock distribution, in a fully automatic manner. The algorithm automatically removes outliers by spatial binning, and subsequently determines the best fitting “strike” of the rupture and its length by projecting the aftershock epicenters onto a set of lines that cross the mainshock epicenter with incremental azimuths. For strike-slip or large dip-slip events, for which the surface projection of the rupture is recti-linear, the calculated strike correlates well with the strike of the fault and the corresponding length, determined from the distribution of aftershocks projected onto the line, agrees well with the rupture length. In the case of a smaller dip-slip rupture with an aspect ratio closer to 1, the procedure gives a measure

  8. Effect of tube-support interaction on the dynamic responses of heat exchanger tubes

    International Nuclear Information System (INIS)

    Shin, Y.S.; Jendrzejczyk, J.A.; Wambsganss, M.W.

    1977-01-01

    Operating heat exchangers have experienced tube damages due to excessive flow-induced vibration. The relatively small inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the tube vibrational characteristics. In attempting a theoretical analysis, questions arise as to the effects of tube-baffle impacting on dynamic responses. Experiments were performed to determine the effects of tube-baffle impacting in vertical/horizontal tube orientation, and in air/water medium on the vibrational characteristics (resonant frequencies, mode shapes, and damping) and displacement response amplitudes of a seven-span tube model. The tube and support conditions were prototypic, and overall length approximately one-third that of a straight tube segment of the steam generator designed for the CRBR. The test results were compared with the analytical results based on the multispan beam with ''knife-edge'' supports

  9. Arthroscintigraphy in suspected rotator cuff rupture

    International Nuclear Information System (INIS)

    Gratz, S.; Behr, T.; Becker, W.; Koester, G.; Vosshenrich, R.; Grabbe, E.

    1998-01-01

    Aim: In order to evaluate the diagnostic efficiency of arthroscintigraphy in suspected rotator cuff ruptures this new imaging procedure was performed 20 times in 17 patients with clinical signs of a rotator cuff lesion. The scintigraphic results were compared with sonography (n=20), contrast arthrography (n=20) and arthroscopy (n=10) of the shoulder joint. Methods: After performing a standard bone scintigraphy with intravenous application of 300 MBq 99m-Tc-methylene diphosphonate (MDP) for landmarking of the shoulder region arthroscintigraphy was performed after an intraarticular injection of 99m-Tc microcolloid (ALBU-RES 400 μCi/5 ml). The application was performed either in direct combination with contrast arthrography (n=10) or ultrasound conducted mixed with a local anesthetic (n=10). Findings at arthroscopical surgery (n=10) were used as the gold standard. Results: In case of complete rotator cuff rupture (n=5), arthroscintigraphy and radiographic arthrography were identical in 5/5. In one patient with advanced degenerative alterations of the shoulder joint radiographic arthrography incorrectly showed a complete rupture which was not seen by arthroscintigraphy and endoscopy. In 3 patients with incomplete rupture, 2/3 results were consistant. A difference was seen in one patient with a rotator cuff, that has been already revised in the past and that suffered of capsulitis and calcification. Conclusion: Arthroscinitgraphy is a sensitive technique for detection of rotator cuff ruptures. Because of the lower viscosity of the active compound, small ruptures can be easily detected, offering additional value over radiographic arthrography and ultrasound, especially for evaluation of incomplete cuff ruptures. (orig.) [de

  10. Toward tsunami early warning system in Indonesia by using rapid rupture durations estimation

    International Nuclear Information System (INIS)

    Madlazim

    2012-01-01

    Indonesia has Indonesian Tsunami Early Warning System (Ina-TEWS) since 2008. The Ina-TEWS has used automatic processing on hypocenter; Mwp, Mw (mB) and Mj. If earthquake occurred in Ocean, depth 7, then Ina-TEWS announce early warning that the earthquake can generate tsunami. However, the announcement of the Ina-TEWS is still not accuracy. Purposes of this research are to estimate earthquake rupture duration of large Indonesia earthquakes that occurred in Indian Ocean, Java, Timor sea, Banda sea, Arafura sea and Pasific ocean. We analyzed at least 330 vertical seismogram recorded by IRIS-DMC network using a direct procedure for rapid assessment of earthquake tsunami potential using simple measures on P-wave vertical seismograms on the velocity records, and the likelihood that the high-frequency, apparent rupture duration, T dur . T dur can be related to the critical parameters rupture length (L), depth (z), and shear modulus (μ) while T dur may be related to wide (W), slip (D), z or μ. Our analysis shows that the rupture duration has a stronger influence to generate tsunami than Mw and depth. The rupture duration gives more information on tsunami impact, Mo/μ, depth and size than Mw and other currently used discriminants. We show more information which known from the rupture durations. The longer rupture duration, the shallower source of the earthquake. For rupture duration greater than 50 s, the depth less than 50 km, Mw greater than 7, the longer rupture length, because T dur is proportional L and greater Mo/μ. Because Mo/μ is proportional L. So, with rupture duration information can be known information of the four parameters. We also suggest that tsunami potential is not directly related to the faulting type of source and for events that have rupture duration greater than 50 s, the earthquakes generated tsunami. With available real-time seismogram data, rapid calculation, rupture duration discriminant can be completed within 4–5 min after an earthquake

  11. Rupture of steam lines between blocks D and G

    International Nuclear Information System (INIS)

    1999-01-01

    Analysis of steam lines rupture between blocks D and G of Ignalina NPP was performed. Model for evaluation of thermo hydrodynamic parameters was developed. Structural analysis of the shaft building was done as well. State of the art codes such as RELAP5, ALGOR, NEPTUNE were used in these calculations

  12. Analysis of fuel pin mechanics in case of flow blockage of a single RBMK channel

    International Nuclear Information System (INIS)

    Pierro, F.; Moretti, F.; Mazzini, D.; D'Auria, F.

    2005-01-01

    The evaluation of the consequences of the pressure tube rupture due to accidental overheating is one of the key elements for addressing an RBMK safety analysis, since it causes the lost of design boundaries against the fission products release. Several events are expected to take place: thermal hydraulic crisis (energy unbalance), fuel overheating, fuel rod damage, pressure tube overheating, pressure tube failure and graphite stack damage, Hydrogen and fission products release. The present work deals with the research activity carried out at ''Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione'' (DIMNP) of the University of Pisa aimed at assessing numerical models for safety analysis of the RBMK-1000. The attention is focused on the modelling of (1) a single fuel channel and its surrounding graphite column for evaluating the transient conditions enabling the different damaging phenomena, (2) a single fuel rod for investigating fuel pin behaviour, (3) the ruptured fuel channel for figuring the magnitude of the hydrodynamic loads acting on fuel rods. Different codes were employed to cover the competences for the investigation of each field; in particular, RELAP5 code for thermal-hydraulics, FRAPCON-3 and FRAPTRAN1-2 codes for fuel pin mechanics, FLUENT-6 for fluid dynamics. The paper discusses the numerical models, the analysis capabilities of numerical models in comparison with available data about the Leningrad NPP 1992 accident. Furthermore, the possibility to draw a failure map identifying the range of the cladding safety respect to the transient condition is outlined. (author)

  13. Categorising YouTube

    DEFF Research Database (Denmark)

    Simonsen, Thomas Mosebo

    2011-01-01

    This article provides a genre analytical approach to creating a typology of the User Generated Content (UGC) of YouTube. The article investigates the construction of navigation processes on the YouTube website. It suggests a pragmatic genre approach that is expanded through a focus on YouTube......’s technological affordances. Through an analysis of the different pragmatic contexts of YouTube, it is argued that a taxonomic understanding of YouTube must be analysed in regards to the vacillation of a user-driven bottom-up folksonomy and a hierarchical browsing system that emphasises a culture of competition...... and which favours the already popular content of YouTube. With this taxonomic approach, the UGC videos are registered and analysed in terms of empirically based observations. The article identifies various UGC categories and their principal characteristics. Furthermore, general tendencies of the UGC within...

  14. State-variable analysis of inelastic deformation of thin-walled tubes. II. Data analysis and simulations

    International Nuclear Information System (INIS)

    Wire, G.L.; Duncan, D.R.; Cannon, N.S.; Johnson, G.D.; Alexopoulos, P.S.; Li, C.Y.

    Inelastic analysis is performed to calculate the deformation of thin-walled, internally pressurized, tube under a variety of loading modes. A state-variable approach was used to describe the material properties. The material parameters of the constitutive equations used were determined based on uniaxial, load relaxation, tensile tests, and internally pressurized tubes under creep and constant-displacement-rate modes of loading. The simulated results were compared with the experimental data. The significance of the comparison is discussed in terms of the validity of a state-variable approach used to describe the deformation properties in mechanical testing

  15. Creep-rupture behavior of candidate Stirling engine iron supperalloys in high-pressure hydrogen. Volume 2: Hydrogen creep-rupture behavior

    Science.gov (United States)

    Bhattacharyya, S.; Peterman, W.; Hales, C.

    1984-01-01

    The creep rupture behavior of nine iron base and one cobalt base candidate Stirling engine alloys is evaluated. Rupture life, minimum creep rate, and time to 1% strain data are analyzed. The 3500 h rupture life stress and stress to obtain 1% strain in 3500 h are also estimated.

  16. 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.

  17. Tube sheet structural analysis of intermediate heat exchanger for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Nakagawa, Y.; Ueno, T.; Fukuda, Y.; Ichimiya, M.

    1983-01-01

    The Prototype Fast Breeder Reactor 'Monju' is the first power generating fast breeder reactor in Japan. We have been designing the components of the plant for manufacturing. Among these is the intermediate heat exchanger (IHX) which exchanges heat between primary and secondary sodium loop. The tube sheet of IHX (shell to ligament junction) is a difficult area from the view point of structural strength design under elevated temperature. To validate the structural integrity of tube sheet we performed the series of inelastic analysis and tube sheet thermal shock test using test pieces and half scale model of actual design. The results of inelastic analyses showed there is little progressive deformation around shell to ligament structural discontinuous junction. Furthermore, thermal shock tests showed no increase of an accumulative deformation. By these analyses and experiments, structural reliability of tube sheet could be shown. (author)

  18. Neck curve polynomials in neck rupture model

    International Nuclear Information System (INIS)

    Kurniadi, Rizal; Perkasa, Yudha S.; Waris, Abdul

    2012-01-01

    The Neck Rupture Model is a model that explains the scission process which has smallest radius in liquid drop at certain position. Old fashion of rupture position is determined randomly so that has been called as Random Neck Rupture Model (RNRM). The neck curve polynomials have been employed in the Neck Rupture Model for calculation the fission yield of neutron induced fission reaction of 280 X 90 with changing of order of polynomials as well as temperature. The neck curve polynomials approximation shows the important effects in shaping of fission yield curve.

  19. Swirl flow analysis in a helical wire inserted tube using CFD code

    International Nuclear Information System (INIS)

    Park, Yusun; Chang, Soon Heung

    2010-01-01

    An analysis on the two-phase flow in a helical wire inserted tube using commercial CFD code, CFX11.0, was performed in bubbly flow and annular flow regions. The analysis method was validated with the experimental results of Takeshima. Bubbly and annular flows in a 10 mm inner diameter tube with varying pitch lengths and inserted wire diameters were simulated using the same analysis methods after validation. The geometry range of p/D was 1-4 and e/D was 0.08-0.12. The results show that the inserted wire with a larger diameter increased swirl flow generation. An increasing swirl flow was seen as the pitch length increased. Regarding pressure loss, smaller pitch lengths and inserted wires with larger diameters resulted in larger pressure loss. The average liquid film thickness increased as the pitch length and the diameter of the inserted wire increased in the annular flow region. Both in the bubbly flow and annular flow regions, the effect of pitch length on swirl flow generation and pressure loss was more significant than that of the inserted wire diameters. Pitch length is a more dominant factor than inserted wire diameter for the design of the swirl flow generator in small diameter tubes.

  20. Vibration impact acoustic emission technique for identification and analysis of defects in carbon steel tubes: Part A Statistical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Halim, Zakiah Abd [Universiti Teknikal Malaysia Melaka (Malaysia); Jamaludin, Nordin; Junaidi, Syarif [Faculty of Engineering and Built, Universiti Kebangsaan Malaysia, Bangi (Malaysia); Yahya, Syed Yusainee Syed [Universiti Teknologi MARA, Shah Alam (Malaysia)

    2015-04-15

    Current steel tubes inspection techniques are invasive, and the interpretation and evaluation of inspection results are manually done by skilled personnel. This paper presents a statistical analysis of high frequency stress wave signals captured from a newly developed noninvasive, non-destructive tube inspection technique known as the vibration impact acoustic emission (VIAE) technique. Acoustic emission (AE) signals have been introduced into the ASTM A179 seamless steel tubes using an impact hammer, and the AE wave propagation was captured using an AE sensor. Specifically, a healthy steel tube as the reference tube and four steel tubes with through-hole artificial defect at different locations were used in this study. The AE features extracted from the captured signals are rise time, peak amplitude, duration and count. The VIAE technique also analysed the AE signals using statistical features such as root mean square (r.m.s.), energy, and crest factor. It was evident that duration, count, r.m.s., energy and crest factor could be used to automatically identify the presence of defect in carbon steel tubes using AE signals captured using the non-invasive VIAE technique.

  1. Application of numerical analysis techniques to eddy current testing for steam generator tubes

    International Nuclear Information System (INIS)

    Morimoto, Kazuo; Satake, Koji; Araki, Yasui; Morimura, Koichi; Tanaka, Michio; Shimizu, Naoya; Iwahashi, Yoichi

    1994-01-01

    This paper describes the application of numerical analysis to eddy current testing (ECT) for steam generator tubes. A symmetrical and three-dimensional sinusoidal steady state eddy current analysis code was developed. This code is formulated by future element method-boundary element method coupling techniques, in order not to regenerate the mesh data in the tube domain at every movement of the probe. The calculations were carried out under various conditions including those for various probe types, defect orientations and so on. Compared with the experimental data, it was shown that it is feasible to apply this code to actual use. Furthermore, we have developed a total eddy current analysis system which consists of an ECT calculation code, an automatic mesh generator for analysis, a database and display software for calculated results. ((orig.))

  2. Steam generator tube integrity program. Semiannual report, August 1995--March 1996

    International Nuclear Information System (INIS)

    Diercks, D.R.; Bakhtiari, S.; Chopra, O.K.

    1997-04-01

    This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of that program in August 1995 through March 1996. The program is divided into five tasks, namely (1) Assessment of Inspection Reliability, (2) Research on ISI (in-service-inspection) Technology, (3) Research on Degradation Modes and Integrity, (4) Development of Methodology and Technical Requirements for Current and Emerging Regulatory Issues, and (5) Program Management. Under Task 1, progress is reported on the preparation of and evaluation of nondestructive evaluation (NDE) techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate burst pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Under Task 2, results are reported on closed-form solutions and finite element electromagnetic modeling of EC probe response for various probe designs and flaw characteristics. Under Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe accident conditions. In addition, crack behavior and stability are being modeled to provide guidance on test facility design, to develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and to predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the cracking and failure of tubes that have been repaired by sleeving, and with a review of literature on this subject

  3. Evaluation on double-wall-tube residual stress distribution of sodium-heated steam generator by neutron diffraction and numerical analysis

    International Nuclear Information System (INIS)

    Kisohara, N.; Suzuki, H.; Akita, K.; Kasahara, N.

    2012-01-01

    A double-wall-tube is nominated for the steam generator heat transfer tube of future sodium fast reactors (SFRs) in Japan, to decrease the possibility of sodium/water reaction. The double-wall-tube consists of an inner tube and an outer tube, and they are mechanically contacted to keep the heat transfer of the interface between the inner and outer tubes by their residual stress. During long term SG operation, the contact stress at the interface gradually falls down due to stress relaxation. This phenomenon might increase the thermal resistance of the interface and degrade the tube heat transfer performance. The contact stress relaxation can be predicted by numerical analysis, and the analysis requires the data of the initial residual stress distributions in the tubes. However, unclear initial residual stress distributions prevent precious relaxation evaluation. In order to resolve this issue, a neutron diffraction method was employed to reveal the tri-axial (radius, hoop and longitudinal) initial residual stress distributions in the double-wall-tube. Strain gauges also were used to evaluate the contact stress. The measurement results were analyzed using a JAEA's structural computer code to determine the initial residual stress distributions. Based on the stress distributions, the structural computer code has predicted the transition of the relaxation and the decrease of the contact stress. The radial and longitudinal temperature distributions in the tubes were input to the structural analysis model. Since the radial thermal expansion difference between the inner (colder) and outer (hotter) tube reduces the contact stress and the tube inside steam pressure contributes to increasing it, the analytical model also took these effects into consideration. It has been conduced that the inner and outer tubes are contacted with sufficient stresses during the plant life time, and that effective heat transfer degradation dose not occur in the double-wall-tube SG. (authors)

  4. Ruptured gastroepiploic artery aneurysm: A case report

    Directory of Open Access Journals (Sweden)

    Ahmad S. Ashrafi

    Full Text Available Introduction: Gastroepiploic artery aneurysms are extremely rare, with few reported cases in the literature. The risk of rupture however, is high and thus warrants attention. Presentation of case: Here we present a rare case of a women who presented to the emergency department in shock and was found to have a ruptured gastroepiploic artery aneurysm during surgical exploration. Suture ligation of the aneurysm was completed. Discussion: Although rare, gastroepiploic artery aneurysms have up to a 90% rate of rupture and therefore require intervention. A laparoscopic approach has been described however, in cases where rupture has occurred, urgent laparotomy and control of hemorrhage is needed. Conclusion: We describe a rare case of a ruptured gastroepiploic aneurysm that was successfully managed with urgent laparotomy and aneurysmal resection. Keywords: Gastroepiploic, Aneurysm, Hemorrhage, Case report

  5. The influence of atmospheric pressure on aortic aneurysm rupture--is the diameter of the aneurysm important?

    Science.gov (United States)

    Urbanek, Tomasz; Juśko, Maciej; Niewiem, Alfred; Kuczmik, Wacław; Ziaja, Damian; Ziaja, Krzysztof

    2015-01-01

    The rate of aortic aneurysm rupture correlates with the aneurysm's diameter, and a higher rate of rupture is observed in patients with larger aneurysms. According to the literature, contradictory results concerning the relationship between atmospheric pressure and aneurysm size have been reported. In this paper, we assessed the influence of changes in atmospheric pressure on abdominal aneurysm ruptures in relationship to the aneurysm's size. The records of 223 patients with ruptured abdominal aneurysms were evaluated. All of the patients had been admitted to the department in the period 1997-2007 from the Silesia region. The atmospheric pressures on the day of the rupture and on the days both before the rupture and between the rupture events were compared. The size of the aneurysm was also considered in the analysis. There were no statistically significant differences in pressure between the days of rupture and the remainder of the days within an analysed period. The highest frequency of the admission of patients with a ruptured aortic aneurysm was observed during periods of winter and spring, when the highest mean values of atmospheric pressure were observed; however, this observation was not statistically confirmed. A statistically non-significant trend towards the higher rupture of large aneurysms (> 7 cm) was observed in the cases where the pressure increased between the day before the rupture and the day of the rupture. This trend was particularly pronounced in patients suffering from hypertension (p = 0.1). The results of this study do not support the hypothesis that there is a direct link between atmospheric pressure values and abdominal aortic aneurysm ruptures.

  6. Comparing slow and fast rupture in laboratory experiments

    Science.gov (United States)

    Aben, F. M.; Brantut, N.; David, E.; Mitchell, T. M.

    2017-12-01

    During the brittle failure of rock, elastically stored energy is converted into a localized fracture plane and surrounding fracture damage, seismic radiation, and thermal energy. However, the partitioning of energy might vary with the rate of elastic energy release during failure. Here, we present the results of controlled (slow) and dynamic (fast) rupture experiments on dry Lanhélin granite and Westerly granite samples, performed under triaxial stress conditions at confining pressures of 50 and 100 MPa. During the tests, we measured sample shortening, axial load and local strains (with 2 pairs of strain gauges glued directly onto the sample). In addition, acoustic emissions (AEs) and changes in seismic velocities were monitored. The AE rate was used as an indicator to manually control the axial load on the sample to stabilize rupture in the quasi-static failure experiments. For the dynamic rupture experiments a constant strain rate of 10-5 s-1 was applied until sample failure. A third experiment, labeled semi-controlled rupture, involved controlled rupture up to a point where the rupture became unstable and the remaining elastic energy was released dynamically. All experiments were concluded after a macroscopic fracture had developed across the whole sample and frictional sliding commenced. Post-mortem samples were epoxied, cut and polished to reveal the macroscopic fracture and the surrounding damage zone. The samples failed with average rupture velocities varying from 5x10-6 m/s up to >> 0.1 m/s. The analyses of AE locations on the slow ruptures reveal that within Westerly granite samples - with a smaller grain size - fracture planes are disbanded in favor of other planes when a geometrical irregularity is encountered. For the coarser grained Lanhélin granite a single fracture plane is always formed, although irregularities are recognized as well. The semi-controlled experiments show that for both rock types the rupture can become unstable in response to these

  7. Ruptured Spleen

    Science.gov (United States)

    ... be caused by various underlying problems, such as mononucleosis and other infections, liver disease, and blood cancers. ... cause a ruptured spleen. For instance, people with mononucleosis — a viral infection that can cause an enlarged ...

  8. The root caused analysis of leakaged heat exchanger tube

    International Nuclear Information System (INIS)

    Shamsudin, Shaiful Rizam; Salleh, M.A.A. Mohd; Rahmat, Azmi; Anuar, Mohd Arif; Harun, Mohd; Zayid, Hafizal; Noor, Mazlee Mohd

    2015-01-01

    AISI type 316L stainless steel was used as a heat exchanger tube material in an inter-cooler column. After less than a year of operation, severe corrosion failures occurred and a transverse opening leakage was observed on one of the heat exchanger tubes. The failed tube was carefully analyzed using various metallurgical laboratory equipments. The root cause of the tube leakage was believed due to the presence of horizontal micro and macro pores as a hydrogen gas entrapment during casting of the parent ingot. The overlapped and gaping pores formed notch on the shell side of the tube surface, and it increasingly evident when the use of a high-energy water-jet and metal brush as cleaning procedure results in an establishment of pitting type local-action corrosion cells penetrated the tube wall. As a result, corrosive fluid in the tube side dissolved into the cooling water, accelerating the corrosion process.

  9. Analysis of steam generator tube sections removed from Gentilly-2 nuclear generating station

    International Nuclear Information System (INIS)

    Semmler, J.; Lockley, A.J.; Doyon, D.

    2010-01-01

    In order to meet the requirements of CSA Standards CAN/CSA N285.4-94, which states, 'A section of one tube in a deposit region shall be removed from one steam generator for metallurgical examination', Gentilly-2 has been removing steam generator tube sections on a regular basis for analysis at Chalk River Laboratories. In 2009 April, sections from the hot leg and the cold leg of a steam generator tube were removed for detailed metallographic examination and characterization. The hot leg tube section covered the area from within the tube sheet up to below the second support plate, and the cold leg tube section covered the area from within the tube sheet to below the third preheater support plate. After a general visual and photographic examination, the area above the tube sheet on the hot leg side where the sludge pile is highest was examined in detail. Visual and macro-photography of the two tube sections within the tube sheet were also examined. Additional metallographic and surface examinations of both tube inner diameter and tube outer diameter, and surface roughness measurements of tube inner diameter were also completed. The surface activities (μCi/cm 2 ) of cold leg and hot leg specimens were measured before and after electrolytic descaling, and major and minor radionuclides were identified; a comparison of the surface activities for hot leg with the values for the cold leg were made. The results from the initial γ-spectroscopy measurements, and the measurements after the descaling of the specimens were used to estimate decontamination factors for each specimen and for each radionuclide. The tube specimens had thin outer diameter oxides; all four steam generators were chemically cleaned in 2005. All specimens had inner diameter deposits; the inner diameter deposits on the cold leg were heavier than those on the hot leg as expected. Primary side oxide loadings of specimens were used to estimate the total oxide inventory in 2009. The oxide

  10. Draining Collars and Lenses in Liquid-Lined Vertical Tubes.

    Science.gov (United States)

    Jensen

    2000-01-01

    The speed at which an annular liquid collar drains under gravity g in a vertical tube of radius a, when the tube has an otherwise thin viscous liquid lining on its interior, is determined by a balance between the collar's weight and viscous shear stresses confined to narrow regions in the neighborhood of the collar's effective contact lines. Whether a collar grows or shrinks in volume as it drains depends on the modified Bond number B=rho g a(2)/(sigmaepsilon), where rho is the fluid density, sigma is its surface tension, and epsilona is the thickness of the thin film immediately ahead of the collar. Asymptotic methods are used here to determine the following nonlinear stability criteria for an individual collar, valid in the limit of small epsilon. For 0draining collars grow in volume and, in sufficiently long tubes, ultimately "snap off" to form stable lenses. For 0.5960drain, so that any lens ultimately ruptures, unless stabilizing intermolecular forces allow the formation of a lamella supported by a macroscopic Plateau border. If surfactant immobilizes the liquid's free surface, these critical values of B are reduced by a factor of 2 but the distance a collar must travel before it snaps off is unchanged. Gravitationally driven snap off is therefore most likely to occur in long tubes with radii substantially less than the capillary lengthscale sigma/rhog)(1/2). Copyright 2000 Academic Press.

  11. Clinical, morphological, and hemodynamic independent characteristic factors for rupture of posterior communicating artery aneurysms.

    Science.gov (United States)

    Zhang, Ying; Jing, Linkai; Liu, Jian; Li, Chuanhui; Fan, Jixing; Wang, Shengzhang; Li, Haiyun; Yang, Xinjian

    2016-08-01

    To identify clinical, morphological, and hemodynamic independent characteristic factors that discriminate posterior communicating artery (PCoA) aneurysm rupture status. 173 patients with single PCoA aneurysms (108 ruptured, 65 unruptured) between January 2012 and June 2014 were retrospectively collected. Patient-specific models based on their three-dimensional digital subtraction angiography images were constructed and analyzed by a computational fluid dynamic method. All variables were analyzed by univariate analysis and multivariate logistic regression analysis. Two clinical factors (younger age and atherosclerosis), three morphological factors (higher aspect ratio, bifurcation type, and irregular shape), and six hemodynamic factors (lower mean and minimum wall shear stress, higher oscillatory shear index, a greater portion of area under low wall shear stress, unstable and complex flow pattern) were significantly associated with PCoA aneurysm rupture. Independent factors characterizing the rupture status were identified as age (OR 0.956, p=0.015), irregular shape (OR 6.709, pPCoA aneurysm rupture were younger age, irregular shape, and low minimum wall shear stress. This may be useful for guiding risk assessments and subsequent treatment decisions for PCoA aneurysms. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

  12. Misdiagnosed Chest Pain: Spontaneous Esophageal Rupture

    Science.gov (United States)

    Inci, Sinan; Gundogdu, Fuat; Gungor, Hasan; Arslan, Sakir; Turkyilmaz, Atila; Eroglu, Atila

    2013-01-01

    Chest pain is one of themost common complaints expressed by patients presenting to the emergency department, and any initial evaluation should always consider life-threatening causes. Esophageal rupture is a serious condition with a highmortality rate. If diagnosed, successful therapy depends on the size of the rupture and the time elapsed between rupture and diagnosis.We report on a 41-year-old woman who presented to the emergency department complaining of left-sided chest pain for two hours. PMID:27122690

  13. A low power x-ray tube for use in energy dispersive x-ray fluorescence analysis

    International Nuclear Information System (INIS)

    Kataria, S.K.; Govil, Rekha; Lal, M.

    1980-01-01

    A low power X-ray tube with thin molybdenum transmission target for use in energy dispersive X-ray fluorescence (ENDXRF) element analysis has been indigenously built, along with its power supply. The X-ray tube has been in operation since August 1979, and it has been operated upto maximum target voltage of 35 KV and tube current upto 200 μA which is more than sufficient for trace element analysis. This X-ray tube has been used alongwith the indigenously built Si(Li) detector X-ray spectrometer with an energy resolution of 200 eV at 5.9 Kev MnKsub(α) X-ray peak for ENDXRF analysis. A simple procedure of calibration has been developed for thin samples based on the cellulose diluted, thin multielement standard pellets. Analytical sensitivities of the order of a few p.p.m. have been obtained with the experimental setup for elements with 20 < = Z < = 38 and 60 < = Z < = 90. A number of X-ray spectra for samples of environmental, biological, agricultural, industrial and metallurgical interest are presented to demonstrate the salient features of the experimental sep up. (auth.)

  14. Acute Iliac Artery Rupture: Endovascular Treatment

    International Nuclear Information System (INIS)

    Chatziioannou, A.; Mourikis, D.; Katsimilis, J.; Skiadas, V.; Koutoulidis, V.; Katsenis, K.; Vlahos, L.

    2007-01-01

    The authors present 7 patients who suffered iliac artery rupture over a 2 year period. In 5 patients, the rupture was iatrogenic: 4 cases were secondary to balloon angioplasty for iliac artery stenosis and 1 occurred during coronary angioplasty. In the last 2 patients, the rupture was secondary to iliac artery mycotic aneurysm. Direct placement of a stent-graft was performed in all cases, which was dilated until extravasation was controlled. Placement of the stent-graft was successful in all the cases, without any complications. The techniques used, results, and mid-term follow-up are presented. In conclusion, endovascular placement of a stent-graft is a quick, minimally invasive, efficient, and safe method for emergency treatment of acute iliac artery rupture, with satisfactory short- and mid-term results

  15. CT diagnosis of ruptured abdominal aortic aneurysm

    International Nuclear Information System (INIS)

    Sacknoff, R.; Novelline, R.A.; Wittenberg, J.; Waltman, A.C.; De Luca, S.A.; Rhea, J.T.; Lawrason, J.N.

    1986-01-01

    Ruptured abdominal aortic aneurysm (AAA) is a life-threatening condition requiring immediate diagnosis and surgery. In a series of 23 consecutive patients scanned by CT for suspected ruptured AAA, CT proved 100% accurate. In seven patients with surgically or pathologically proved ruptured AAA, CT demonstrated a similar distribution of hemorrhage into the perirenal space and to a lesser degree into the anterior and posterior pararenal spaces. The 16 true-negative examinations included ten in patients with unruptured AAA and six in patients with other diseases. The authors conclude that patients in stable condition with suspected ruptured AAA should be examined by CT

  16. Evaluation of finger A3 pulley rupture in the crimp grip position - a magnetic resonance imaging cadaver study

    Energy Technology Data Exchange (ETDEWEB)

    Bayer, Thomas; Uder, Michael; Janka, Rolf [University of Erlangen-Nuremberg, Department of Radiology, Erlangen (Germany); Adler, Werner [University of Erlangen-Nuremberg, Department of Biometry and Epidemiology, Erlangen (Germany); Schweizer, Andreas [Balgrist, University of Zurich, Department of Orthopaedics, Zurich (Switzerland); Schoeffl, Isabelle [Klinikum Bamberg, Department of Pediatrics, Bamberg (Germany)

    2015-09-15

    The correct diagnosis of an A3 pulley rupture is challenging for musculoskeletal radiologists. An A3 pulley rupture should in theory influence the shape of the proximal interphalangeal joint volar plate (VP) and the amount of bowstringing at level of the VP during finger flexion. The purpose of this study was to perform MRI with metric analysis of the VP configuration and VP bowstringing in cadaver fingers in the crimp grip position and to determine cut points for A3 pulley rupture. MRI in the crimp grip position was performed in 21 cadaver fingers with artificially created flexor tendon pulley tears (fingers with A3 pulley rupture n = 16, fingers without A3 pulley rupture n = 5). The distances of the translation of the VP relative to the middle phalanx base, the distances between the flexor tendons and the VP body, and the distances between the flexor tendon and bone (TB) were measured. Statistical analysis showed significantly lower VP translation distances and significantly higher VP tendon distances if the A3 pulley was ruptured. A2 TB and A4 TB distances did not differ significantly in specimens with and without A3 pulley rupture. The optimal cut points for A3 pulley rupture were a VP translation distance <2.8 mm and a VP tendon distance >1.4 mm. Reduction of the VP translation distance and augmentation of the VP tendon distance are suitable indirect signs of A3 pulley rupture. (orig.)

  17. Spontaneous rupture of adrenal metastasis from hepatocellular carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Chae Hun; Kim, Hyun Jin; Park, Soo Youn; Hwang, Seong Su; Choi, Hyun Joo [St. Vincent Hospital, Suwon (Korea, Republic of)

    2007-03-15

    Rupture of adrenal tumor from various primary origins is a rather rare event. We report here on a ruptured adrenal metastasis from hepatocellular carcinoma, and this ruptured metastasis was observed at the time of the initial diagnosis.

  18. Descriptive analysis of YouTube music therapy videos.

    Science.gov (United States)

    Gooding, Lori F; Gregory, Dianne

    2011-01-01

    The purpose of this study was to conduct a descriptive analysis of music therapy-related videos on YouTube. Preliminary searches using the keywords music therapy, music therapy session, and "music therapy session" resulted in listings of 5000, 767, and 59 videos respectively. The narrowed down listing of 59 videos was divided between two investigators and reviewed in order to determine their relationship to actual music therapy practice. A total of 32 videos were determined to be depictions of music therapy sessions. These videos were analyzed using a 16-item investigator-created rubric that examined both video specific information and therapy specific information. Results of the analysis indicated that audio and visual quality was adequate, while narrative descriptions and identification information were ineffective in the majority of the videos. The top 5 videos (based on the highest number of viewings in the sample) were selected for further analysis in order to investigate demonstration of the Professional Level of Practice Competencies set forth in the American Music Therapy Association (AMTA) Professional Competencies (AMTA, 2008). Four of the five videos met basic competency criteria, with the quality of the fifth video precluding evaluation of content. Of particular interest is the fact that none of the videos included credentialing information. Results of this study suggest the need to consider ways to ensure accurate dissemination of music therapy-related information in the YouTube environment, ethical standards when posting music therapy session videos, and the possibility of creating AMTA standards for posting music therapy related video.

  19. Preliminary Stress Analysis of an IHX Tube Support Plate in Prototype SFR

    International Nuclear Information System (INIS)

    Kim, Sung Kyun; Koo, Gyeong Hoi

    2013-01-01

    In this paper, the structural integrity about the conceptual design of IHX tube support plate was reviewed and the design should be changed because of its high stress concentration at the outer rim area. For reducing its maximum stress, two alternatives were proposed and reviewed for the structural integrity point of view. In both proposing support designs, the maximum stress decreases up to the stress design limit. Tube support plates (TSPs) of the intermediate heat exchanger (IHX) in Prototype GenIV Sodium Cooled Fast Reactor (PGSFR) act to horizontally support IHX tubes against hydraulic loadings and they have numerous flow holes where a primary sodium flows downward and secondary sodium flows upward. Due to its many penetrations, its geometric shape is quite complex and structurally its integrity is quite weaker than other parts. In this study, we investigated the structural integrity of the conceptually designed IHX tube support plate. In addition, TSP's supporting concepts were proposed to increase its structural integrity, and confirmed its integrity by using a finite element analysis

  20. Metrics for comparing dynamic earthquake rupture simulations

    Science.gov (United States)

    Barall, Michael; Harris, Ruth A.

    2014-01-01

    Earthquakes are complex events that involve a myriad of interactions among multiple geologic features and processes. One of the tools that is available to assist with their study is computer simulation, particularly dynamic rupture simulation. A dynamic rupture simulation is a numerical model of the physical processes that occur during an earthquake. Starting with the fault geometry, friction constitutive law, initial stress conditions, and assumptions about the condition and response of the near‐fault rocks, a dynamic earthquake rupture simulation calculates the evolution of fault slip and stress over time as part of the elastodynamic numerical solution (Ⓔ see the simulation description in the electronic supplement to this article). The complexity of the computations in a dynamic rupture simulation make it challenging to verify that the computer code is operating as intended, because there are no exact analytic solutions against which these codes’ results can be directly compared. One approach for checking if dynamic rupture computer codes are working satisfactorily is to compare each code’s results with the results of other dynamic rupture codes running the same earthquake simulation benchmark. To perform such a comparison consistently, it is necessary to have quantitative metrics. In this paper, we present a new method for quantitatively comparing the results of dynamic earthquake rupture computer simulation codes.

  1. Hepatic Rupture Induced by Spontaneous Intrahepatic Hematoma

    Directory of Open Access Journals (Sweden)

    Jin-bao Zhou

    2018-01-01

    Full Text Available The etiology of hepatic rupture is usually secondary to trauma, and hepatic rupture induced by spontaneous intrahepatic hematoma is clinically rare. We describe here a 61-year-old female patient who was transferred to our hospital with hepatic rupture induced by spontaneous intrahepatic hematoma. The patient had no history of trauma and had a history of systemic lupus erythematosus for five years, taking a daily dose of 5 mg prednisone for treatment. The patients experienced durative blunt acute right upper abdominal pain one day after satiation, which aggravated in two hours, accompanied by dizziness and sweating. Preoperative diagnosis was rupture of the liver mass. Laparotomy revealed 2500 mL fluid consisting of a mixture of blood and clot in the peritoneal cavity. A 3.5 cm × 2.5 cm rupture was discovered on the hepatic caudate lobe near the vena cava with active arterial bleeding, and a 5  × 6 cm hematoma was reached on the right posterior lobe of the liver. Abdominal computed tomography (CT and laparotomy revealed spontaneous rupture of intrahepatic hematoma with hemorrhagic shock. The patient was successfully managed by suturing the rupture of the hepatic caudate lobe and clearing part of the hematoma. The postoperative course was uneventful, and the patient was discharged after two weeks of hospitalization.

  2. Physical therapy in the conservative treatment for anterior cruciate ligament rupture followed by contralateral rupture: case report

    OpenAIRE

    Almeida, Gabriel Peixoto Leão; Arruda, Gilvan de Oliveira; Marques, Amélia Pasqual

    2014-01-01

    Although the surgical reconstruction be the obvious indication for the anterior cruciate ligament (ACL) lesion, there is no consensus on whether the results of surgery are superior to those obtained with nonsurgical management. The objective of this report was to describe a case of nonsurgical treatment for ACL rupture followed by a contralateral rupture. A 28-year-old female practitioner of muay-thai and handball suffered a non-contact ACL rupture in the left knee, and three months after the...

  3. Risk factors and perinatal outcome of uterine rupture in a low-resource setting.

    Science.gov (United States)

    Igwegbe, Anthony Osita; Eleje, George Uchenna; Udegbunam, Onyebuchi Izuchukwu

    2013-11-01

    Uterine rupture has continued to be a catastrophic feature of obstetric practice especially in the low-resource settings. This study determined the incidence, predisposing factors, treatment options and feto-maternal outcome of ruptured uterus. A 10-year retrolective study of all cases of uterine ruptures that were managed in Nnamdi Azikiwe University Teaching Hospital, Nnewi, Nigeria between 1st January, 2001 and 31st December, 2010 was undertaken. The proforma was initially used for data collection, which was transferred to a data sheet before entering them into the Epi-info software. Analysis was done using Epi info 2008 (version 3.5.1). Out of 5,585 deliveries over the study period, 47 had uterine rupture, giving an incidence of 0.84% or 1 in 119 deliveries. All the patients were multiparous and majority (63.8%) was unbooked. Traumatic (iatrogenic) rupture predominated (72.1%). Uterine repair with (55.8%) or without (34.9%) bilateral tubal ligation was the commonest surgery performed. Case fatality rate was 16.3%, while the perinatal mortality rate was 88.4%. Average duration of hospitalization following uterine rupture was 10.3 days. Uterine rupture constituted a major obstetric emergency in the study hospital and its environs. The incidence, maternal and perinatal mortalities were high. The traumatic/iatrogenic ruptures constituted the majority of cases, hence, majority of the cases are preventable. There is therefore a dire need for education of our women on health-related issues, utilization of available health facilities, adequate supervision of labour and provision of facilities for emergency obstetric care.

  4. Numerical analysis of mass transfer with graphite oxidation in a laminar flow of multi-component gas mixture through a circular tube

    International Nuclear Information System (INIS)

    Ogawa, Masuro

    1992-10-01

    In the present paper, mass transfer has been numerically studied in a laminar flow through a circular graphite tube to evaluate graphite corrosion rate and generation rate of carbon monoxide during a pipe rupture accident in a high temperature gas cooled reactor. In the analysis, heterogeneous (graphite oxidation and graphite/carbon dioxide reaction) and homogeneous (carbon monoxide combustion) chemical reactions were dealt in the multi-component gas mixture; helium, oxygen, carbon monoxide and carbon dioxide. Multi-component diffusion coefficients were used in a diffusion term. Mass conservation equations of each gas component, mass conservation equation and momentum conservation equations of the gas mixture were solved by using SIMPLE algorism. Chemical reactions between graphite and oxygen, graphite and carbon dioxide, and carbon monoxide combustion were taken into account in the present numerical analysis. An energy equation for the gas mixture was not solved and temperature was held to be constant in order to understand basic mass transfer characteristics without heat transfer. But, an energy conservation equation for single component gas was added to know heat transfer characteristics without mass transfer. The effects of these chemical reactions on the mass transfer coefficients were quantitatively and qualitatively clarified in the range of 50 to 1000 of inlet Reynolds numbers, 0 to 0.5 of inlet oxygen mass fraction and 800 to 1600degC of temperature. (author)

  5. Use of ICD-10 codes to monitor uterine rupture

    DEFF Research Database (Denmark)

    Thisted, Dorthe L A; Mortensen, Laust Hvas; Hvidman, Lone

    2014-01-01

    OBJECTIVES: Uterine rupture is a rare but severe complication in pregnancies after a previous cesarean section. In Denmark, the monitoring of uterine rupture is based on reporting of relevant diagnostic codes to the Danish Medical Birth Registry (MBR). The aim of our study was to examine the vali......OBJECTIVES: Uterine rupture is a rare but severe complication in pregnancies after a previous cesarean section. In Denmark, the monitoring of uterine rupture is based on reporting of relevant diagnostic codes to the Danish Medical Birth Registry (MBR). The aim of our study was to examine...... uterine ruptures, the sensitivity and specificity of the codes for uterine rupture were 83.8% and 99.1%, respectively. CONCLUSION: During the study period the monitoring of uterine rupture in the MBR was inadequate....

  6. Traumatic Fundal Rupture of unscarred Uterus in a Primigravida ...

    African Journals Online (AJOL)

    Background: Uterine rupture is an infrequent but life threatening obstetric emergency. Rupture of previously scarred uterus is often encountered especially in multiparous women, but the traumatic rupture of an unscarred primigravid uterus as presented here is a relatively rare event. We report a case of rupture of an ...

  7. Analysis of the dynamic response of a double rupture disc assembly to simulated sodium-water reaction pressure pulses

    International Nuclear Information System (INIS)

    Leonard, J.R.

    1980-03-01

    A series of double rupture disc experiments were conducted in 1979 to evaluate the dynamic response characteristics of this pressure relief apparatus. The tests were performed in a facility with water simulating sodium and rising pressure pulses representative of the pressure increase resulting from a water/steam leak from a steam generator into sodium in the intermediate heat transport system of a breeder reactor power plant. Maximum source pressures ranged in magnitude from 50 psi to 800 psi. Dynamic response characteristics of each of the two rupture discs were similar to those observed in larger scale sodium-water experiments conducted in the Series I and Series II Large Leak Test Program at the Energy Technology Engineering Center. The SRI double rupture disc dynamic behavior was found to be consistent and amendable to modelling in the TRANSWRAP II computer code. A series of correlations which represent rupture disc buckling parameters were developed for use in the TRANSWRAP II code. The semi-empirical modeling of the rupture discs in the TRANSWRAP II code showed very good agreement with the experimental results

  8. YouTube and ‘psychiatry’

    Science.gov (United States)

    Gordon, Robert; Miller, John; Collins, Noel

    2015-01-01

    YouTube is a video-sharing website that is increasingly used to share and disseminate health-related information, particularly among younger people. There are reports that social media sites, such as YouTube, are being used to communicate an anti-psychiatry message but this has never been confirmed in any published analysis of YouTube clip content. This descriptive study revealed that the representation of ‘psychiatry’ during summer 2012 was predominantly negative. A subsequent smaller re-analysis suggests that the negative portrayal of ‘psychiatry’ on YouTube is a stable phenomenon. The significance of this and how it could be addressed are discussed. PMID:26755987

  9. Linguine sign in musculoskeletal imaging: calf silicone implant rupture.

    Science.gov (United States)

    Duryea, Dennis; Petscavage-Thomas, Jonelle; Frauenhoffer, Elizabeth E; Walker, Eric A

    2015-08-01

    Imaging findings of breast silicone implant rupture are well described in the literature. On MRI, the linguine sign indicates intracapsular rupture, while the presence of silicone particles outside the fibrous capsule indicates extracapsular rupture. The linguine sign is described as the thin, wavy hypodense wall of the implant within the hyperintense silicone on T2-weighted images indicative of rupture of the implant within the naturally formed fibrous capsule. Hyperintense T2 signal outside of the fibrous capsule is indicative of an extracapsular rupture with silicone granuloma formation. We present a rare case of a patient with a silicone calf implant rupture and discuss the MRI findings associated with this condition.

  10. Performance analysis of double basin solar still with evacuated tubes

    International Nuclear Information System (INIS)

    Hitesh N Panchal; Shah, P. K.

    2013-01-01

    Solar still is a very simple device, which is used for solar distillation process. In this research work, double basin solar still is made from locally available materials. Double basin solar still is made in such a way that, outer basin is exposed to sun and lower side of inner basin is directly connected with evacuated tubes to increase distillate output and reducing heat losses of a solar still. The overall size of the lower basin is about 1006 mm x 325 mm x 380 mm, the outer basin is about 1006 mm x 536 mm x 100 mm Black granite gravel is used to increase distillate output by reducing quantity of brackish or saline water in the both basins. Several experiments have conducted to determine the performance of a solar still in climate conditions of Mehsana (latitude of 23 degree 59' and longitude of 72 degree 38'), Gujarat, like a double basin solar still alone, double basin solar still with different size black granite gravel, double basin solar still with evacuated tubes and double basin solar still with evacuated tubes and different size black granite gravel. Experimental results show that, connecting evacuated tubes with the lower side of the inner basin increases daily distillate output of 56% and is increased by 60%, 63% and 67% with average 10 mm, 20 mm and 30 mm size black granite gravel. Economic analysis of present double basin solar still is 195 days. (authors)

  11. Physics of Earthquake Rupture Propagation

    Science.gov (United States)

    Xu, Shiqing; Fukuyama, Eiichi; Sagy, Amir; Doan, Mai-Linh

    2018-05-01

    A comprehensive understanding of earthquake rupture propagation requires the study of not only the sudden release of elastic strain energy during co-seismic slip, but also of other processes that operate at a variety of spatiotemporal scales. For example, the accumulation of the elastic strain energy usually takes decades to hundreds of years, and rupture propagation and termination modify the bulk properties of the surrounding medium that can influence the behavior of future earthquakes. To share recent findings in the multiscale investigation of earthquake rupture propagation, we held a session entitled "Physics of Earthquake Rupture Propagation" during the 2016 American Geophysical Union (AGU) Fall Meeting in San Francisco. The session included 46 poster and 32 oral presentations, reporting observations of natural earthquakes, numerical and experimental simulations of earthquake ruptures, and studies of earthquake fault friction. These presentations and discussions during and after the session suggested a need to document more formally the research findings, particularly new observations and views different from conventional ones, complexities in fault zone properties and loading conditions, the diversity of fault slip modes and their interactions, the evaluation of observational and model uncertainties, and comparison between empirical and physics-based models. Therefore, we organize this Special Issue (SI) of Tectonophysics under the same title as our AGU session, hoping to inspire future investigations. Eighteen articles (marked with "this issue") are included in this SI and grouped into the following six categories.

  12. Acute Pectoralis Major Rupture Captured on Video

    Directory of Open Access Journals (Sweden)

    Alejandro Ordas Bayon

    2016-01-01

    Full Text Available Pectoralis major (PM ruptures are uncommon injuries, although they are becoming more frequent. We report a case of a PM rupture in a young male who presented with axillar pain and absence of the anterior axillary fold after he perceived a snap while lifting 200 kg in the bench press. Diagnosis of PM rupture was suspected clinically and confirmed with imaging studies. The patient was treated surgically, reinserting the tendon to the humerus with suture anchors. One-year follow-up showed excellent results. The patient was recording his training on video, so we can observe in detail the most common mechanism of injury of PM rupture.

  13. Dynamic Measurements of Plastic Deformation in a Water-Filled Aluminum Tube in Response to Detonation of a Small Explosives Charge

    Directory of Open Access Journals (Sweden)

    Harold Sandusky

    1999-01-01

    Full Text Available Experiments have been conducted to benchmark computer code calculations for the dynamic interaction of explosions in water with structures. Aluminum cylinders with a length slightly more than twice their diameter were oriented vertically, sealed on the bottom by a thin plastic sheet, and filled with distilled water. An explosive charge suspended in the center of the tube plastically deformed but did not rupture the wall. Tube wall velocity, displacement, and strain were directly measured. The agreement among the three sets of dynamic data and the agreement of the terminal displacement measurements with the residual deformation were excellent.

  14. Earthquake Rupture at Focal Depth, Part I: Structure and Rupture of the Pretorius Fault, TauTona Mine, South Africa

    Science.gov (United States)

    Heesakkers, V.; Murphy, S.; Reches, Z.

    2011-12-01

    We analyze the structure of the Archaean Pretorius fault in TauTona mine, South Africa, as well as the rupture-zone that recently reactivated it. The analysis is part of the Natural Earthquake Laboratory in South African Mines (NELSAM) project that utilizes the access to 3.6 km depth provided by the mining operations. The Pretorius fault is a ~10 km long, oblique-strike-slip fault with displacement of up to 200 m that crosscuts fine to very coarse grain quartzitic rocks in TauTona mine. We identify here three structural zones within the fault-zone: (1) an outer damage zone, ~100 m wide, of brittle deformation manifested by multiple, widely spaced fractures and faults with slip up to 3 m; (2) an inner damage zone, 25-30 m wide, with high density of anastomosing conjugate sets of fault segments and fractures, many of which carry cataclasite zones; and (3) a dominant segment, with a cataclasite zone up to 50 cm thick that accommodated most of the Archaean slip of the Pretorius fault, and is regarded as the `principal slip zone' (PSZ). This fault-zone structure indicates that during its Archaean activity, the Pretorius fault entered the mature fault stage in which many slip events were localized along a single, PSZ. The mining operations continuously induce earthquakes, including the 2004, M2.2 event that rejuvenated the Pretorius fault in the NELSAM project area. Our analysis of the M2.2 rupture-zone shows that (1) slip occurred exclusively along four, pre-existing large, quasi-planer segments of the ancient fault-zone; (2) the slipping segments contain brittle cataclasite zones up to 0.5 m thick; (3) these segments are not parallel to each other; (4) gouge zones, 1-5 mm thick, composed of white `rock-flour' formed almost exclusively along the cataclasite-host rock contacts of the slipping segments; (5) locally, new, fresh fractures branched from the slipping segments and propagated in mixed shear-tensile mode; (6) the maximum observed shear displacement is 25 mm in

  15. Experimental Study of Concrete-filled Carbon Fiber Reinforced Polymer Tube with Internal Reinforcement under Axially Loading

    Directory of Open Access Journals (Sweden)

    Wenbin SUN

    2014-12-01

    Full Text Available Comparing with the circular concrete columns confined with fiber reinforced polymer (FRP wrap or tube, the rectilinear confined columns were reported much less. Due to the non-uniform distribution of confining pressure in the rectilinear confined columns, the FRP confinement effectiveness was significant reduced. This paper presents findings of an experimental program where nine prefabricated rectangular cross-section CFRP tubes with CFRP integrated crossties filled concrete to form concrete-filled FRP tube (CFFT short columns and three plain concrete control specimens were tested. All specimens were axially loaded until failure. The rest results showed that the stress-strain curves of CFFTs consisted of two distinct branches, an ascending branch before the concrete peak stress was reaches and a second branch that terminated when the tube ruptured, and that the CFFTs with integrated crossties experienced most uniform confinement pressure distribution. Test research also found that the stress-strain curves of CFFTs indicated an increase in ductility. These demonstrate that this confinement system can produce higher lateral confinement stiffness. DOI: http://dx.doi.org/10.5755/j01.ms.20.4.6035

  16. 3D analysis of thermal exchange in finned batteries. A comparison between round and elliptical tubes

    International Nuclear Information System (INIS)

    Valdiserri, P.

    2001-01-01

    In this paper a numerical 3D analysis of the thermal exchange in air-cooled finned batteries has been carried out. Speed and temperature values in each hub of the numerical simulation domain have been reckoned both at different air flows and with different shapes of the tubes. The thermal power exchanged between tubes and air is obtained by the simulation of a numerical model of a finned battery with round section tubes and is compared to the values obtained for three batteries with elliptical section tubes. The comparison has been performed for different values of the air input speed [it

  17. ACL Rupture in Collegiate Wrestler

    Directory of Open Access Journals (Sweden)

    Lindsay A. Palmer

    2016-05-01

    Full Text Available Objective: To educate others on unique Anterior Cruciate Ligament tears and percentage of usage of the ACL in normal daily function. Background: Patient is an eighteen year old male participating in wrestling and football at the time of the injury. Patient now only participates in wrestling. No previous knee or chronic injuries were reported prior to this injury. Patient was playing football during the time of injury. The patient stated that he planted his foot down and was tackled at the same time when the injury occurred. The patient felt his knee twist and buckle. Patient complained of clicking inside the knee and had minimal swelling. He also complained of it being difficult to bear weight at the time. The patient did not seek further treatment until two months after the injury occurred when he received an MRI. His MRI showed a positive finding for an Anterior Cruciate Ligament rupture. His previous Athletic Trainer could not find a positive diagnosis for the patient prior to the MRI. Differential Diagnosis: Possible meniscal or ACL injury. Treatment: Doctors officially diagnosed the injury as a complete rupture of the ACL. The patient did not receive surgery immediately. Doctors have stated that he only uses about 50% of his ACL on a daily basis compared to a normal person who uses about 95% of their ACL daily. Because of this, the patient played on his rupture for seven months before receiving surgery. He played a whole season of high school football and a whole season of wrestling his senior year with the ACL ruptured. The patient only used a brace for better comfort during the seven months. The patient then received reconstructive surgery to repair the rupture. A hamstring tendon graft was used to repair the ruptured ACL. Because a tendon was taken from the hamstring, patient experienced a tight ACL and hamstring of the left leg post-surgery. The patient participated in Physical Therapy for five months to strengthen and stretch the new

  18. Describing Soils: Calibration Tool for Teaching Soil Rupture Resistance

    Science.gov (United States)

    Seybold, C. A.; Harms, D. S.; Grossman, R. B.

    2009-01-01

    Rupture resistance is a measure of the strength of a soil to withstand an applied stress or resist deformation. In soil survey, during routine soil descriptions, rupture resistance is described for each horizon or layer in the soil profile. The lower portion of the rupture resistance classes are assigned based on rupture between thumb and…

  19. Gastrostomy Tube (G-Tube)

    Science.gov (United States)

    ... any of these problems: a dislodged tube a blocked or clogged tube any signs of infection (including redness, swelling, or warmth at the tube site; discharge that's yellow, green, or foul-smelling; fever) excessive bleeding or drainage from the tube site severe abdominal pain lasting ...

  20. 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.