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Sample records for tube rupture analysis

  1. 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

  2. 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

  3. 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.

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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)

  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. 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)

  18. 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)

  19. 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

  20. 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.

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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)

  8. 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

  9. 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.

  10. 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)

  11. 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 ...

  12. 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)

  13. 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

  14. Sensitivity Analysis of Dousing Spray Trip on Radioactive Release in Pressure Tube Rupture Accident with Both End Fitting Failures

    Energy Technology Data Exchange (ETDEWEB)

    Jang, M. S.; Kang, H. S; Kim, S. R. [NESS, Daejeon (Korea, Republic of)

    2015-10-15

    We analyzed the sensitivity analysis of dousing spray trip conditions on radioactive release. In terms of conservativeness, the set 1 trip would be more appropriate in RR analysis than set 2 trip, which is the general condition of RR analysis. Radioactive releases from the containment building is related to containment air pressure, which increases by the coolant discharge from loss of coolant accident and the actuation conditions of dousing spray and so on. In LOCA analysis, the dousing spray trip conditions are set for the analysis objectives; for peak pressure (PP), for pressure signal (PS), for radioactive release (RR) and etc. In RR analysis, we would determine the dousing spray trip condition to increase radioactive release to the public for conservatism. Therefore, we carried out the sensitivity analysis of dousing spray trip condition on radioactive release from containment building using GOTHIC and SMART program for CANDU.

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. 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

  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. 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.)

  3. 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

  4. 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

  5. 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

  6. 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

  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. 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

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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)

  14. 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

  15. 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...

  16. 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.

  17. 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

  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. 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

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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.)

  5. 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

  6. 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

  7. 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.)

  8. 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.

  9. 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

  10. 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

  11. 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)

  12. 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.)

  13. 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)

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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)

  20. 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.)

  1. 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.

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. Rupture

    CERN Multimedia

    Association du personnel

    2006-01-01

    Our Director-General is indifferent to the tradition of concertation foreseen in our statutes and is "culturally" unable to associate the Staff Association with problem-solving in staff matters. He drags his heels as long as possible before entering into negotiations, presents "often misleading" solutions at the last minute which he only accepts to change once a power struggle has been established. Faced with this rupture and despite its commitment to concertation between gentlemen. The results of the poll in which the staff is invited to participate this week. We therefore need your support to state our claims to the Governing Bodies. The Staff Association proposes a new medium of communication and thus hopes to show that it is ready for future negotiations. The pages devoted to the Staff Association are presented in a more informative, reactive and factual manner and in line with the evolution of the social situation at CERN. We want to establish strong and continuous ties between the members of CERN and ou...

  8. 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.)

  9. 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)

  10. 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.

  11. 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

  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. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.)

  19. 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.)

  20. 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)

  1. 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.

  2. 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.

  3. 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)

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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).

  9. 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

  10. 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

  11. 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)

  12. 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)

  13. 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.)

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.)

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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.

  5. 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)

  6. Analysis of the Steam Generator Tubes Rupture Initiating Event

    International Nuclear Information System (INIS)

    Trillo, A.; Minguez, E.; Munoz, R.; Melendez, E.; Sanchez-Perea, M.; Izquierd, J.M.

    1998-01-01

    In PSA studies, Event Tree-Fault Tree techniques are used to analyse to consequences associated with the evolution of an initiating event. The Event Tree is built in the sequence identification stage, following the expected behaviour of the plant in a qualitative way. Computer simulation of the sequences is performed mainly to determine the allowed time for operator actions, and do not play a central role in ET validation. The simulation of the sequence evolution can instead be performed by using standard tools, helping the analyst obtain a more realistic ET. Long existing methods and tools can be used to automatism the construction of the event tree associated to a given initiator. These methods automatically construct the ET by simulating the plant behaviour following the initiator, allowing some of the systems to fail during the sequence evolution. Then, the sequences with and without the failure are followed. The outcome of all this is a Dynamic Event Tree. The work described here is the application of one such method to the particular case of the SGTR initiating event. The DYLAM scheduler, designed at the Ispra (Italy) JRC of the European Communities, is used to automatically drive the simulation of all the sequences constituting the Event Tree. Similarly to the static Event Tree, each time a system is demanded, two branches are open: one corresponding to the success and the other to the failure of the system. Both branches are followed by the plant simulator until a new system is demanded, and the process repeats. The plant simulation modelling allows the treatment of degraded sequences that enter into the severe accident domain as well as of success sequences in which long-term cooling is started. (Author)

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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%).

  2. 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.

  3. 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 ...

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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)

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. Correlation of energy balance method to dynamic pipe rupture analysis

    International Nuclear Information System (INIS)

    Kuo, H.H.; Durkee, M.

    1983-01-01

    When using an energy balance approach in the design of pipe rupture restraints for nuclear power plants, the NRC specifies in its Standard Review Plan 3.6.2 that the input energy to the system must be multiplied by a factor of 1.1 unless a lower value can be justified. Since the energy balance method is already quite conservative, an across-the-board use of 1.1 to amplify the energy input appears unneccessary. The paper's purpose is to show that this 'correlation factor' could be substantially less than unity if certain design parameters are met. In this paper, result of nonlinear dynamic analyses were compared to the results of the corresponding analyses based on the energy balance method which assumes constant blowdown forces and rigid plastic material properties. The appropriate correlation factors required to match the energy balance results with the dynamic analyses results were correlated to design parameters such as restraint location from the break, yield strength of the energy absorbing component, and the restraint gap. It is shown that the correlation factor is related to a single nondimensional design parameter and can be limited to a value below unity if appropriate design parameters are chosen. It is also shown that the deformation of the restraints can be related to dimensionless system parameters. This, therefore, allows the maximum restraint deformation to be evaluated directly for design purposes. (orig.)

  18. 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 ...

  19. 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,

  20. 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.

  1. 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

  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. 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 ...

  4. 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 ...

  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. 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

  8. Failure analysis of tubes with wastages

    International Nuclear Information System (INIS)

    Prachuktam, S.; Reich, M.; Rajan, J.

    1979-01-01

    A finite element method for large strain calculation using the constitutive relation due to Hill was developed. This constitutive relation relates the co-rotational rate of the Kirchoff stress and deformation rate tensor which leads to a symmetric structure stiffness. This method is used to calculate failure pressures of degraded tubes

  9. 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.

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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.)

  3. 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)

  4. 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)

  5. 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

  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. 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.

  8. 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.

  9. 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

  10. 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 ...

  11. 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)

  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. 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)

  14. 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.

  15. 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.)

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. 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.)

  1. 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.

  2. 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.)

  3. 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)

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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.

  12. 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.

  13. 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

  14. Thermomechanical CSM analysis of a superheater tube in transient state

    Science.gov (United States)

    Taler, Dawid; Madejski, Paweł

    2011-12-01

    The paper presents a thermomechanical computational solid mechanics analysis (CSM) of a pipe "double omega", used in the steam superheaters in circulating fluidized bed (CFB) boilers. The complex cross-section shape of the "double omega" tubes requires more precise analysis in order to prevent from failure as a result of the excessive temperature and thermal stresses. The results have been obtained using the finite volume method for transient state of superheater. The calculation was carried out for the section of pipe made of low-alloy steel.

  15. 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.

  16. 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

  17. Dynamic analysis of complex tube systems in heat exchangers

    International Nuclear Information System (INIS)

    Kouba, J.; Dvorak, P.

    1985-01-01

    Using a computation model, a dynamic analysis was made of tube assemblies of heat exchanger bundles by the finite element method. The algorithm is presented for determining the frequency mode properties, based on the Sturm sequences combined with inverse vector iteration. The results obtained using the method are compared with those obtained by analytical solution and by the transfer matrix method, this for the cases of both eigenvibrations and resonance vibrations. The results are in very good agreement. For the first four eigenfrequencies, the calculation error is less than 1.5% as against the analytical solution. (J.B.). 4 tabs., 8 figs., 5 refs

  18. 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%

  19. 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 ...

  20. Characteristics of elongated and ruptured anterior cruciate ligament grafts: An analysis of 21 consecutive revision cases

    Directory of Open Access Journals (Sweden)

    Kohei Iio

    2017-04-01

    Conclusion: The location of the original femoral tunnel was more proximal in patients with elongated grafts than in those with ruptured grafts. Different bone tunnel position from native ACL might lead to graft elongation.

  1. 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.

  2. 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

  3. 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)

  4. Tests and analysis on steam generator tube failure propagation

    International Nuclear Information System (INIS)

    Tanabe, Hiromi

    1990-01-01

    The understanding of leak enlargement and failure propagation behavior is essential to select a design basis leak (DBL) of LMFBR steam generators. Therefore, various series of experiments, such as self-enlargement tests, target wastage tests, failure propagation tests were conducted in a wide range of leak using test facilities of SWAT at PNC/OEC. Especially, in the large leak tests, potential of overheating failure was investigated under a prototypical steam cooling condition inside target tubes. In the small leak, the difference of wastage resistivity was clarified among several tube materials such as 9-chrome steels. In regard to an analytical approach, a computer code LEAP (Leak Enlargement and Propagation) was developed on the basis of all of these experimental results. The code was used to validate the previously selected DBL of the prototype reactor, Monju, steam generator. This approach proved to be successful in spite of somewhat over-conservatism in the analysis. Moreover, LEAP clarified the effectiveness of a rapid steam dump and an enhanced leak detection system. The code improvement toward a realistic analysis is desired, however, to lessen the DBL for a future large plant and then the re-evaluation of the experimental data such as the size of secondary failure is under way. (author). 4 refs, 8 figs, 1 tab

  5. 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.

  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. 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)

  8. 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.

  9. LOFT ECC Pitot Tube and Thermocouple Rake Penetration thermal analysis

    International Nuclear Information System (INIS)

    Tolan, B.J.

    1977-01-01

    A thermal analysis of the LOFT ECC Pitot Tube and Thermocouple Rake Penetration was performed using COUPLE, a two-dimensional finite element computer code. Four transients which conservatively cover all transients the rake will be exposed to were included in this analysis in order to comply with the ASME Code Section III requirements. The transients conservatively cover hot and cold leg operation, and nuclear and nonnuclear operation. The four transients include the LOCE with ECC injection transient, the single control rod drop transient, the scram transient, and the heatup with 0 to 100% load change transient. Temperature distributions in the rake were obtained for each of the four transients and several plots of node temperatures vs. time are given

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. Identification of earthquakes that generate tsunamis in Java and Nusa Tenggara using rupture duration analysis

    Energy Technology Data Exchange (ETDEWEB)

    Pribadi, S., E-mail: sugengpribadimsc@gmail.com [Tsunami Warning Information Division, Indonesian Meteorological Climatological and Geophysical Agency (BMKG), Jalan Angkasa I No. 2, Jakarta13920 and Graduate Student of Earth Sciences, Faculty of Earth Sciences and Technology, Bandung Institute of T (Indonesia); Puspito, N. T.; Yudistira, T.; Afnimar,; Ibrahim, G. [Global Geophysics Research Group, Faculty of Mining and Petroleum Engineering, Bandung Institute of Technology (ITB), Jalan Ganesha 10, Bandung 40132 (Indonesia); Laksono, B. I. [Database Maintenance Division, Indonesian Meteorological Climatological and Geophysical Agency (BMKG), Jalan Angkasa I No.2, Jakarta 13920 (Indonesia); Adnan, Z. [Database Maintenance Division, Indonesian Meteorological Climatological and Geophysical Agency (BMKG), Jalan Angkasa I No. 2, Jakarta 13920 and Graduate Student of Earth Sciences, Faculty of Earth Sciences and Technology, Bandung Institute of Technol (Indonesia)

    2014-09-25

    Java and Nusa Tenggara are the tectonically active of Sunda arc. This study discuss the rupture duration as a manifestation of the power of earthquake-generated tsunami. We use the teleseismic (30° - 90°) body waves with high-frequency energy Seismometer is from IRIS network as amount 206 broadband units. We applied the Butterworth high bandpass (1 - 2 Hz) filtered. The arrival and travel times started from wave phase of P - PP which based on Jeffrey Bullens table with TauP program. The results are that the June 2, 1994 Banyuwangi and the July 17, 2006 Pangandaran earthquakes identified as tsunami earthquakes with long rupture duration (To > 100 second), medium magnitude (7.6 < Mw < 7.9) and located near the trench. The others are 4 tsunamigenic earthquakes and 3 inland earthquakes with short rupture duration start from To > 50 second which depend on its magnitude. Those events are located far from the trench.

  15. Thermal analysis of gyrotron traveling-wave tube collector

    International Nuclear Information System (INIS)

    Zheng Zhiqing; Luo Yong; Jiang Wei; Tang Yong

    2013-01-01

    In order to solve cooling problem of the gyrotron traveling-wave tube(TWT) collector and guarantee the gyrotron TWT's reliability and stability, the electron trajectories in the gyrotron TWT are simulated using CST electron simulation software. Thermal analysis of the collector with finite element software ANSYS is performed. The ways of applying boundary that affects the distribution of collector temperature are compared. The influence of the water temperature and flow rate on collector temperature distribution under actual heat fluxes (boundary condition) is researched. The size and number of collector fins are optimized, and a relatively perfect structure is obtained finally. The result estimated by simulation is consistent with the experiment and proves that the model and method employed in this work are suitable. (authors)

  16. 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

  17. 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

  18. 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.

  19. Inelastic analysis of finite length and depth cracked tubes

    International Nuclear Information System (INIS)

    Reich, M.; Prachuktam, S.; Gardner, D.

    1977-01-01

    Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdowns. Typical PWR steam generator units contain thousands of long straight tubes with U-bend sections. These tubes are primarily made from alloy 600 and their sizes vary between 3 / 4 '' and 7 / 8 '' (1.905 cm and 2.223 cm) in diameter with nominal thicknesses of 0.043'' to 0.053'' (0.109 cm to 0.135 cm). Since alloy 600 (and the previously used 304-SS tubes) are ductile, high toughness materials LEFM (linear elastic fracture mechanics) criteria do not apply. 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

  20. 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

  1. Analysis of Reactor Vessel Lower Head Penetration Tube Failure

    International Nuclear Information System (INIS)

    Stempniewicz, Marek

    1999-01-01

    This paper presents results of two studies, performed to investigate the behavior of the reactor vessel penetration tubes in case of relocation of molten material into the tubes. The first study is on the CORVIS drain line experiment 03/1. Results of pre-test calculations are presented, and compared to the later obtained experimental data. The timing of the drain line melting and the velocity of the debris flowing inside the drain line were predicted correctly, but the penetration depth was clearly underestimated. If the calculations are done using different correlation for the melt-to-wall convective heat transfer, the results are closer to the experiment. It cannot however be concluded that the alternative correlation is more appropriate until other uncertainties are clarified. The second study presents calculations performed for GKN Dodewaard CRD, instrument tubes and drain line. Calculations were performed to estimate whether the tubes have a chance to withstand the first attack of the melt and thus postpone vessel failure until the water in the lower plenum evaporates. Calculations were performed assuming that the melt can move into the tubes without any resistance, e.g. presence of water in the tubes was not taken into account. The results indicate that the critical penetration of the GKN vessel, which is most likely to fail, is the drain line. Results also indicate that external flooding should prevent early tube failure, at least in case of low vessel pressure. (author)

  2. Analysis of defect tubes of fast reactor heat exchanger

    International Nuclear Information System (INIS)

    Rukhlyada, N.Ya.

    2014-01-01

    The experimental Auger electron spectroscopy and X-ray diffraction microanalysis data of laboratory investigations of defect tubes of heat exchanger with sodium coolant are presented. Element distribution through depth of corrosion layers form on the side of coolant (sodium) and on the surface contacting with steam in heat exchanger tube is studied. Sodium presence through all thickness of the tube is determined. It is shown that treatment of 12Cr18N9 steel surface by plasma pulses decreases intergranular corrosion susceptibility. It is related with structural changes of surface layer (∼ 20 μm), its enrichment by chromium and formation of chromium oxide protective film [ru

  3. Identification of earthquakes that generate tsunamis in Java and Nusa Tenggara using rupture duration analysis

    International Nuclear Information System (INIS)

    Pribadi, S.; Puspito, N. T.; Yudistira, T.; Afnimar,; Ibrahim, G.; Laksono, B. I.; Adnan, Z.

    2014-01-01

    Java and Nusa Tenggara are the tectonically active of Sunda arc. This study discuss the rupture duration as a manifestation of the power of earthquake-generated tsunami. We use the teleseismic (30° - 90°) body waves with high-frequency energy Seismometer is from IRIS network as amount 206 broadband units. We applied the Butterworth high bandpass (1 - 2 Hz) filtered. The arrival and travel times started from wave phase of P - PP which based on Jeffrey Bullens table with TauP program. The results are that the June 2, 1994 Banyuwangi and the July 17, 2006 Pangandaran earthquakes identified as tsunami earthquakes with long rupture duration (To > 100 second), medium magnitude (7.6 50 second which depend on its magnitude. Those events are located far from the trench

  4. 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

  5. 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

  6. Analysis of Fan Waves in a Laboratory Model Simulating the Propagation of Shear Ruptures in Rocks

    Science.gov (United States)

    Tarasov, B. G.; Sadovskii, V. M.; Sadovskaya, O. V.

    2017-12-01

    The fan-shaped mechanism of rotational motion transmission in a system of elastically bonded slabs on flat surface, simulating the propagation of shear ruptures in super brittle rocks, is analyzed. Such ruptures appear in the Earth's crust at seismogenic depths. They propagate due to the nucleation of oblique tensile microcracks, leading to the formation of a fan domino-structure in the rupture head. A laboratory physical model was created which demonstrates the process of fan-structure wave propagation. Equations of the dynamics of rotational motion of slabs as a mechanical system with a finite number of degrees of freedom are obtained. Based on the Merson method of solving the Cauchy problem for systems of ordinary differential equations, the computational algorithm taking into account contact interaction of slabs is developed. Within the framework of a simplified mathematical model of dynamic behavior of a fan-shaped system in the approximation of a continuous medium, the approximate estimates of the length of a fan depending on the velocity of its motion are obtained. It is shown that in the absence of friction a fan can move with any velocity that does not exceed the critical value, which depends on the size, the moment of inertia of slabs, the initial angle and the elasticity coefficient of bonds. In the presence of friction a fan stops. On the basis of discrete and continuous models, the main qualitative features of the behavior of a fan-structure moving under the action of applied tangential forces, whose values in a laboratory physical model are regulated by a change in the inclination angle of the rupture plane, are analyzed. Comparison of computations and laboratory measurements and observations shows good correspondence between the results.

  7. 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.

  8. 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)

  9. 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

  10. 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

  11. 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.

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. 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

  17. Analysis of the Kaplan turbine draft tube effect

    Energy Technology Data Exchange (ETDEWEB)

    Motycak, L; Skotak, A; Obrovsky, J, E-mail: motycak.vhs@cbeng.c [CKD Blansko Engineering, a.s., Capkova 2357/5, Blansko 67801 (Czech Republic)

    2010-08-15

    The aim of this paper is to present information about possible problems and errors which can appear during numerical analyses of low head Kaplan turbines with a view to the runner - draft tube interaction. The setting of numerical model, grid size, used boundary conditions are the interface definition between runner and draft tube are discussed. There are available data from physical model tests which gives a great opportunity to compare CFD and experiment results and on the basis of this comparison to determine the approach to the CFD flow modeling. The main purpose for the Kaplan turbine model measurement was to gather the information about real flow field. The model tests were carried out in new hydraulic laboratory of CKD Blansko Engineering. The model tests were focused on the detailed velocity measurements downstream of the runner by differential pressure probe and on the velocity measurement downstream of the draft tube elbow by Particle Image Velocimetry method (PIV). The data from CFD simulation were compared to the velocity measurement results. In the paper also the design of the original draft tube modification due to flow improvement is discussed in the case of the Kaplan turbine uprating project. The results of the draft tube modification were confirmed by model tests in the hydraulic laboratory as well.

  18. Analysis of the Kaplan turbine draft tube effect

    Science.gov (United States)

    Motycak, L.; Skotak, A.; Obrovsky, J.

    2010-08-01

    The aim of this paper is to present information about possible problems and errors which can appear during numerical analyses of low head Kaplan turbines with a view to the runner - draft tube interaction. The setting of numerical model, grid size, used boundary conditions are the interface definition between runner and draft tube are discussed. There are available data from physical model tests which gives a great opportunity to compare CFD and experiment results and on the basis of this comparison to determine the approach to the CFD flow modeling. The main purpose for the Kaplan turbine model measurement was to gather the information about real flow field. The model tests were carried out in new hydraulic laboratory of CKD Blansko Engineering. The model tests were focused on the detailed velocity measurements downstream of the runner by differential pressure probe and on the velocity measurement downstream of the draft tube elbow by Particle Image Velocimetry method (PIV). The data from CFD simulation were compared to the velocity measurement results. In the paper also the design of the original draft tube modification due to flow improvement is discussed in the case of the Kaplan turbine uprating project. The results of the draft tube modification were confirmed by model tests in the hydraulic laboratory as well.

  19. Analysis of the Kaplan turbine draft tube effect

    International Nuclear Information System (INIS)

    Motycak, L; Skotak, A; Obrovsky, J

    2010-01-01

    The aim of this paper is to present information about possible problems and errors which can appear during numerical analyses of low head Kaplan turbines with a view to the runner - draft tube interaction. The setting of numerical model, grid size, used boundary conditions are the interface definition between runner and draft tube are discussed. There are available data from physical model tests which gives a great opportunity to compare CFD and experiment results and on the basis of this comparison to determine the approach to the CFD flow modeling. The main purpose for the Kaplan turbine model measurement was to gather the information about real flow field. The model tests were carried out in new hydraulic laboratory of CKD Blansko Engineering. The model tests were focused on the detailed velocity measurements downstream of the runner by differential pressure probe and on the velocity measurement downstream of the draft tube elbow by Particle Image Velocimetry method (PIV). The data from CFD simulation were compared to the velocity measurement results. In the paper also the design of the original draft tube modification due to flow improvement is discussed in the case of the Kaplan turbine uprating project. The results of the draft tube modification were confirmed by model tests in the hydraulic laboratory as well.

  20. Condensation Analysis of Steam/Air Mixtures in Horizontal Tubes

    International Nuclear Information System (INIS)

    Lee, Kwon Yeong; Bae, Sung Won; Kim, Moo Hwan

    2008-01-01

    Perhaps the most common flow configuration in which a convective condensation occurs is a flow in a horizontal circular tube. This configuration is encountered in air-conditioning and refrigeration condensers as well as condensers in Rankine power cycles. Although a convective condensation is also sometimes contrived to occur in a co-current vertical downward flow, a horizontal flow is often preferred because the flow can be repeatedly passed through the heat exchanger core in a serpentine fashion without trapping liquid or vapor in the return bends. Many researchers have investigated a in-tube condensation for horizontal heat exchangers. However, almost all of them obtained tube section-averaged data without a noncondensable gas. Recently, Wu and Vierow have experimentally studied the condensation of steam in a horizontal heat exchanger with air present. In order to measure the condenser tube inner surface temperatures and to calculate the local heat fluxes, they developed an innovative thermocouple design that allowed for nonintrusive measurements. Here we developed a theoretical model using the heat and mass analogy to analyze a steam condensation with a noncondensable gas in horizontal tubes

  1. 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.

  2. 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)

  3. 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...

  4. Analysis of independent failure assumptions on postulated secondary high energy line ruptures

    International Nuclear Information System (INIS)

    Hollingsworth, S.D.

    1977-01-01

    Postulated ruptures of the main steam piping in pressurized water reactors result in large amounts of steam being removed from the secondary system. Since the energy removal rate could be many times that of nominal design power, there may be a rapid cooldown of the primary coolant system and a positive addition of reactivity to the reactor core. The Westinghouse protection system design concept incorporates features that trip the reactor, isolate the main steamlines and provide for automatic alternate shutdown capability in the form of boric acid solution injection into the primary coolant system. At the most limiting time in life (end of life) the reactivity calculated to be inserted by the cooldown is sufficient to overcome the shutdown margin predicted to be available from control rods with the most reactive rod in the fully withdrawn position. Because the boron injected into the core may be delayed due to system responses, there is potential that the reactor core could return critical and return to power. The extremely adverse radial power distributions caused by the fully withdrawn control rod causes localized high power densities that could lead to reduced heat transfer capability (DNB). Because of the large amount of stored energy in the reactor coolant system at full power, the cooldown and subsequent return to power is more severe when calculated from a shutdown, hot zero power condition. It is shown that the protection system design has large margins to protect against adverse core effects following a steamline rupture

  5. Analysis of Tube Drawing Process – A Finite Element Approach ...

    African Journals Online (AJOL)

    In this paper the effect of die semi angle on drawing load in cold tube drawing has been investigated numerically using the finite element method. The equation governing the stress distribution was derived and solved using Galerkin finite element method. An isoparametric formulation for the governing equation was utilized ...

  6. Stress analysis of steam generator row-1 tubes

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Ryu, Woo Seog; Lee, Ho Jin; Kim, Sung Chung

    2000-01-01

    Residual stresses induced in U-bending and tube-to-tubesheet joining processes of PWR's steam generator row-1 tube were measured by X-ray method and Hole-Drilling Method(HDM). The stresses resulting from the internal pressure and the temperature gradient in the steam generator were also estimated theoretically. In U-bent regions, the residual stresses at extrados were induced with compressive stress(-), and its maximum value reached -319 Mpa in axial direction at ψ=0 .deg. in position. Maximum tensile residual stress of 170 MPa was found to be at the flank side at position of ψ=90 deg., i.e., at apex region. In tube-to-tubesheet joining methods, the residual stresses induced by the explosive joint method were found to be lower than that by the mechanical roll method. The gradient of residual stress along the expanded tube was highest at the transition region, and the residual stress in circumferential direction was found to be higher than the residual stress in axial direction. Hoop stress due to an internal pressure between primary and secondary side was analyzed to be 76 MPa and thermal stress was 45 MPa

  7. 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)

  8. 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

  9. Numerical Methods for the Analysis of Power Transformer Tank Deformation and Rupture Due to Internal Arcing Faults.

    Science.gov (United States)

    Yan, Chenguang; Hao, Zhiguo; Zhang, Song; Zhang, Baohui; Zheng, Tao

    2015-01-01

    Power transformer rupture and fire resulting from an arcing fault inside the tank usually leads to significant security risks and serious economic loss. In order to reveal the essence of tank deformation or explosion, this paper presents a 3-D numerical computational tool to simulate the structural dynamic behavior due to overpressure inside transformer tank. To illustrate the effectiveness of the proposed method, a 17.3 MJ and a 6.3 MJ arcing fault were simulated on a real full-scale 360MVA/220kV oil-immersed transformer model, respectively. By employing the finite element method, the transformer internal overpressure distribution, wave propagation and von-Mises stress were solved. The numerical results indicate that the increase of pressure and mechanical stress distribution are non-uniform and the stress tends to concentrate on connecting parts of the tank as the fault time evolves. Given this feature, it becomes possible to reduce the risk of transformer tank rupture through limiting the fault energy and enhancing the mechanical strength of the local stress concentrative areas. The theoretical model and numerical simulation method proposed in this paper can be used as a substitute for risky and costly field tests in fault overpressure analysis and tank mitigation design of transformers.

  10. Numerical Methods for the Analysis of Power Transformer Tank Deformation and Rupture Due to Internal Arcing Faults

    Science.gov (United States)

    Yan, Chenguang; Hao, Zhiguo; Zhang, Song; Zhang, Baohui; Zheng, Tao

    2015-01-01

    Power transformer rupture and fire resulting from an arcing fault inside the tank usually leads to significant security risks and serious economic loss. In order to reveal the essence of tank deformation or explosion, this paper presents a 3-D numerical computational tool to simulate the structural dynamic behavior due to overpressure inside transformer tank. To illustrate the effectiveness of the proposed method, a 17.3MJ and a 6.3MJ arcing fault were simulated on a real full-scale 360MVA/220kV oil-immersed transformer model, respectively. By employing the finite element method, the transformer internal overpressure distribution, wave propagation and von-Mises stress were solved. The numerical results indicate that the increase of pressure and mechanical stress distribution are non-uniform and the stress tends to concentrate on connecting parts of the tank as the fault time evolves. Given this feature, it becomes possible to reduce the risk of transformer tank rupture through limiting the fault energy and enhancing the mechanical strength of the local stress concentrative areas. The theoretical model and numerical simulation method proposed in this paper can be used as a substitute for risky and costly field tests in fault overpressure analysis and tank mitigation design of transformers. PMID:26230392

  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. 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. 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

  14. RELAP5 analysis of reflux condensation behavior in heat transfer tube bundle of a steam generator

    International Nuclear Information System (INIS)

    Minami, Noritoshi; Chikusa, Toshiaki; Nagae, Takashi; Murase, Michio

    2007-01-01

    In case of loss of the residual heat removal system and other alternative cooling methods under mid-loop operation during shutdown of the pressurized water reactor plant, reflux condensation in the steam generator (SG) may be an effective heat removal mechanism. In reflux condensation experiments 7.2c with injection of nitrogen gas using the BETHSY facility in France, which is a scale model of a pressurized water reactor plant, 34 heat transfer tubes were divided into two kinds of flow patterns, which were steam forward flow and nitrogen reverse flow. In this study, we simulated the BETHSY experiments using the transient analysis code RELAP5. Modifying calculation equations for interfacial friction force and wall friction force between the inlet plenum and heat transfer tubes, nitrogen reverse flow was successfully simulated. In calculations with alteration of the flow area ratio to two flow channels for the heat transfer tube bundle, the number of active tubes with the maximum nitrogen recirculation flow rate agreed rather well with the observed number of active tubes. In calculations with three flow channels for the heat transfer tube bundle, the average number of active tubes in several calculations with different flow area ratios of the three flow channels predicted the number of active tubes well. (author)

  15. Failure Analysis of Alumina Reinforced Aluminum Microtruss and Tube Composites

    Science.gov (United States)

    Chien, Hsueh Fen (Karen)

    The energy absorption capacity of cellular materials can be dramatically increased by applying a structural coating. This thesis examined the failure mechanisms of alumina reinforced 3003 aluminum alloy microtrusses and tubes. Alumina coatings were produced by hard anodizing and by plasma electrolytic oxidation (PEO). The relatively thin and discontinuous oxide coating at the hinge acted as a localized weak spot which triggered a chain reaction of failure, including oxide fracture, oxide spallation, oxide penetration to the aluminum core and severe local plastic deformation of the core. For the PEO microtrusses, delamination occurred within the oxide coating resulting in a global strut buckling failure mode. A new failure mode for the anodized tubes was observed: (i) axisymmetric folding of the aluminum core, (ii) longitudinal fracture, and (iii) alumina pulverization. Overall, the alumina coating enhanced the buckling resistance of the composites, while the aluminum core supported the oxide during the damage propagation.

  16. On the Quantitative Analysis of Liquid Flow in Physiological Tubes.

    Science.gov (United States)

    1982-12-01

    cri- copharyngeal sphincter which is aided by skeletal muscle (Vantrap- pen and hellemans, 1980) relaxes to accept the bolus and the gastro - esophageal ...lower ( gastro -) esophageal junction during peristalsis resulting from the interaction of gastric, esophageal and thoracic pressures. PIP is a pressure...higher than the downstream pressure and a flow velocity profile with no reflux (syn.: retropulsion). The 5 Pumping in Biological Tubes a. Peristaltic

  17. Analysis of Ketones by Selected Ion Flow Tube Mass Spectrometry

    Czech Academy of Sciences Publication Activity Database

    Smith, D.; Wang, T.; Španěl, Patrik

    2003-01-01

    Roč. 17, - (2003), s. 2655-2660 ISSN 0951-4198 R&D Projects: GA ČR GA202/03/0827; GA ČR GA203/02/0737 Institutional research plan: CEZ:AV0Z4040901 Keywords : mass spectrometry * selected ion flow tube * ketones Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 2.789, year: 2003

  18. 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

  19. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    International Nuclear Information System (INIS)

    Bucholz, J.A.

    2000-01-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source

  20. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    2000-07-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

  1. The single chest tube versus double chest tube application after pulmonary lobectomy: A systematic review and meta-analysis

    Directory of Open Access Journals (Sweden)

    Xuefei Zhang

    2016-01-01

    Conclusion: Compared with the double chest tube, the single chest tube significantly decreases amount of drainage, duration of chest tube drainage, pain score, the number of patients who need thoracentesis, and cost. Although there is convincing evidence to confirm the results mentioned herein, they still need to be confirmed by large-sample, multicenter, randomized, controlled trials.

  2. 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

  3. 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

  4. 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)

  5. Analysis of reactor installation accident WWER-1000 caused by collector head rupture of steam generator

    International Nuclear Information System (INIS)

    Sviridenko, I.I.

    2005-01-01

    System classification of passive heat defence of nuclear dangerous objects with usage of heat pipes. The article is continuation of series of works devoted to research and development of promising passive cooling system of nuclear dangerous objects which use evaporating-condensing heat transfer devices of closed type - heat pipes and two-phase thermosyphons. Classification of autonomous systems of passive heat defence with usage low temperature heat tubes is given. Base classificational lectures are observed. Advantages n disadvantages of various circuit (scheme) solutions are analyzed

  6. A comparative Thermal Analysis of conventional parabolic receiver tube and Cavity model tube in a Solar Parabolic Concentrator

    Science.gov (United States)

    Arumugam, S.; Ramakrishna, P.; Sangavi, S.

    2018-02-01

    Improvements in heating technology with solar energy is gaining focus, especially solar parabolic collectors. Solar heating in conventional parabolic collectors is done with the help of radiation concentration on receiver tubes. Conventional receiver tubes are open to atmosphere and loose heat by ambient air currents. In order to reduce the convection losses and also to improve the aperture area, we designed a tube with cavity. This study is a comparative performance behaviour of conventional tube and cavity model tube. The performance formulae were derived for the cavity model based on conventional model. Reduction in overall heat loss coefficient was observed for cavity model, though collector heat removal factor and collector efficiency were nearly same for both models. Improvement in efficiency was also observed in the cavity model’s performance. The approach towards the design of a cavity model tube as the receiver tube in solar parabolic collectors gave improved results and proved as a good consideration.

  7. A Systematic Review and Meta-Analysis on Economic Comparison Between Endovascular Coiling Versus Neurosurgical Clipping for Ruptured Intracranial Aneurysms.

    Science.gov (United States)

    Zhang, Xiaoxi; Li, Li; Hong, Bo; Xu, Yi; Liu, Yuan; Huang, Qinghai; Liu, Jianmin

    2018-05-01

    Healthcare expenditures and cost reduction have been under critical surveillance in all countries and are critical for policymakers. This review aims at qualitatively and quantitatively analyzing the difference of hospital costs and length of stay between endovascular coiling versus neurosurgical clipping in ruptured intracranial aneurysms (RAs). MEDLINE, the Cochrane database, Embase, and the Web of Science database were searched and evaluated independently by 2 authors according to the Newcastle-Ottawa Scale for cohort studies describing economic hospital cost or length of stay in patients with RAs. A total of 8 studies were included, describing 24,219 RAs treated with neurosurgical clipping and 24,962 RAs with endovascular coiling. Meta-analysis revealed that the total hospital costs (THCs) were similar between coiling versus clipping in RAs (standard mean difference [SMD], -0.05; 95% confidence interval [CI], -0.12 to 0.22; I 2  = 99%; P = 0.50). Subgroup analysis showed that THCs of clipping and coiling were similar in ruptured aneurysms in the United States. However, in South Korea, the THCs of coiling were significantly higher than clipping. In the long run, 1-year medical costs of endovascular treatment were significantly lower than that of clipping in RAs (SMD, 0.15; 95% CI, 0.05-0.25; I 2  = 66%; P = 0.005). In addition, the length of stay of coiled patients was significantly shorter than clipped patients (SMD, 0.29; 95% CI, 0.13-0.45; I 2  = 96%; P China, coiling was more expensive. The length of stay was much shorter in coiled patients in all countries. Copyright © 2018. Published by Elsevier Inc.

  8. Meteorology in ruptured abdominal aortic aneurysm: an institutional study and a meta-analysis of published studies reporting atmospheric pressure.

    Science.gov (United States)

    Takagi, H; Watanabe, T; Mizuno, Y; Kawai, N; Umemoto, T

    2014-12-01

    The aim of this paper was to determine whether weather factors including atmospheric pressure are associated with the occurrence of ruptured abdominal aortic aneurysm (RAAA). We investigated our institutional experiences of RAAA in more than 150 patients during 8 years. Further, we performed a meta-analysis of published studies reporting the influence of atmospheric pressure on RAAA. We retrospectively evaluated 152 patients who underwent surgery for RAAA (including ruptured iliac arterial aneurysm) at our institute between 1 January 2006 and 31 December 2013. Daily regional meteorological data (in the nearest weather station located 3.5 km from the hospital) were obtained online from Japan Meteorological Agency. To identify comparative studies of mean atmospheric pressure on the day with RAAA versus that on the day without RAAA, MEDLINE and EMBASE were searched through January 2014 using Web-based search engines (PubMed and OVID). Mean sea level atmospheric pressure, delta mean atmospheric pressure (difference between mean sea level atmospheric pressure on the day and that on the previous day), and sunshine duration on the day with RAAA were significantly lower than those on the day without RAAA: 1012.43±7.44 versus 1013.71±6.49 hPa, P=0.039, -1.18±5.15 versus 0.05±5.62 hPa, P=0.005; and 4.76±3.76 versus 5.47±3.88 h, P=0.026; respectively. A pooled analysis of 8 studies (including our institutional study) demonstrated that mean atmospheric pressure on the day with RAAA was significantly lower than that on the day without RAAA: standardized mean difference, -0.09; 95% confidence interval, -0.14 to -0.04; P=0.0009. Atmospheric pressure on the day with RAAA appears lower than that on the day without RAAA. Atmospheric pressure may be associated with the occurrence of RAAA.

  9. 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.

  10. 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)

  11. 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.

  12. Shell and Double Concentric Tube Heat Exchanger Calculations and Analysis

    Directory of Open Access Journals (Sweden)

    Basma Abbas Abdulmajeed

    2015-01-01

    Full Text Available This study concerns a new type of heat exchangers, which is that of shell-and-double concentric tube heat exchangers. The case studies include both design calculations and performance calculations. The new heat exchanger design was conducted according to Kern method. The volumetric flow rates were 3.6 m3/h and 7.63 m3/h for the hot oil and water respectively. The experimental parameters studied were: temperature, flow rate of hot oil, flow rate of cold water and pressure drop. A comparison was made for the theoretical and experimental results and it was found that the percentage error for the hot oil outlet temperature was (- 1.6%. The percentage errors for the pressure drop in the shell and in the concentric tubes were (17.2% and (- 39% respectively. For cold water outlet temperature, the percentage error was (- 3.3%, while it was (18% considering the pressure drop in the annulus formed. The percentage error for the total power consumed was (-10.8% A theoretical comparison was made between the new design and the conventional heat exchanger from the point of view of, length, mass, pressure drop and total power consumed.

  13. Economic analysis comparing induction of labor and expectant management in women with preterm prelabor rupture of membranes between 34 and 37 weeks (PPROMEXIL trial)

    NARCIS (Netherlands)

    Vijgen, S.M.; Ham, D.P. van der; Bijlenga, D.; Beek, J.J. van; Bloemenkamp, K.W.; Kwee, A.; Groenewout, M.; Kars, M.M.; Kuppens, S.; Mantel, G.; Molkenboer, J.F.; Mulder, A.L.; Nijhuis, J.G.; Pernet, P.J.; Porath, M.; Woiski, M.D.; Weinans, M.J.; Wijngaarden, W.J. van; Wildschut, H.I.J.; Akerboom, B.; Sikkema, J.M.; Willekes, C.; Mol, B.W.; Opmeer, B.C.; et al.,

    2014-01-01

    OBJECTIVE: To compare the costs of induction of labor and expectant management in women with preterm prelabor rupture of membranes (PPROM). DESIGN: Economic analysis based on a randomized clinical trial. SETTING: Obstetric departments of eight academic and 52 non-academic hospitals in the

  14. Economic analysis comparing induction of labor and expectant management in women with preterm prelabor rupture of membranes between 34 and 37 weeks (PPROMEXIL trial)

    NARCIS (Netherlands)

    Vijgen, Sylvia M. C.; Van der Ham, David P.; Bijlenga, Denise; Van Beek, Johannes J.; Bloemenkamp, Kitty W. M.; Kwee, Anneke; Groenewout, Mariet; Kars, Michael M.; Kuppens, Simone; Mantel, Gerald; Molkenboer, Jan F. M.; Mulder, Antonius L. M.; Nijhuis, Jan G.; Pernet, Paula J. M.; Porath, Martina; Woiski, Mallory D.; Weinans, Martin J. N.; Van Wijngaarden, Wim J.; Wildschut, Hajo I. J.; Akerboom, Bertina; Sikkema, J. Marko; Willekes, Christine; Mol, Ben W. J.; Opmeer, Brent C.

    ObjectiveTo compare the costs of induction of labor and expectant management in women with preterm prelabor rupture of membranes (PPROM). DesignEconomic analysis based on a randomized clinical trial. SettingObstetric departments of eight academic and 52 non-academic hospitals in the Netherlands.

  15. Economic analysis comparing induction of labor and expectant management in women with preterm prelabor rupture of membranes between 34 and 37 weeks (PPROMEXIL trial)

    NARCIS (Netherlands)

    Vijgen, Sylvia M. C.; van der Ham, David P.; Bijlenga, Denise; van Beek, Johannes J.; Bloemenkamp, Kitty W. M.; Kwee, Anneke; Groenewout, Mariët; Kars, Michael M.; Kuppens, Simone; Mantel, Gerald; Molkenboer, Jan F. M.; Mulder, Antonius L. M.; Nijhuis, Jan G.; Pernet, Paula J. M.; Porath, Martina; Woiski, Mallory D.; Weinans, Martin J. N.; van Wijngaarden, Wim J.; Wildschut, Hajo I. J.; Akerboom, Bertina; Sikkema, J. Marko; Willekes, Christine; Mol, Ben W. J.; Opmeer, Brent C.

    2014-01-01

    To compare the costs of induction of labor and expectant management in women with preterm prelabor rupture of membranes (PPROM). Economic analysis based on a randomized clinical trial. Obstetric departments of eight academic and 52 non-academic hospitals in the Netherlands. Women with PPROM near

  16. 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)

  17. 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

  18. Bayes Analysis and Reliability Implications of Stress-Rupture Testing a Kevlar/Epoxy COPV Using Temperature and Pressure Acceleration

    Science.gov (United States)

    Phoenix, S. Leigh; Kezirian, Michael T.; Murthy, Pappu L. N.

    2009-01-01

    Composite Overwrapped Pressure Vessels (COPVs) that have survived a long service time under pressure generally must be recertified before service is extended. Flight certification is dependent on the reliability analysis to quantify the risk of stress rupture failure in existing flight vessels. Full certification of this reliability model would require a statistically significant number of lifetime tests to be performed and is impractical given the cost and limited flight hardware for certification testing purposes. One approach to confirm the reliability model is to perform a stress rupture test on a flight COPV. Currently, testing of such a Kevlar49 (Dupont)/epoxy COPV is nearing completion. The present paper focuses on a Bayesian statistical approach to analyze the possible failure time results of this test and to assess the implications in choosing between possible model parameter values that in the past have had significant uncertainty. The key uncertain parameters in this case are the actual fiber stress ratio at operating pressure, and the Weibull shape parameter for lifetime; the former has been uncertain due to ambiguities in interpreting the original and a duplicate burst test. The latter has been uncertain due to major differences between COPVs in the database and the actual COPVs in service. Any information obtained that clarifies and eliminates uncertainty in these parameters will have a major effect on the predicted reliability of the service COPVs going forward. The key result is that the longer the vessel survives, the more likely the more optimistic stress ratio model is correct. At the time of writing, the resulting effect on predicted future reliability is dramatic, increasing it by about one "nine," that is, reducing the predicted probability of failure by an order of magnitude. However, testing one vessel does not change the uncertainty on the Weibull shape parameter for lifetime since testing several vessels would be necessary.

  19. 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)

  20. 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)

  1. 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.

  2. 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.

  3. Contribution to the heat transfer analysis of substitute refrigerants in evaporator tubes with smooth or enhanced tube surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Kattan, N

    1997-12-31

    The substitution of CFC refrigerants in refrigeration systems, heat pumps and organic Rankine cycles for heat recovery, requests a good knowledge of heat transfer properties of substitute fluids. A new test facility has been built at the Laboratory for Industrial Energy Systems (LENI) to contribute to this international effort. It consists of two sets of concentric tubes allowing either annular or inside tube convective boiling with a counter current water flow heating to be studied. A new data base including heat transfer coefficients and pressure drop measurements for four new refrigerants (R123, R134A, R402A and R404A) and three older refrigerants (R11, R12 and R502) has been collected. Flow boiling measurements covered a broad range of mass velocities, vapor qualities and heat fluxes. Some of the tests included plain tubes and others enhanced surface tubes (microfilms from Wieland) in horizontal and vertical orientations. An improved Wilson plot technique, that covers both the transition and turbulent flow regimes of the water flowing in the annular channel for the inside tube boiling tests, is proposed to overcome the severe limitations of conventional Wilson plots, to improve accuracy and to facilitate data processing. Mean flow boiling heat transfer coefficients were measured for R12 and R134A evaporating inside a horizontal plain tube and for R11 and R123 evaporating inside a horizontal plain tube. Local flow boiling heat transfer coefficients were measured for : R134A, R123, R404A and R502 evaporating inside a horizontal plain tube, for R134A and R123 evaporating inside a horizontal microfin tube and for R134 evaporating inside a vertical microfin tube. In addition microfin heat transfer augmentation relative to plain tube test data was investigated. The measured heat transfer coefficients were compared to different existing inside tube flow boiling correlations. (author) figs., tabs., refs.

  4. 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

  5. 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.

  6. Numerical Analysis on the Compressible Flow Characteristics of Supersonic Jet Caused by High-Pressure Pipe Rupture Using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong-Kil; Yoon, Jun-Kyu [Gachon Univ., Sungnam (Korea, Republic of); Kim, Kwang-Chu [KEPCO-E& C, Kimchun (Korea, Republic of)

    2017-10-15

    A rupture in a high-pressure pipe causes the fluid in the pipe to be discharged in the atmosphere at a high speed resulting in a supersonic jet that generates the compressible flow. This supersonic jet may display complicated and unsteady behavior in general . In this study, Computational Fluid Dynamics (CFD) analysis was performed to investigate the compressible flow generated by a supersonic jet ejected from a high-pressure pipe. A Shear Stress Transport (SST) turbulence model was selected to analyze the unsteady nature of the flow, which depends upon the various gases as well as the diameter of the pipe. In the CFD analysis, the basic boundary conditions were assumed to be as follows: pipe of diameter 10 cm, jet pressure ratio of 5, and an inlet gas temperature of 300 K. During the analysis, the behavior of the shockwave generated by a supersonic jet was observed and it was found that the blast wave was generated indirectly. The pressure wave characteristics of hydrogen gas, which possesses the smallest molecular mass, showed the shortest distance to the safety zone. There were no significant difference observed for nitrogen gas, air, and oxygen gas, which have similar molecular mass. In addition, an increase in the diameter of the pipe resulted in the ejected impact caused by the increased flow rate to become larger and the zone of jet influence to extend further.

  7. Non-linear analysis up to rupture of a model of a multi-cavity prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Rebora, B.; Uffer, F.; Zimmermann, T.

    1977-01-01

    Within the frame of a German-Swiss agreement concerning the project of a high-temperature nuclear plant (HHT), the Swiss Federal Institute for Reactor Research (EIR, in Wuerlingen) has developed an integrated variant of an helium-cooled high temperature reactor of 3x500 Mwe. A test on a model (1:20) of this prestressed concrete nuclear vessel with multiple cavities has been carried out under the supervision of 'Bonnard et Gardel ingenieurs-conseils SA (BG). The aim of this analysis is to determine the mechanism of ruin and ultimate load of the structure. In addition, comparison with the results of the test emphasizes the mathematical model with a view to its utilisation for the analysis of any prestressed concrete nuclear vessel. The principal interest of this paper is to show the accuracy of non-linear analysis of a complex massive structure with the test results and the evolution of the behaviour of the structure from the apparition of the first crack up to the ruin by rupture of the steel wires. (Auth.)

  8. 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

  9. 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.

  10. 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.

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

    NARCIS (Netherlands)

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

    2017-01-01

    Background: 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. Methods:

  12. 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

  13. 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.

  14. 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.

  15. Single-cell analysis of pyroptosis dynamics reveals conserved GSDMD-mediated subcellular events that precede plasma membrane rupture.

    Science.gov (United States)

    de Vasconcelos, Nathalia M; Van Opdenbosch, Nina; Van Gorp, Hanne; Parthoens, Eef; Lamkanfi, Mohamed

    2018-04-17

    Pyroptosis is rapidly emerging as a mechanism of anti-microbial host defense, and of extracellular release of the inflammasome-dependent cytokines interleukin (IL)-1β and IL-18, which contributes to autoinflammatory pathology. Caspases 1, 4, 5 and 11 trigger this regulated form of necrosis by cleaving the pyroptosis effector gasdermin D (GSDMD), causing its pore-forming amino-terminal domain to oligomerize and perforate the plasma membrane. However, the subcellular events that precede pyroptotic cell lysis are ill defined. In this study, we triggered primary macrophages to undergo pyroptosis from three inflammasome types and recorded their dynamics and morphology using high-resolution live-cell spinning disk confocal laser microscopy. Based on quantitative analysis of single-cell subcellular events, we propose a model of pyroptotic cell disintegration that is initiated by opening of GSDMD-dependent ion channels or pores that are more restrictive than recently proposed GSDMD pores, followed by osmotic cell swelling, commitment of mitochondria and other membrane-bound organelles prior to sudden rupture of the plasma membrane and full permeability to intracellular proteins. This study provides a dynamic framework for understanding cellular changes that occur during pyroptosis, and charts a chronological sequence of GSDMD-mediated subcellular events that define pyroptotic cell death at the single-cell level.

  16. Thermographic image analysis for classification of ACL rupture disease, bone cancer, and feline hyperthyroid, with Gabor filters

    Science.gov (United States)

    Alvandipour, Mehrdad; Umbaugh, Scott E.; Mishra, Deependra K.; Dahal, Rohini; Lama, Norsang; Marino, Dominic J.; Sackman, Joseph

    2017-05-01

    Thermography and pattern classification techniques are used to classify three different pathologies in veterinary images. Thermographic images of both normal and diseased animals were provided by the Long Island Veterinary Specialists (LIVS). The three pathologies are ACL rupture disease, bone cancer, and feline hyperthyroid. The diagnosis of these diseases usually involves radiology and laboratory tests while the method that we propose uses thermographic images and image analysis techniques and is intended for use as a prescreening tool. Images in each category of pathologies are first filtered by Gabor filters and then various features are extracted and used for classification into normal and abnormal classes. Gabor filters are linear filters that can be characterized by the two parameters wavelength λ and orientation θ. With two different wavelength and five different orientations, a total of ten different filters were studied. Different combinations of camera views, filters, feature vectors, normalization methods, and classification methods, produce different tests that were examined and the sensitivity, specificity and success rate for each test were produced. Using the Gabor features alone, sensitivity, specificity, and overall success rates of 85% for each of the pathologies was achieved.

  17. 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

  18. An investigation on compression strength analysis of commercial aluminium tube to aluminium 2025 tube plate by using TIG welding process

    Energy Technology Data Exchange (ETDEWEB)

    Kannan, S., E-mail: kannan.dgl201127@gmail.com [Department of Mechanical Engineering and Mining Machinery Engineering, Indian Institute of Technology (ISM), Dhanbad, Jharkhand, India, 826004 (India); Senthil Kumaran, S., E-mail: sskumaran@ymail.com [Research and Development Center, Department of Mechanical Engineering, RVS Educational Trust' s Group of Institutions, RVS School of Engineering and Technology, Dindigul, Tamilnadu, India, 624005 (India); Kumaraswamidhas, L.A., E-mail: lakdhas1978@gmail.com [Department of Mechanical Engineering and Mining Machinery Engineering, Indian School of Mines University, Dhanbad, Jharkhand, India, 826004 (India)

    2016-05-05

    In this present study, Tungsten inert gas (TIG) welding was applied to weld the dissimilar materials and authenticate the mechanical and metallurgical properties of tube to tube plate made up of commercial aluminium and Al 2025 respectively using an Zirconiated tungsten electrode along with filler material aluminium ER 2219. In total, twenty five pieces has been subjected to compression strength and hardness value to evaluate the optimal joint strength. The three optimization technique has been used in this experiment. Taguchi L{sub 25} orthogonal array is used to identify the most influencing process parameter which affects the joint strength. ANOVA method is measured for both compression strength and hardness to calculate the percentage of contribution for each process parameter. Genetic algorithm is used to validate the results obtained from the both experimental value and optimization value. The micro structural study is depicted the welding joints characterization in between tube to tube plate joints. The radiograph test is conducted to prove the welds are non-defective and no flaws are found during the welding process. The mechanical property of compression strength and hardness has been measured to obtain the optimal joint strength of the welded sample was about 174.846 MPa and 131.364 Hv respectively. - Highlights: • Commercial Al tube and Al 2025 tube plate successfully welded by TIG welding. • Compression strength and hardness value proves to obtain optimal joint strength. • The maximum compression and hardness was achieved in various input parameters.

  19. An investigation on compression strength analysis of commercial aluminium tube to aluminium 2025 tube plate by using TIG welding process

    International Nuclear Information System (INIS)

    Kannan, S.; Senthil Kumaran, S.; Kumaraswamidhas, L.A.

    2016-01-01

    In this present study, Tungsten inert gas (TIG) welding was applied to weld the dissimilar materials and authenticate the mechanical and metallurgical properties of tube to tube plate made up of commercial aluminium and Al 2025 respectively using an Zirconiated tungsten electrode along with filler material aluminium ER 2219. In total, twenty five pieces has been subjected to compression strength and hardness value to evaluate the optimal joint strength. The three optimization technique has been used in this experiment. Taguchi L_2_5 orthogonal array is used to identify the most influencing process parameter which affects the joint strength. ANOVA method is measured for both compression strength and hardness to calculate the percentage of contribution for each process parameter. Genetic algorithm is used to validate the results obtained from the both experimental value and optimization value. The micro structural study is depicted the welding joints characterization in between tube to tube plate joints. The radiograph test is conducted to prove the welds are non-defective and no flaws are found during the welding process. The mechanical property of compression strength and hardness has been measured to obtain the optimal joint strength of the welded sample was about 174.846 MPa and 131.364 Hv respectively. - Highlights: • Commercial Al tube and Al 2025 tube plate successfully welded by TIG welding. • Compression strength and hardness value proves to obtain optimal joint strength. • The maximum compression and hardness was achieved in various input parameters.

  20. Regression analysis of pulsed eddy current signals for inspection of steam generator tube support structures

    International Nuclear Information System (INIS)

    Buck, J.; Underhill, P.R.; Mokros, S.G.; Morelli, J.; Krause, T.W.; Babbar, V.K.; Lepine, B.

    2015-01-01

    Nuclear steam generator (SG) support structure degradation and fouling can result in damage to SG tubes and loss of SG efficiency. Conventional eddy current technology is extensively used to detect cracks, frets at supports and other flaws, but has limited capabilities in the presence of multiple degradation modes or fouling. Pulsed eddy current (PEC) combined with principal components analysis (PCA) and multiple linear regression models was examined for the inspection of support structure degradation and SG tube off-centering with the goal of extending results to include additional degradation modes. (author)

  1. 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.

  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. Buckling analysis of micro- and nano-rods/tubes based on nonlocal Timoshenko beam theory

    International Nuclear Information System (INIS)

    Wang, C M; Zhang, Y Y; Ramesh, Sai Sudha; Kitipornchai, S

    2006-01-01

    This paper is concerned with the elastic buckling analysis of micro- and nano-rods/tubes based on Eringen's nonlocal elasticity theory and the Timoshenko beam theory. In the former theory, the small scale effect is taken into consideration while the effect of transverse shear deformation is accounted for in the latter theory. The governing equations and the boundary conditions are derived using the principle of virtual work. Explicit expressions for the critical buckling loads are derived for axially loaded rods/tubes with various end conditions. These expressions account for a better representation of the buckling behaviour of micro- and nano-rods/tubes where small scale effect and transverse shear deformation effect are significant. By comparing it with the classical beam theories, the sensitivity of the small scale effect on the buckling loads may be observed

  4. The application of ductile-fracture analysis to predictions of pressure-tube failure

    International Nuclear Information System (INIS)

    Simpson, L.A.

    1981-08-01

    Progress during the past six years towards establishing a method for predicting critical crack length in a reactor pressure tube, based on data from tests on small fracture-mechanics specimens, is reviewed. The disadvantages of relying on data from burst tests alone are described along with the benefits of a small-specimen method. It is clear from the work reviewed that only an approach that can account for the ability of the presssure tube material to increase its crack-growth resistance during stable crack extension is suitable for the prediction of critical crack length. A method that utilizes crack-growth resistance curves based on crack-opening displacement, or the J integral, is described, along with a large body of experimental data. It is concluded that the resistance curve approach provides a viable method for the analysis of fracture in pressure tubes that can greatly improve our understanding of the material's behaviour

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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.))

  10. The value of emergency CT studies in spontaneous rupture of hepatocellular carcinoma. Analysis for tumor protrusion and hemorrhagic ascites

    Energy Technology Data Exchange (ETDEWEB)

    Ishihara, Makiko; Kobayashi, Hisashi; Ichikawa, Taro; Cho, Keiichi; Gemma, Kazuhito; Kumazaki, Tatsuo [Nippon Medical School, Tokyo (Japan)

    1997-12-01

    CT characteristics of spontaneous rupture of HCC (n=13) were reviewed retrospectively, and the value of emergency CT studies in this disease was evaluated. Especially, tumor protrusion ratio (TPR) and ascitic CT numbers were measured to for comparison with the data for unruptured HCCs and ordinary, (e.g., non-hemorrhagic) ascites (n=13). As a result, except for diffuse type HCCs, the TPR was significantly higher than for the unruptured HCCs. Nine cases had intraperitoneal HDAs, and the laterality of the HDAs corresponded with that of the ruptured tumors in 8 cases. Also, the ascitic CT numbers apart from the HDA were still higher than the ordinary ascites. Therefore, a high TPR, HDAs adjacent to the tumor, and elevated ascitic CT numbers are important CT manifestations indicating HCC rupture. Diffuse HCCs, however, require careful clinical evaluation. (author)

  11. The value of emergency CT studies in spontaneous rupture of hepatocellular carcinoma. Analysis for tumor protrusion and hemorrhagic ascites

    International Nuclear Information System (INIS)

    Ishihara, Makiko; Kobayashi, Hisashi; Ichikawa, Taro; Cho, Keiichi; Gemma, Kazuhito; Kumazaki, Tatsuo

    1997-01-01

    CT characteristics of spontaneous rupture of HCC (n=13) were reviewed retrospectively, and the value of emergency CT studies in this disease was evaluated. Especially, tumor protrusion ratio (TPR) and ascitic CT numbers were measured to for comparison with the data for unruptured HCCs and ordinary, (e.g., non-hemorrhagic) ascites (n=13). As a result, except for diffuse type HCCs, the TPR was significantly higher than for the unruptured HCCs. Nine cases had intraperitoneal HDAs, and the laterality of the HDAs corresponded with that of the ruptured tumors in 8 cases. Also, the ascitic CT numbers apart from the HDA were still higher than the ordinary ascites. Therefore, a high TPR, HDAs adjacent to the tumor, and elevated ascitic CT numbers are important CT manifestations indicating HCC rupture. Diffuse HCCs, however, require careful clinical evaluation. (author)

  12. Reconstruction versus conservative treatment after rupture of the anterior cruciate ligament: cost effectiveness analysis

    Directory of Open Access Journals (Sweden)

    Farshad Mazda

    2011-11-01

    Full Text Available Abstract Background The decision whether to treat conservatively or reconstruct surgically a torn anterior cruciate ligament (ACL is an ongoing subject of debate. The high prevalence and associated public health burden of torn ACL has led to continuous efforts to determine the best therapeutic approach. A critical evaluation of benefits and expenditures of both treatment options as in a cost effectiveness analysis seems well-suited to provide valuable information for treating physicians and healthcare policymakers. Methods A literature review identified four of 7410 searched articles providing sufficient outcome probabilities for the two treatment options for modeling. A transformation key based on the expert opinions of 25 orthopedic surgeons was used to derive utilities from available evidence. The cost data for both treatment strategies were based on average figures compiled by Orthopaedic University Hospital Balgrist and reinforced by Swiss national statistics. A decision tree was constructed to derive the cost-effectiveness of each strategy, which was then tested for robustness using Monte Carlo simulation. Results Decision tree analysis revealed a cost effectiveness of 16,038 USD/0.78 QALY for ACL reconstruction and 15,466 USD/0.66 QALY for conservative treatment, implying an incremental cost effectiveness of 4,890 USD/QALY for ACL reconstruction. Sensitivity analysis of utilities did not change the trend. Conclusion ACL reconstruction for reestablishment of knee stability seems cost effective in the Swiss setting based on currently available evidence. This, however, should be reinforced with randomized controlled trials comparing the two treatment strategies.

  13. Spectral analysis in thin tubes with axial heterogeneities

    KAUST Repository

    Ferreira, Rita; Mascarenhas, M. Luí sa; Piatnitski, Andrey

    2015-01-01

    In this paper, we present the 3D-1D asymptotic analysis of the Dirichlet spectral problem associated with an elliptic operator with axial periodic heterogeneities. We extend to the 3D-1D case previous 3D-2D results (see [10]) and we analyze the special case where the scale of thickness is much smaller than the scale of the heterogeneities and the planar coefficient has a unique global minimum in the periodic cell. These results are of great relevance in the comprehension of the wave propagation in nanowires showing axial heterogeneities (see [17]).

  14. Experience of steam generator tube examination in the hot laboratory of EDF: analysis of recent events concerning the secondary side

    International Nuclear Information System (INIS)

    Thebault, Y.; Bouvier, O. de; Boccanfuso, M.; Coquio, N.; Barbe, V.; Molinie, E.

    2011-01-01

    Until 2010, more than 60 steam generator (SG) tubes have been removed and analysed in the EDF hot laboratory of CEIDRE/Chinon. This article is particularly related to three recent events that lead to the extraction of several tubes dedicated to laboratory destructive examinations. The first event that constitutes a first occurrence on the EDF Park, concerns the detection of a circumferential crack on the external surface of a tube located at tube support plate elevation. After this observation, several tubes have been extracted from Bugey 3 and Fessenheim 2 nuclear power plants with steam generators equipped with 600 MA bundle. The other two events concern the consequences of chemical cleaning of the tube bundle steam generators. The examples chosen are from Cruas 4 et Chinon B2 units whose tubes were extracted following non destructive testing performed immediately after or at the completion of cycle following the chemical cleaning. In the case of Cruas 4, Eddy Current Testing (ET) were performed for requalification of steam Generators after chemical cleaning. They allowed the detection of an indication located at the bottom of tube for a large number of tubes; the ET signal was similar to that corresponding to 'deposit' corrosion. Moreover, inspections of Chinon-B2 SGs at the end of the operation cycle following the chemical cleaning, showed the presence of conductor deposits at the bottom of some tubes. The first part of this document presents the major results of laboratory examinations of the pulled tubes of Bugey 3 and Fessenheim 2 and their analysis. Hypothesis concerning damage mechanisms of the tubes are also proposed. The second part of the paper relates the results of the laboratory examinations of the pulled tubes of Cruas 4 and Chinon B 2 after chemical cleaning and their analysis. (authors)

  15. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation

  16. 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

  17. Attitudes towards schizophrenia on YouTube: A content analysis of Finnish and Greek videos.

    Science.gov (United States)

    Athanasopoulou, Christina; Suni, Sanna; Hätönen, Heli; Apostolakis, Ioannis; Lionis, Christos; Välimäki, Maritta

    2016-01-01

    To investigate attitudes towards schizophrenia and people with schizophrenia presented in YouTube videos. We searched YouTube using the search terms "schizophrenia" and "psychosis" in Finnish and Greek language on April 3rd, 2013. The first 20 videos from each search (N = 80) were retrieved. Deductive content analysis was first applied for coding and data interpretation and it was followed by descriptive statistical analysis. A total of 52 videos were analyzed (65%). The majority of the videos were in the "Music" category (50%, n = 26). Most of the videos (83%, n = 43) tended to present schizophrenia in a negative way, while less than a fifth (17%, n = 9) presented schizophrenia in a positive or neutral way. Specifically, the most common negative attitude towards schizophrenia was dangerousness (29%, n = 15), while the most often identified positive attitude was objective, medically appropriate beliefs (21%, n = 11). All attitudes identified were similarly present in the Finnish and Greek videos, without any statistically significant difference. Negative presentations of schizophrenia are most likely to be accessed when searching YouTube for schizophrenia in Finnish and Greek language. More research is needed to investigate to what extent, if any, YouTube viewers' attitudes are affected by the videos they watch.

  18. 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.

  19. Statistical mechanical analysis of LMFBR fuel cladding tubes

    International Nuclear Information System (INIS)

    Poncelet, J.-P.; Pay, A.

    1977-01-01

    The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. (Auth.)

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. Accuracy of physical diagnostic tests for assessing ruptures of the anterior curciate ligament: a meta-analysis.

    NARCIS (Netherlands)

    Scholten, R.J.P.M.; Opstelten, W.; Plas, C.G. van der; Bijl, D.; Devillé, W.L.J.M.; Bouter, L.M.

    2003-01-01

    OBJECTIVE: This systematic review summarizes the evidence on the accuracy of tests for assessing ACL ruptures of the knee. SEARCH STRATEGY: A computerized search of MEDLINE (1966-2003) and EMBASE (1980-2003) with additional reference tracking. SELECTION CRITERIA: Articles included were written in

  7. Ultrasound elastography of the lower uterine segment in women with a previous cesarean section: Comparison of in-/ex-vivo elastography versus tensile-stress-strain-rupture analysis.

    Science.gov (United States)

    Seliger, Gregor; Chaoui, Katharina; Lautenschläger, Christine; Jenderka, Klaus-Vitold; Kunze, Christian; Hiller, Grit Gesine Ruth; Tchirikov, Michael

    2018-06-01

    The purpose of this study was to assess, if the biomechanical properties of the lower uterine segment (LUS) in women with a previous cesarean section (CS) can be determined by ultrasound (US) elastography. The first aim was to establish an ex-vivo LUS tensile-stress-strain-rupture(break point) analysis with the possibility of simultaneously using US elastography. The second aim was to investigate the relationship between measurement results of LUS stiffness using US elastography in-/ex-vivo with results of tensile-stress-strain-rupture analysis, and to compare different US elastography LUS-stiffness-measurement methods ex-vivo. An explorative experimental, in-/ex-vivo US study of women with previous CS was conducted. LUS elasticity was measured by point Shear Wave Elastography (pSWE) and bidimensional Shear-Wave-Elastography (2D-SWE) first in-vivo during preoperative examination within 24 h before repeat CS (including resection of the thinnest part of the LUS = uterine scar area during CS), second within 1 h after operation during the ex-vivo experiment, followed by tensile-stress-strain-rupture analysis. Pearson's correlation coefficient and scatter plots, Bland-Altman plots and paired T-tests, were used. Thirty three women were included in the study; elastography measurements n = 1412. The feasibility of ex-vivo assessment of LUS by quantitative US elastography using pSWE and 2D-SWE to detect stiffness of LUS was demonstrated. The strongest correlation with tensile-stress-strain analysis was found in the US elastography examination carried out with 2D-SWE (0.78, p break point - as a surrogate marker for the risk of rupture of the LUS after CS - is linearly dependent on the thickness of the LUS in the scar area (Coefficient of correlation: 0.79, p even at less stroke/strain than would be expected by their thickness. This study confirms that US elastography can help in determining viscoelastic properties of the LUS in women with a previous CS. The

  8. 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

  9. Knowledge networking on Sociology: network analysis of blogs, YouTube videos and tweets about Sociology

    Directory of Open Access Journals (Sweden)

    Julián Cárdenas

    2017-06-01

    Full Text Available While mainstream scientific knowledge production have been widely studied in recent years with the development of scientometrics and bibliometrics, an emergent number of studies have focused on alternative sources of production and dissemination of knowledge such as blogs, YouTube videos and comments on Twitter. These online sources of knowledge become relevant in fields such as Sociology, where some academics seek to bring the sociological knowledge to the general population. To explore which knowledge on Sociology is produced and disseminated, and how is organized in these online sources, we analyze the knowledge networking of blogs, YouTube videos and tweets on Twitter using network analysis approach. Specifically, the present research analyzes the hyperlink network of the main blogs on Sociology, the networks of tags used to classify videos on Sociology hosted on YouTube, and the network of hashtags linked to #sociología on Twitter. The main results point out the existence of a cohesive and strongly connected community of blogs on Sociology, the very low presence of YouTube videos on Sociology in Spanish, and Sociology on Twitter is linked to others social sciences, classical scholars and social media

  10. Analysis of microwave amplifier and frequency multiplier tube with a multipactor electron gun

    International Nuclear Information System (INIS)

    Yokoo, Kuniyoshi; Ono, Shoichi; Tai, Dong-Zhe.

    1983-01-01

    The performance analysis was made for a multipactor microwave tube with the aim of realizing a microwave amplifier or a frequency multiplier tube with a multipactor cathode with high efficiency and high power. The possibility for producing the multipactor tube with high efficiency and high power was shown by using effectively the characteristics of the multipactor cathode which emits pulsed electron current with narrow band, synchronizing with high frequency period. As the operating conditions for the multipactor cathode, it was shown that the wide spacing of the cathode was needed for the operation in high operating power, and the narrow spacing was needed for the operation in high efficiency and for reducing power consumption. It was also shown that there were the best values of the high-frequency voltage for the cathode operation. The study by the simulation for the multipactor cathode and for the acceleration zone of electron current was also performed to examine the possible performance for a microwave amplifier and a frequency multiplier tube. For the use of the multipactor cathode with a spacing of 1 mm, the conversion efficiency for d. c. input power was 86, 56 and 31 % for the primary, the secondary and the tertiary harmonic wave amplifications, respectively. (Asami, T.)

  11. 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)

  12. 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

  13. 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.

  14. A combined experimental and FE analysis procedure to evaluate tensile behavior of zircaloy pressure tubes

    International Nuclear Information System (INIS)

    Samal, M.K.; Vaze, K.K.; Balakrishnan, K.S.; Anantharaman, S.

    2012-01-01

    Determination of transverse mechanical properties from the ring type of specimens directly machined from the nuclear reactor pressure tubes is not straightforward because of the presence of combined membrane as well as bending stresses arising in the loaded condition. In this work, we have performed ring-tensile tests on the un-irradiated ring tensile specimen using two split semi-cylindrical mandrels as the loading device. A 3-D finite element (FE) analysis was performed in order to determine the material true stress-strain curve by comparing experimental load-displacement data with those predicted by FE analysis. In order to validate the methodology, miniaturized tensile specimens were machined from these tubes and tested. It was observed that the stress-strain data as obtained from ring tensile specimen could describe the load displacement curve of the miniaturized flat tensile specimen very well. (author)

  15. 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

  16. 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

  17. Management of chest tubes after pulmonary resection: a systematic review and meta-analysis.

    Science.gov (United States)

    Coughlin, Shaun M; Emmerton-Coughlin, Heather M A; Malthaner, Richard

    2012-08-01

    We performed a systematic review and meta-analysis to determine the effect of suction with water seal, compared with water seal alone, applied to intra pleural chest tubes on the duration of air leaks in patients undergoing pulmonary surgery. We searched MEDLINE, EMBASE and the Cochrane Central Register of Controlled Trials to find randomized controlled trials (RCTs) comparing the effect of the 2 methods on the duration of air leaks. Trials were systematically assessed for eligibility and validity. Data were extracted in duplicate and pooled across studies using a random-effects model. The search yielded 7 RCTs that met the eligibility criteria. No difference was identified between the 2 methods in duration of air leak (weighted mean difference [WMD] 1.15 days, favours water seal; 95% confidence interval [CI] -0.64 to 2.94), time to discharge (WMD 2.19 d, favours water seal; 95% CI -0.63 to 5.01), duration of chest tubes (WMD 0.96 d, favours water seal; 95% CI -0.12 to 2.05) or incidence of prolonged air leaks (absolute risk reduction [ARR] 0.04, favours water seal; 95% CI -0.01 to 0.09). Water seal was associated with a significantly increased incidence of postoperative pneumothorax (ARR -0.14, 95% CI -0.21 to -0.07). No differences were identified in terms of duration of air leak, incidence of prolonged air leak, duration of chest tubes and duration of hospital stay when chest tubes were placed to suction rather than water seal. Chest tube suction appears to be superior to water seal in reducing the incidence of pneumothorax; however, the clinical significance of this finding is unclear.

  18. 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.

  19. Ruptured Ectopic Pregnancy

    Directory of Open Access Journals (Sweden)

    Valentina Park

    2016-09-01

    Full Text Available History of present illness: A 21-year-old female presented with sudden onset suprapubic abdominal pain associated with dysuria. The patient also experienced near syncope during bowel movements three times three days ago without falling or losing consciousness. She denied fever, nausea, and vomiting. She stated that she was five weeks pregnant by last menstrual period. She had an ultrasound a few weeks before that showed no intrauterine pregnancy, but she had not followed up for additional testing. Significant findings: The patient’s serum beta-hCG was 5,637 mIU/mL. The transvaginal ultrasound showed an empty uterus with free fluid posteriorly in the pelvis and Pouch of Douglas (00:00. A 4.5 cm heterogeneous mass was visible in the left adnexa concerning for an ectopic pregnancy (00:10. Discussion: Ectopic pregnancies are a leading cause of maternal morbidity and mortality, as well as decreased fertility.1,2 Differentiating between an ectopic pregnancy and a normal early pregnancy may be difficult, since ultrasound and quantitative beta-hCG may show inconclusive results.3,4 Patients who have used fertility treatment may further complicate the picture because they are at risk for heterotypic pregnancies.5 Ectopic pregnancies most commonly implant in the fallopian tube, but may alternatively implant in the ovary, cervix, abdomen, or uterine cornua.4 Ultrasonography may show an empty uterus, adnexal mass, pelvic free fluid, or an extra-uterine gestational sac, yolk sac, and/or embryo.6 Treatment options for ectopic pregnancy include surgery or methotrexate.2,4 Some patients may be candidates for close outpatient surveillance if the diagnosis is unclear or in very limited cases for early, non-ruptured ectopic pregnancies.2,4

  20. The integral analysis of 40 mm diameter pipe rupture in cooling system of fusion facility W7-X with ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Kačegavičius, Tomas, E-mail: Tomas.Kacegavicius@lei.lt; Povilaitis, Mantas, E-mail: Mantas.Povilaitis@lei.lt

    2015-12-15

    Highlights: • The analysis of loss-of-coolant accident (LOCA) in W7-X facility. • Burst disc is sufficient to prevent pressure inside the plasma vessel exceeding 110 kPa. • Developed model of the cooling system adequately represents the expected phenomena. - Abstract: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental facility of stellarator type, which is currently being built at the Max-Planck-Institute for Plasmaphysics located in Greifswald, Germany. W7-X shall demonstrate that in future the energy could be produced in such type of fusion reactors. The safety analysis is required before the operation of the facility could be started. A rupture of 40 mm diameter pipe, which is connected to the divertor unit (module for plasma cooling) to ensure heat removal from the vacuum vessel in case of no-plasma operation mode “baking” is one of the design basis accidents to be investigated. During “baking” mode the vacuum vessel structures and working fluid – water are heated to the temperature 160 °C. This accident was selected for the detailed analysis using integral code ASTEC, which is developed by IRSN (France) and GRS mbH (Germany). This paper presents the integral analysis of W7-X response to a selected accident scenario. The model of the main cooling circuit and “baking” circuit was developed for ASTEC code. There were analysed two cases: (1) rupture of a pipe connected to the upper divertor unit and (2) rupture of a pipe connected to the lower divertor unit. The results of analysis showed that in both cases the water is almost completely released from the units into the plasma vessel. In both cases the pressure in the plasma vessel rapidly increases and in 28 s the set point for burst disc opening is reached preventing further pressurisation.

  1. 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

  2. Mid-term cost-effectiveness analysis of open and endovascular repair for ruptured abdominal aortic aneurysm.

    Science.gov (United States)

    Rollins, K E; Shak, J; Ambler, G K; Tang, T Y; Hayes, P D; Boyle, J R

    2014-02-01

    Emergency endovascular repair (EVAR) for ruptured abdominal aortic aneurysm (rAAA) may have lower operative mortality rates than open surgical repair. Concerns remain that the early survival benefit after EVAR for rAAA may be offset by late reinterventions. The aim of this study was to compare reintervention rates and cost-effectiveness of EVAR and open repair for rAAA. A retrospective analysis was undertaken of patients with rAAA undergoing EVAR or open repair over 6 years. A health economic model developed for the cost-effectiveness of elective EVAR was used in the emergency setting. Sixty-two patients (mean age 77·9 years) underwent EVAR and 85 (mean age 75·9 years) had open repair of rAAA. Median follow-up was 42 and 39 months respectively. There was no significant difference in 30-day mortality rates after EVAR and open repair (18 and 26 per cent respectively; P = 0·243). Reintervention rates were also similar (32 and 31 per cent; P = 0·701). The mean cost per patient was €26,725 for EVAR and €30,297 for open repair, and the cost per life-year gained was €7906 and €9933 respectively (P = 0·561). Open repair had greater initial costs: longer procedural times (217 versus 178·5 min; P < 0·001) and intensive care stay (5·0 versus 1·0 days; P = 0·015). Conversely, EVAR had greater reintervention (€156,939 versus €35,335; P = 0·001) and surveillance (P < 0·001) costs. There was no significant difference in reintervention rates after EVAR or open repair for rAAA. EVAR was as cost-effective at mid-term follow-up. The increased procedural costs of open repair are not outweighed by greater surveillance and reintervention costs after EVAR. © 2014 BJS Society Ltd. Published by John Wiley & Sons Ltd.

  3. 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.

  4. 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

  5. Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

    International Nuclear Information System (INIS)

    Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

    1983-01-01

    A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

  6. Numerical Analysis of Turbulent Flow around Tube Bundle by Applying CAD Best Practice Guideline

    International Nuclear Information System (INIS)

    Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong; Cheng, Ae Ju

    2013-01-01

    In this study, the numerical analysis of a turbulent flow around both a staggered and an incline tube bundle was conducted using Annoys Cfx V. 13, a commercial CAD software. The flow was assumed to be steady, incompressible, and isothermal. According to the CAD Best Practice Guideline, the sensitivity study for grid size, accuracy of the discretization scheme for convection term, and turbulence model was conducted, and its result was compared with the experimental data to estimate the applicability of the CAD Best Practice Guideline. It was concluded that the CAD Best Practice Guideline did not always guarantee an improvement in the prediction performance of the commercial CAD software in the field of tube bundle flow

  7. Parametric study and performance analysis of hybrid rocket motors with double-tube configuration

    Science.gov (United States)

    Yu, Nanjia; Zhao, Bo; Lorente, Arnau Pons; Wang, Jue

    2017-03-01

    The practical implementation of hybrid rocket motors has historically been hampered by the slow regression rate of the solid fuel. In recent years, the research on advanced injector designs has achieved notable results in the enhancement of the regression rate and combustion efficiency of hybrid rockets. Following this path, this work studies a new configuration called double-tube characterized by injecting the gaseous oxidizer through a head end injector and an inner tube with injector holes distributed along the motor longitudinal axis. This design has demonstrated a significant potential for improving the performance of hybrid rockets by means of a better mixing of the species achieved through a customized injection of the oxidizer. Indeed, the CFD analysis of the double-tube configuration has revealed that this design may increase the regression rate over 50% with respect to the same motor with a conventional axial showerhead injector. However, in order to fully exploit the advantages of the double-tube concept, it is necessary to acquire a deeper understanding of the influence of the different design parameters in the overall performance. In this way, a parametric study is carried out taking into account the variation of the oxidizer mass flux rate, the ratio of oxidizer mass flow rate injected through the inner tube to the total oxidizer mass flow rate, and injection angle. The data for the analysis have been gathered from a large series of three-dimensional numerical simulations that considered the changes in the design parameters. The propellant combination adopted consists of gaseous oxygen as oxidizer and high-density polyethylene as solid fuel. Furthermore, the numerical model comprises Navier-Stokes equations, k-ε turbulence model, eddy-dissipation combustion model and solid-fuel pyrolysis, which is computed through user-defined functions. This numerical model was previously validated by analyzing the computational and experimental results obtained for

  8. 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

  9. 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.

  10. Direct ophthalmoscopy on YouTube: analysis of instructional YouTube videos' content and approach to visualization.

    Science.gov (United States)

    Borgersen, Nanna Jo; Henriksen, Mikael Johannes Vuokko; Konge, Lars; Sørensen, Torben Lykke; Thomsen, Ann Sofia Skou; Subhi, Yousif

    2016-01-01

    Direct ophthalmoscopy is well-suited for video-based instruction, particularly if the videos enable the student to see what the examiner sees when performing direct ophthalmoscopy. We evaluated the pedagogical effectiveness of instructional YouTube videos on direct ophthalmoscopy by evaluating their content and approach to visualization. In order to synthesize main themes and points for direct ophthalmoscopy, we formed a broad panel consisting of a medical student, junior and senior physicians, and took into consideration book chapters targeting medical students and physicians in general. We then systematically searched YouTube. Two authors reviewed eligible videos to assess eligibility and extract data on video statistics, content, and approach to visualization. Correlations between video statistics and contents were investigated using two-tailed Spearman's correlation. We screened 7,640 videos, of which 27 were found eligible for this study. Overall, a median of 12 out of 18 points (interquartile range: 8-14 key points) were covered; no videos covered all of the 18 points assessed. We found the most difficulties in the approach to visualization of how to approach the patient and how to examine the fundus. Time spent on fundus examination correlated with the number of views per week (Spearman's ρ=0.53; P=0.029). Videos may help overcome the pedagogical issues in teaching direct ophthalmoscopy; however, the few available videos on YouTube fail to address this particular issue adequately. There is a need for high-quality videos that include relevant points, provide realistic visualization of the examiner's view, and give particular emphasis on fundus examination.

  11. Modification of hydrogen determinator for total hydrogen analysis in irradiated zircaloy cladding tube

    International Nuclear Information System (INIS)

    Park, Soon Dal; Choi, Kwnag Soon; Kim, Jong Goo; Joe, Kih Soo; Kim, Won Ho

    1999-01-01

    A hydrogen determinator was modified and installed in the glove box to analyse total hydrogen content in irradiated zircaloy tube. The analysis method of hydrogen is Inert Gas Fusion(IGF)-Thermal Conductivity Detection(TCD). The hydrogen recoveries of no tin method using Ti and Zr matrix standards, respectively, were available within 3 μg of hydrogen. Also the smaller size of sample showed the better hydrogen recovery. It was found that the hydrogen standard of Ti matrix is available to hydrogen analysis in zircaloy sample. The mean radioactivity of irradiated zircaloy sample was 10 mR/hr and hydrogen concentration was 130 ppm

  12. Statistical analysis of failure time in stress corrosion cracking of fuel tube in light water reactor

    International Nuclear Information System (INIS)

    Hirao, Keiichi; Yamane, Toshimi; Minamino, Yoritoshi

    1991-01-01

    This report is to show how the life due to stress corrosion cracking breakdown of fuel cladding tubes is evaluated by applying the statistical techniques to that examined by a few testing methods. The statistical distribution of the limiting values of constant load stress corrosion cracking life, the statistical analysis by making the probabilistic interpretation of constant load stress corrosion cracking life, and the statistical analysis of stress corrosion cracking life by the slow strain rate test (SSRT) method are described. (K.I.)

  13. 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)

  14. A Therapist’s Review of Process: Rupture and repair cycles in relational transactional analysis psychotherapy for a client with a dismissive attachment style: ‘Martha’

    Directory of Open Access Journals (Sweden)

    Silvia Baba Neal

    2017-07-01

    Full Text Available This article is a therapist review of the process that occurred during a systematic case study of psychotherapy with ‘Martha’, a female client who presented with depression, anxiety, alexithymia and dismissive/avoidant attachment style.  Assessment, diagnosis of the client and treatment direction is described, followed by a detailed account of the therapeutic process through 12 sessions and 2 post-therapy interviews. Analysis team results are summarised, indicating support for the therapist’s identification of issues during the process of the therapy. Particular attention is paid by the analysis team two points of rupture and repair, with pragmatic evaluation confirming that the relational struggles between therapist and client seemed pivotal in generating positive change. Citation - APA format: Baba Neal, S. (2017. A Therapist’s Review of Process: Rupture and repair cycles in relational transactional analysis psychotherapy for a client with a dismissive attachment style: ‘Martha’. International Journal of Transactional Analysis Research & Practice, 8(2, 24-34.

  15. 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.

  16. Local chemical and thermal-hydraulic analysis of U-tube steam generators

    International Nuclear Information System (INIS)

    Lee, J.Y.; No, H.C.

    1990-01-01

    In order to know how pH distribution affects corrosion in a U-tube steam generator, a study of the combination of water chemistry and thermal-hydraulic conditions is suggested. A two-fluid (unequal velocity and unequal temperature) formulation is proposed to describe the convective transport of volatile species in each phase, and a spherical bubble model is developed on the basis of the penetration theory to describe the interfacial mass transfer. The thermal-hydraulic local conditions are obtained by the U-tube steam generator design analysis code FAUST which is based on the three-dimensional two-fluid model. The results of the present study are compared with dynamic equilibrium model calculations. This study shows that, in contrast with dynamic equilibrium calculations, the pH is lower in the cold-leg side than in the hot-leg side because of liquid recirculation. Just above the tube sheet, however, the lower void fraction in this region than that in the hot-leg region results in higher pH, which agrees with the prediction of the dynamic equilibrium model. (orig.)

  17. Engineering analysis of ITER In-Vessel Viewing System guide tube

    Energy Technology Data Exchange (ETDEWEB)

    Casal, Natalia, E-mail: natalia.casal@iter.org [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Bates, Philip [Fusion for Energy, Barcelona (Spain); Bede, Ottó [Oxford Technologies Ltd., Abingdon (United Kingdom); Damiani, Carlo; Dubus, Gregory [Fusion for Energy, Barcelona (Spain); Omran, Hassan [Oxford Technologies Ltd., Abingdon (United Kingdom); Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France); Puiu, Adrian [Fusion for Energy, Barcelona (Spain); Reichle, Roger; Suárez, Alejandro; Walker, Christopher; Walsh, Michael [ITER Organization, Route de Vinon sur Verdon, St Paul-lez-Durance (France)

    2015-10-15

    Highlights: • Conceptual design of IVVS Loads action on IVVS Dominant loads. • Seismic and baking conditions. • No active cooling needed for IVVS. • IVVS requires additional support points to avoid excessive deformation. - Abstract: The In Vessel Viewing System (IVVS) will be one of the essential machine diagnostic systems at ITER to provide information about the status of in-vessel and plasma facing components and to evaluate the dust inside the Vacuum Vessel. The current design consists of six scanning probes and their deployment systems, which are placed in dedicated ports at the divertor level. These units are located in resident guiding tubes 10 m long, which allow the IVVS probes to go from their storage location to the scanning position by means of a simple straight translation. Moreover, each resident tube is supported inside the corresponding Vacuum Vessel and Cryostat port extensions, which are part of the primary confinement barrier. As the Vacuum Vessel and the Cryostat will move with respect to each other during operation (especially during baking) and during incidents and accidents (disruptions, vertical displacement events, seismic events), the structural integrity of the resident tube and the surrounding vacuum boundaries would be compromised if the required flexibility and supports are not appropriately assured. This paper focuses on the integration of the present design of the IVVS into the Vacuum Vessel and Cryostat environment. It presents the adopted strategy to withstand all the main interfacing loads without damaging the confinement barriers and the corresponding analysis supporting it.

  18. YouTube as a source of information on skin bleaching: a content analysis.

    Science.gov (United States)

    Basch, C H; Brown, A A; Fullwood, M D; Clark, A; Fung, I C-H; Yin, J

    2018-06-01

    Skin bleaching is a common, yet potentially harmful body modification practice. To describe the characteristics of the most widely viewed YouTube™ videos related to skin bleaching. The search term 'skin bleaching' was used to identify the 100 most popular English-language YouTube videos relating to the topic. Both descriptive and specific information were noted. Among the 100 manually coded skin-bleaching YouTube videos in English, there were 21 consumer-created videos, 45 internet-based news videos, 30 television news videos and 4 professional videos. Excluding the 4 professional videos, we limited our content categorization and regression analysis to 96 videos. Approximately 93% (89/96) of the most widely viewed videos mentioned changing how you look and 74% (71/96) focused on bleaching the whole body. Of the 96 videos, 63 (66%) of videos showed/mentioned a transformation. Only about 14% (13/96) mentioned that skin bleaching is unsafe. The likelihood of a video selling a skin bleaching product was 17 times higher in internet videos compared with consumer videos (OR = 17.00, 95% CI 4.58-63.09, P YouTube video on skin bleaching was uploaded by an internet source. Videos made by television sources mentioned more information about skin bleaching being unsafe, while consumer-generated videos focused more on making skin-bleaching products at home. © 2017 British Association of Dermatologists.

  19. FAILURE ANALYSIS IN TUBING OF AIR PREHEATER OF BOILER FROM A SUGARCANE MILL

    Directory of Open Access Journals (Sweden)

    Joner Oliveira Alves

    2014-10-01

    Full Text Available The increased demand for energy from sugarcane bagasse has made the sugar and alcohol mills search alternatives to reduce maintenance of the boilers, releasing more time to the production. The stainless steel use has become one of the main tools for such reduction. However, specification errors can lead to premature failures. This work reports the factors that led tubes of AISI 409 stainless steel fail after half season when applied in a air preheater of boiler from a sugarcane mill. In such application, the AISI 304 lasts about 15 seasons and the carbon steel about 3. A tube sent by the sugar mill was characterized by wet chemical analysis, optical microscopy and EDS. Results indicated chloride formation on the internal walls of the tube, which combined with the environment, accelerated the corrosion process. The carbon steel showed high lifetime due to a 70% higher thickness. Due to the work condictions is recommended the use of stainless steels with higher corrosion resistance, such as the traditional AISI 304 or the ferritic AISI 444, the last presents better thermal exchange.

  20. [Is the use of plastic capillary tubes justified for blood gases analysis?].

    Science.gov (United States)

    Daurès, Marie-Françoise; Bozonnat, Marie-Cécile; Cristol, Jean-Paul

    2011-01-01

    Some clinical units, such as neonatal or maternity units, preferentially use capillary tubes when analysing blood gases. Using glass tubes is delicate and nurses must recollect blood when breaking. In order to eliminate this problem, we tested flexible, plastic capillary tubes in both the above mentionned units and in our biochemistry laboratory. Each unit, where glass tubes were habitually used, tested 200 flexible, plastic capillary tubes. In addition, the nursing staffed filled out a questionnaire concerned tube usage. Both units clearly preferred using the flexible tubes. In the laboratory, results for blood gas analyses were compared between rigid glass and flexible plastic capillary tubes for 112 patients. Concordance tests did not showed significant differences between the two tube types, except for hematocrit and total haemoglobin. A questionnaire was also presented to the lab technician, who confirmed the easier usability of plastic capillary tubes.

  1. ePatients on YouTube: Analysis of Four Experiences From the Patients' Perspective

    Science.gov (United States)

    Gómez-Zúñiga, Beni; Pousada, Modesta; Hernández-Encuentra, Eulàlia; Armayones, Manuel

    2012-01-01

    Background Many patients share their personal experiences and opinions using online video platforms. These videos are watched by millions of health consumers and health care professionals. Although it has become a popular phenomenon, little is known about patients who share videos online and why they do so. Objective We aimed to explore the motivations and challenges faced by patients who share videos about their health and experiences on YouTube. As part of a conference discussion, we asked several patients actively engaged on YouTube to make a video explaining their motivations. This paper discusses these videos. Methods In this qualitative study, we performed an analysis of the videos created by 4 patients about their self-reported motivations and challenges they face as YouTube users. First, two judges compared the transcriptions and decided the exact wording when confusing content was found. Second, two judges categorized the content of the videos to identify the major themes. Results Four main categories emerged: (1) the origin or cause for making the first video, (2) the objectives that they achieve by continuing to make videos, (3) the perception of community, and (4) the negative consequences of the experience. Conclusions The main reason for making videos was to bridge the gap between traditional health information about their diseases and everyday life. The first consequence of sharing their life on YouTube was a loss of privacy. However, they also experienced the positive effects of expressing their feelings, being part of a large community of peers, and helping others to deal with a chronic condition. PMID:25075229

  2. YouTube as a source of COPD patient education: A social media content analysis

    Science.gov (United States)

    Stellefson, Michael; Chaney, Beth; Ochipa, Kathleen; Chaney, Don; Haider, Zeerak; Hanik, Bruce; Chavarria, Enmanuel; Bernhardt, Jay M.

    2014-01-01

    Objective Conduct a social media content analysis of COPD patient education videos on YouTube. Methods A systematic search protocol was used to locate 223 videos. Two independent coders evaluated each video to determine topics covered, media source(s) of posted videos, information quality as measured by HONcode guidelines for posting trustworthy health information on the Internet, and viewer exposure/engagement metrics. Results Over half the videos (n=113, 50.7%) included information on medication management, with far fewer videos on smoking cessation (n=40, 17.9%). Most videos were posted by a health agency or organization (n=128, 57.4%), and the majority of videos were rated as high quality (n=154, 69.1%). HONcode adherence differed by media source (Fisher’s Exact Test=20.52, p=.01), with user-generated content (UGC) receiving the lowest quality scores. Overall level of user engagement as measured by number of “likes,” “favorites,” “dislikes,” and user comments was low (mdn range = 0–3, interquartile (IQR) range = 0–16) across all sources of media. Conclusion Study findings suggest that COPD education via YouTube has the potential to reach and inform patients, however, existing video content and quality varies significantly. Future interventions should help direct individuals with COPD to increase their engagement with high-quality patient education videos on YouTube that are posted by reputable health organizations and qualified medical professionals. Patients should be educated to avoid and/or critically view low-quality videos posted by individual YouTube users who are not health professionals. PMID:24659212

  3. Discrimination of DPRK M5.1 February 12th, 2013 Earthquake as Nuclear Test Using Analysis of Magnitude, Rupture Duration and Ratio of Seismic Energy and Moment

    Science.gov (United States)

    Salomo Sianipar, Dimas; Subakti, Hendri; Pribadi, Sugeng

    2015-04-01

    On February 12th, 2013 morning at 02:57 UTC, there had been an earthquake with its epicenter in the region of North Korea precisely around Sungjibaegam Mountains. Monitoring stations of the Preparatory Commission for the Comprehensive Nuclear Test-Ban Treaty Organization (CTBTO) and some other seismic network detected this shallow seismic event. Analyzing seismograms recorded after this event can discriminate between a natural earthquake or an explosion. Zhao et. al. (2014) have been successfully discriminate this seismic event of North Korea nuclear test 2013 from ordinary earthquakes based on network P/S spectral ratios using broadband regional seismic data recorded in China, South Korea and Japan. The P/S-type spectral ratios were powerful discriminants to separate explosions from earthquake (Zhao et. al., 2014). Pribadi et. al. (2014) have characterized 27 earthquake-generated tsunamis (tsunamigenic earthquake or tsunami earthquake) from 1991 to 2012 in Indonesia using W-phase inversion analysis, the ratio between the seismic energy (E) and the seismic moment (Mo), the moment magnitude (Mw), the rupture duration (To) and the distance of the hypocenter to the trench. Some of this method was also used by us to characterize the nuclear test earthquake. We discriminate this DPRK M5.1 February 12th, 2013 earthquake from a natural earthquake using analysis magnitude mb, ms and mw, ratio of seismic energy and moment and rupture duration. We used the waveform data of the seismicity on the scope region in radius 5 degrees from the DPRK M5.1 February 12th, 2013 epicenter 41.29, 129.07 (Zhang and Wen, 2013) from 2006 to 2014 with magnitude M ≥ 4.0. We conclude that this earthquake was a shallow seismic event with explosion characteristics and can be discriminate from a natural or tectonic earthquake. Keywords: North Korean nuclear test, magnitude mb, ms, mw, ratio between seismic energy and moment, ruptures duration

  4. Status of selected ion flow tube MS: accomplishments and challenges in breath analysis and other areas.

    Science.gov (United States)

    Smith, David; Španěl, Patrik

    2016-06-01

    This article reflects our observations of recent accomplishments made using selected ion flow tube MS (SIFT-MS). Only brief descriptions are given of SIFT-MS as an analytical method and of the recent extensions to the underpinning analytical ion chemistry required to realize more robust analyses. The challenge of breath analysis is given special attention because, when achieved, it renders analysis of other air media relatively straightforward. Brief overviews are given of recent SIFT-MS breath analyses by leading research groups, noting the desirability of detection and quantification of single volatile biomarkers rather than reliance on statistical analyses, if breath analysis is to be accepted into clinical practice. A 'strengths, weaknesses, opportunities and threats' analysis of SIFT-MS is made, which should help to increase its utility for trace gas analysis.

  5. Iterative reconstruction for quantitative computed tomography analysis of emphysema: consistent results using different tube currents

    Directory of Open Access Journals (Sweden)

    Yamashiro T

    2015-02-01

    Full Text Available Tsuneo Yamashiro,1 Tetsuhiro Miyara,1 Osamu Honda,2 Noriyuki Tomiyama,2 Yoshiharu Ohno,3 Satoshi Noma,4 Sadayuki Murayama1 On behalf of the ACTIve Study Group 1Department of Radiology, Graduate School of Medical Science, University of the Ryukyus, Nishihara, Okinawa, Japan; 2Department of Radiology, Osaka University Graduate School of Medicine, Suita, Osaka, Japan; 3Department of Radiology, Kobe University Graduate School of Medicine, Kobe, Hyogo, Japan; 4Department of Radiology, Tenri Hospital, Tenri, Nara, Japan Purpose: To assess the advantages of iterative reconstruction for quantitative computed tomography (CT analysis of pulmonary emphysema. Materials and methods: Twenty-two patients with pulmonary emphysema underwent chest CT imaging using identical scanners with three different tube currents: 240, 120, and 60 mA. Scan data were converted to CT images using Adaptive Iterative Dose Reduction using Three Dimensional Processing (AIDR3D and a conventional filtered-back projection mode. Thus, six scans with and without AIDR3D were generated per patient. All other scanning and reconstruction settings were fixed. The percent low attenuation area (LAA%; < -950 Hounsfield units and the lung density 15th percentile were automatically measured using a commercial workstation. Comparisons of LAA% and 15th percentile results between scans with and without using AIDR3D were made by Wilcoxon signed-rank tests. Associations between body weight and measurement errors among these scans were evaluated by Spearman rank correlation analysis. Results: Overall, scan series without AIDR3D had higher LAA% and lower 15th percentile values than those with AIDR3D at each tube current (P<0.0001. For scan series without AIDR3D, lower tube currents resulted in higher LAA% values and lower 15th percentiles. The extent of emphysema was significantly different between each pair among scans when not using AIDR3D (LAA%, P<0.0001; 15th percentile, P<0.01, but was not

  6. Head losses prediction and analysis in a bulb turbine draft tube under different operating conditions using unsteady simulations

    Science.gov (United States)

    Wilhelm, S.; Balarac, G.; Métais, O.; Ségoufin, C.

    2016-11-01

    Flow prediction in a bulb turbine draft tube is conducted for two operating points using Unsteady RANS (URANS) simulations and Large Eddy Simulations (LES). The inlet boundary condition of the draft tube calculation is a rotating two dimensional velocity profile exported from a RANS guide vane- runner calculation. Numerical results are compared with experimental data in order to validate the flow field and head losses prediction. Velocity profiles prediction is improved with LES in the center of the draft tube compared to URANS results. Moreover, more complex flow structures are obtained with LES. A local analysis of the predicted flow field using the energy balance in the draft tube is then introduced in order to detect the hydrodynamic instabilities responsible for head losses in the draft tube. In particular, the production of turbulent kinetic energy next to the draft tube wall and in the central vortex structure is found to be responsible for a large part of the mean kinetic energy dissipation in the draft tube and thus for head losses. This analysis is used in order to understand the differences in head losses for different operating points. The numerical methodology could then be improved thanks to an in-depth understanding of the local flow topology.

  7. Numerical analysis on the condensation heat transfer and pressure drop characteristics of the horizontal tubes of modular shell and tube-bundle heat exchanger

    International Nuclear Information System (INIS)

    Ko, Seung Hwan; Park, Hyung Gyu; Kim, Charn Jung; Park, Byung Kyu

    2001-01-01

    A numerical analysis of the heat and mass transfer and pressure drop characteristics in modular shell and tube bundle heat exchanger was carried out. Finite concept method based on FVM and κ-ε turbulent model were used for this analysis. Condensation heat transfer enhanced total heat transfer rate 4∼8% higher than that of dry heat exchanger. With increasing humid air inlet velocity, temperature and relative humidity, and with decreasing heat exchanger aspect ratio and cooling water velocity, total heat and mass transfer rate could be increased. Cooling water inlet velocity had little effect on total heat transfer

  8. Test beam analysis of the first CMS drift tube muon chamber

    CERN Document Server

    Albajar, C; Arce, P; Autermann, C; Bellato, M; Benettoni, M; Benvenuti, Alberto C; Bontenackels, M; Caballero, J; Cavallo, F R; Cerrada, M; Cirio, R; Colino, N; Conti, E; de la Cruz, B; Dal Corso, F; Dallavalle, G M; Fernández, C; Fernández de Troconiz, J; Fouz-Iglesias, M C; García-Abia, P; García-Raboso, A; Gasparini, F; Gasparini, U; Giacomelli, P; Gonella, F; Gulmini, M; Hebbeker, T; Hermann, S; Höpfner, K; Jiménez, I; Josa-Mutuberria, I; Lacaprara, S; Marcellini, S; Mariotti, C; Maron, G; Maselli, S; Meneguzzo, Anna Teresa; Monaco, V; Montanari, A; Montanari, C; Montecassiano, F; Navarria, Francesco Luigi; Odorici, F; Passaseo, M; Pegoraro, M; Peroni, C; Perrotta, A; Puerta, J; Reithler, H; Romero, A; Romero, L; Ronchese, P; Rossi, A; Rovelli, T; Sacchi, R; Sowa, M; Staiano, A; Toniolo, N; Torassa, E; Vaniev, V; Vanini, S; Ventura, Sandro; Villanueva, C; Willmott, C; Zotto, P L; Zumerle, G

    2004-01-01

    In October 2001 the first produced CMS Barrel Drift Tube (DT) Muon Chamber was tested at the CERN Gamma Irradiation Facility (GIF) using a muon beam. A Resistive Plate Chamber (RPC) was attached to the top of the DT chamber, and, for the first time, both detectors were operated coupled together. The performance of the DT chamber was studied for several operating conditions, and for gamma rates similar to the ones expected at LHC. In this paper we present the data analysis; the results are considered fully satisfactory.

  9. Test beam analysis of the first CMS drift tube muon chamber

    International Nuclear Information System (INIS)

    Albajar, C.; Amapane, N.; Arce, P.; Autermann, C.; Bellato, M.; Benettoni, M.; Benvenuti, A.; Bontenackels, M.; Caballero, J.; Cavallo, F.R.; Cerrada, M.; Cirio, R.; Colino, N.; Conti, E.; Cruz, B. de la; Corso, F. dal; Dallavalle, G.M.; Fernandez, C.; Troconiz, J.F. de; Fouz, M.C.; Garcia-Abia, P.; Garcia-Raboso, A.; Gasparini, F.; Gasparini, U.; Giacomelli, P.; Gonella, F.; Gulmini, M.; Hebbeker, T.; Hermann, S.; Hoepfner, K.; Jimenez, I.; Josa, I.; Lacaprara, S.; Marcellini, S.; Mariotti, C.; Maron, G.; Maselli, S.; Meneguzzo, A.T.; Monaco, V.; Montanari, A.; Montanari, C.; Montecassiano, F.; Navarria, F.L.; Odorici, F.; Passaseo, M.; Pegoraro, M.; Peroni, C.; Perrotta, A.; Puerta, J.; Reithler, H.; Romero, A.; Romero, L.; Ronchese, P.; Rossi, A.; Rovelli, T.; Sacchi, R.; Sowa, M.; Staiano, A.; Toniolo, N.; Torassa, E.; Vaniev, V.; Vanini, S.; Ventura, S.; Villanueva, C.; Willmott, C.; Zotto, P.; Zumerle, G.

    2004-01-01

    In October 2001 the first produced CMS Barrel Drift Tube (DT) Muon Chamber was tested at the CERN Gamma Irradiation Facility (GIF) using a muon beam. A Resistive Plate Chamber (RPC) was attached to the top of the DT chamber, and, for the first time, both detectors were operated coupled together. The performance of the DT chamber was studied for several operating conditions, and for gamma rates similar to the ones expected at LHC. In this paper we present the data analysis; the results are considered fully satisfactory

  10. The YouTube Jihadists: A Social Network Analysis of Al-Muhajiroun’s Propaganda Campaign

    Directory of Open Access Journals (Sweden)

    Jytte Klausen

    2012-03-01

    Full Text Available Producers of Al-Qaeda inspired propaganda have shifted their operations in recent years from closed membership online forums to mainstream social networking platforms. Using social network analysis, we show that behind the apparent proliferation of such sources, YouTube account holders associated with incarnations of the British al-Muhajiroun collude to post propaganda and violent content. European groups commonly use American platforms and domain names registered with American companies. Seeking shelter under speech rights granted by the First Amendment, they evade European laws against incitement and hate speech.

  11. Test beam analysis of the first CMS drift tube muon chamber

    Energy Technology Data Exchange (ETDEWEB)

    Albajar, C.; Amapane, N.; Arce, P.; Autermann, C.; Bellato, M.; Benettoni, M.; Benvenuti, A.; Bontenackels, M.; Caballero, J.; Cavallo, F.R.; Cerrada, M.; Cirio, R.; Colino, N.; Conti, E.; Cruz, B. de la; Corso, F. dal; Dallavalle, G.M.; Fernandez, C.; Troconiz, J.F. de E-mail: jorge.troconiz@uam.es; Fouz, M.C.; Garcia-Abia, P.; Garcia-Raboso, A.; Gasparini, F.; Gasparini, U.; Giacomelli, P.; Gonella, F.; Gulmini, M.; Hebbeker, T.; Hermann, S.; Hoepfner, K.; Jimenez, I.; Josa, I.; Lacaprara, S.; Marcellini, S.; Mariotti, C.; Maron, G.; Maselli, S.; Meneguzzo, A.T.; Monaco, V.; Montanari, A.; Montanari, C.; Montecassiano, F.; Navarria, F.L.; Odorici, F.; Passaseo, M.; Pegoraro, M.; Peroni, C.; Perrotta, A.; Puerta, J.; Reithler, H.; Romero, A.; Romero, L.; Ronchese, P.; Rossi, A.; Rovelli, T.; Sacchi, R.; Sowa, M.; Staiano, A.; Toniolo, N.; Torassa, E.; Vaniev, V.; Vanini, S.; Ventura, S.; Villanueva, C.; Willmott, C.; Zotto, P.; Zumerle, G

    2004-06-11

    In October 2001 the first produced CMS Barrel Drift Tube (DT) Muon Chamber was tested at the CERN Gamma Irradiation Facility (GIF) using a muon beam. A Resistive Plate Chamber (RPC) was attached to the top of the DT chamber, and, for the first time, both detectors were operated coupled together. The performance of the DT chamber was studied for several operating conditions, and for gamma rates similar to the ones expected at LHC. In this paper we present the data analysis; the results are considered fully satisfactory.

  12. A content analysis of smoking fetish videos on YouTube: regulatory implications for tobacco control.

    Science.gov (United States)

    Kim, Kyongseok; Paek, Hye-Jin; Lynn, Jordan

    2010-03-01

    This study examined the prevalence, accessibility, and characteristics of eroticized smoking portrayal, also referred to as smoking fetish, on YouTube. The analysis of 200 smoking fetish videos revealed that the smoking fetish videos are prevalent and accessible to adolescents on the website. They featured explicit smoking behavior by sexy, young, and healthy females, with the content corresponding to PG-13 and R movie ratings. We discuss a potential impact of the prosmoking image on youth according to social cognitive theory, and implications for tobacco control.

  13. Storing of Extracts in Polypropylene Microcentrifuge Tubes Yields Contaminant Peak During Ultra-flow Liquid Chromatographic Analysis

    OpenAIRE

    Kshirsagar, Parthraj R.; Hegde, Harsha; Pai, Sandeep R.

    2016-01-01

    Background and Aim: This study was designed to understand the effect of storage in polypropylene microcentrifuge tubes and glass vials during ultra-flow liquid chromatographic (UFLC) analysis. Materials and Methods: One ml of methanol was placed in polypropylene microcentrifuge tubes (PP material, Autoclavable) and glass vials (Borosilicate) separately for 1, 2, 4, 8, 10, 20, 40, and 80 days intervals stored at ?4?C. Results: Contaminant peak was detected in methanol stored in polypropylene m...

  14. Premature rupture of membranes

    Science.gov (United States)

    ... gov/ency/patientinstructions/000512.htm Premature rupture of membranes To use the sharing features on this page, ... water that surrounds your baby in the womb. Membranes or layers of tissue hold in this fluid. ...

  15. Ruptured submitral aneurysm

    Directory of Open Access Journals (Sweden)

    V. Shukla

    2016-09-01

    Full Text Available Submitral aneurysm is a rare entity, with around few hundred cases reported till date. Presentation can be varied. We describe here a case of submitral aneurysm in a young male with rupture into the left atrium cavity.

  16. Achilles Tendon Rupture

    Science.gov (United States)

    ... is a strong fibrous cord that connects the muscles in the back of your calf to your heel bone. If you overstretch your Achilles tendon, it can tear (rupture) completely or just partially. If your Achilles ...

  17. 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)

  18. 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.

  19. A Local Condensation Analysis Representing Two-phase Annular Flow in Condenser/radiator Capillary Tubes

    Science.gov (United States)

    Karimi, Amir

    1991-01-01

    NASA's effort for the thermal environmental control of the Space Station Freedom is directed towards the design, analysis, and development of an Active Thermal Control System (ATCS). A two phase, flow through condenser/radiator concept was baselined, as a part of the ATCS, for the radiation of space station thermal load into space. The proposed condenser rejects heat through direct condensation of ATCS working fluid (ammonia) in the small diameter radiator tubes. Analysis of the condensation process and design of condenser tubes are based on the available two phase flow models for the prediction of flow regimes, heat transfer, and pressure drops. The prediction formulas use the existing empirical relationships of friction factor at gas-liquid interface. An attempt is made to study the stability of interfacial waves in two phase annular flow. The formulation is presented of a stability problem in cylindrical coordinates. The contribution of fluid viscosity, surface tension, and transverse radius of curvature to the interfacial surface is included. A solution is obtained for Kelvin-Helmholtz instability problem which can be used to determine the critical and most dangerous wavelengths for interfacial waves.

  20. Automating data analysis during the inspection of boiler tubes using line scanning thermography

    Science.gov (United States)

    Ley, Obdulia; Momeni, Sepand; Ostroff, Jason; Godinez, Valery

    2012-05-01

    Failures in boiler waterwalls can occur when a relatively small amount of corrosion and loss of metal have been experienced. This study presents our efforts towards the application of Line Scanning Thermography (LST) for the analysis of thinning in boiler waterwall tubing. LST utilizes a line heat source to thermally excite the surface to be inspected and an infrared detector to record the transient surface temperature increase observed due to the presence of voids, thinning or other defects. In waterwall boiler tubes the defects that can be detected using LST correspond to corrosion pitting, hydrogen damage and wall thinning produced by inadequate burner heating or problems with the water chemistry. In this paper we discuss how the LST technique is implemented to determine thickness from the surface temperature data, and we describe our efforts towards developing a semiautomatic analysis tool to speed up the time between scanning, reporting and implementing repairs. We compare the density of data produced by the common techniques used to assess wall thickness and the data produced by LST.

  1. Demonstration Project Relating to Stress Analysis of SWAGE-Autofrettaged and Re-Autofrettaged Gun Tubes

    National Research Council Canada - National Science Library

    Parker, Anthony P

    2008-01-01

    ... treatment hydraulic re-autofrettage of a swaged tube. APP collaborated with Benet staff in: (a) Predicting permanent OD strains for a re-autofrettaged gun tube manufactured from HB7 steel. and (b...

  2. Alzheimer’s Disease in Social Media: Content Analysis of YouTube Videos

    OpenAIRE

    Tang, Weizhou; Olscamp, Kate; Choi, Seul Ki; Friedman, Daniela B

    2017-01-01

    Background Approximately 5.5 million Americans are living with Alzheimer’s disease (AD) in 2017. YouTube is a popular platform for disseminating health information; however, little is known about messages specifically regarding AD that are being communicated through YouTube. Objective This study aims to examine video characteristics, content, speaker characteristics, and mobilizing information (cues to action) of YouTube videos focused on AD. Methods Videos uploaded to YouTube from 2013 to 20...

  3. Failure analysis of steam generator tubes with dented and wastage configurations

    International Nuclear Information System (INIS)

    Reich, M.; Prachuktam, S.; Gardner, D.; Goradia, H.; Bezler, P.; Kao, K.

    1978-03-01

    The occurrence of PWR steam generator tube cracking, denting, and wastage has been reported in the recent literature. As indicated by its title, this paper concerns itself with the inelastic structural response of the tubes that result from various assumed monotonic as well as cyclic loading conditions, which ultimately could lead to the tube failure

  4. Investigating the Physics Behind VLFEs and LFEs: Analysis Based on Dynamic Rupture Models with Ductile-like Friction

    Science.gov (United States)

    Wu, B.; Oglesby, D. D.; Ghosh, A.; LI, B.

    2017-12-01

    Very low frequency earthquakes (VLFE) and low frequency earthquakes (LFE) are two main types of seismic signal that are observed during slow earthquakes. These phenomena differ from standard ("fast") earthquakes in many ways. In contrast to seismic signals generated by standard earthquakes, these two types of signal lack energy at higher frequencies, and have very low stress drops of around 10 kPa. In addition, the Moment-Duration scaling relationship shown by VLFEs and LFEs is linear(M T) instead of M T^3 for regular earthquakes. However, if investigated separately over a small range magnitudes and durations, the scaling relationship for each is somewhat closer to M T^3, not M T. The physical mechanism of VLFEs and LFEs is still not clear, although some models have explored this issue [e.g., Gomberg, 2016b]. Here we investigate the behavior of dynamic rupture models with a ductile-like viscous frictional property [Ando et al., 2010; Nakata et al., 2011; Ando et al., 2012] on a single patch. In the model's framework, VLFE source patches are characterized by a high viscous damping term η and a larger area( 25km^2), while sources that approach LFE properties have a low viscous damping term η and smaller patch area(<0.5km^2). Using both analytical and numerical analyses, we show how and why this model may help to explain current observations. This model supports the idea that VLFEs and LFEs are distinct events, possibly rupturing distinct patches with their own stress dynamics [Hutchison and Ghosh, 2016]. The model also makes predictions that can be tested in future observational experiments.

  5. Development of the tube bundle structure for fluid-structure interaction analysis model - Intermediate Report -

    International Nuclear Information System (INIS)

    Yoon, Kyung Ho; Kim, Jae Yong; Lee, Kang Hee; Lee, Young Ho; Kim, Hyung Kyu

    2009-07-01

    Tube bundle structures within a Boiler or heat exchanger are laid the fluid-structure, thermal-structure and fluid-thermal-structure coupled boundary condition. In these complicated boundary conditions, Fluid-structure interaction (FSI) occurs when fluid flow causes deformation of the structure. This deformation, in turn, changes the boundary conditions for the fluid flow. The structural analysis have been executed as follows. First of all, divide the fluid and structural analysis discipline, and then independently analyzed each other. However, the fluid dynamic force effect the behavior of the structure, and the vibration amplitude of the structure to fluid. FSI analysis model was separately created fluid and structure model, and then defined the fsi boundary condition, and simultaneously analyzed in one domain. The analysis results were compared with those of the experimental method for validating the analysis model. Flow-induced vibration test was executed with single rod configuration. The vibration amplitudes of a fuel rod were measured by the laser vibro-meter system in x and y-direction. The analyses results were not closely with the test data, but the trend was very similar with the test result. In fsi coupled analysis case, the turbulent model was very important with the reliability of the accuracy of the analysis model. Therefore, the analysis model will be needed to further study

  6. 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.

  7. 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.)

  8. Analysis of radiative heat transfer impact in cross-flow tube and fin heat exchangers

    Directory of Open Access Journals (Sweden)

    Hanuszkiewicz-Drapała Małgorzata

    2016-03-01

    Full Text Available A cross-flow, tube and fin heat exchanger of the water – air type is the subject of the analysis. The analysis had experimental and computational form and was aimed for evaluation of radiative heat transfer impact on the heat exchanger performance. The main element of the test facility was an enlarged recurrent segment of the heat exchanger under consideration. The main results of measurements are heat transfer rates, as well as temperature distributions on the surface of the first fin obtained by using the infrared camera. The experimental results have been next compared to computational ones coming from a numerical model of the test station. The model has been elaborated using computational fluid dynamics software. The computations have been accomplished for two cases: without radiative heat transfer and taking this phenomenon into account. Evaluation of the radiative heat transfer impact in considered system has been done by comparing all the received results.

  9. A RETROSPECTIVE ANALYSIS OF ACUTE APPENDICITIS, RUPTURED APPENDICITIS AND THE LEVEL OF LEUKOCYTOSIS IN PAEDIATRIC SURGICAL PATIENTS OF NELSON MANDELA CENTRAL HOSPITAL.

    Science.gov (United States)

    Mtimba, L; Dhaffala, A; Molaoa, S Z

    2017-06-01

    Appendicectomy is the most commonly performed operation worldwide. The diagnosis is predominantly based on clinical findings. Some patients will clinically be unclear if ruptured or acute inflamed appendicitis; the level of white cell count has been used as the predictor for ruptured appendicitis. This was a retrospective chart review of paediatric surgical patients admitted at Nelson Mandela Central Hospital, Mthatha South Africa. A total of 214 patients with a diagnosis of acute appendicitis. Overall, the ruptured appendicitis was 62% and 38% were inflamed appendicitis. Nature of the acute appendicitis: White cell count, Inflamed, Ruptured, Total p-value 30 0 4 4. This study has demonstrated that in patients who are diagnosed with acute appendicitis clinically, the normal white cell count does not necessarily rule out ruptured acute appendicitis. But the risks of ruptured acute appendicitis increase with the increase level of white cell count.

  10. Development of the tube bundle structure for fluid-structure interaction analysis model

    International Nuclear Information System (INIS)

    Yoon, Kyung Ho; Kim, Jae Yong

    2010-02-01

    Tube bundle structures within a Boiler or heat exchanger are laid the fluid-structure, thermal-structure and fluid-thermal-structure coupled boundary condition. In these complicated boundary conditions, Fluid-structure interaction (FSI) occurs when fluid flow causes deformation of the structure. This deformation, in turn, changes the boundary conditions for the fluid flow. The structural analysis discipline, and then independently analyzed each other. However, the fluid dynamic force effect the behavior of the structure, and the vibration amplitude of the structure to fluid. FSI analysis model was separately created fluid and structure model, and then defined the fsi boundary condition, and simultaneously analyzed in one domain. The analysis results were compared with those of the experimental method for validating the analysis model. Flow-induced vibration test was executed with single rod configuration. The vibration amplitudes of a fuel rod were measured by the laser vibro-meter system in x and y-direction. The analyses results were not closely with the test data, but the trend was very similar with the test result. In fsi coupled analysis case, the turbulent model was very important with the reliability of the accuracy of the analysis model. Therefore, the analysis model will be needed to further study

  11. Analysis of the design of an X-ray tube using Monte Carlo

    International Nuclear Information System (INIS)

    Pena V, J. D.; Sosa A, M. A.; Ceron, P. V.; Vallejo, M. A.; Vega C, H. R.

    2017-10-01

    In this paper we present the Monte Carlo analysis of the X-rays produced by a rotating X-ray tube of the Siemens brand that is used in tomographs for clinical use. The work was done with the MCNP6 code with which the tube was modeled and the primary X-ray spectra produced during the interaction of monoenergetic electrons of 130 keV were calculated. The X-ray spectra were obtained by varying some parameters such as: the angle of the anode (15 to 20 degrees), the type of target (Tungsten, Molybdenum and Rhodium) and the thickness of the filter (3, 5, 10 and 15 mm). In order to have a good statistic 10 7 stories were used. Though the estimators f2 and f5 the X-ray spectra and the total fluencies were estimated. This information will be used to calculate the dose absorbed in the lens and the thyroid gland in patients undergoing radio diagnosis procedures. (Author)

  12. Failure Analysis of Cracked FS-85 Tubing and ASTAR-811C End Caps

    International Nuclear Information System (INIS)

    ME Petrichek

    2006-01-01

    Failure analyses were performed on cracked FS-85 tubing and ASTAR-811C and caps which had been fabricated as components of biaxial creep specimens meant to support materials testing for the NR Space program. During the failure analyses of cracked FS-85 tubing, it was determined that the failure potentially could be due to two effects: possible copper contamination from the EDM (electro-discharge machined) recast layer and/or an insufficient solution anneal. to prevent similar failures in the future, a more formal analysis should be done after each processing step to ensure the quality of the material before further processing. During machining of the ASTAR-811FC rod to form end caps for biaxial creep specimens, linear defects were observed along the center portion of the end caps. These defects were only found in material that was processed from the top portion of the ingot. The linear defects were attributed to a probable residual ingot pipe that was not removed from the ingot. During the subsequent processing of the ingot to rod, the processing temperatures were not high enough to allow self healing of the ingot's residual pipe defect. To prevent this from occurring in the future, it is necessary to ensure that complete removal of the as-melted ingot pipe is verified by suitable non-destructive evaluation (NDE)

  13. Analysis and resolution of service water system heat exchanger tube failures at Clinton Power Station

    International Nuclear Information System (INIS)

    Bhayana, G.K.

    1992-01-01

    Microbiologically Influenced (or Induced) Corrosion (MIC) is generally prominent in a hospitable open loop environment with warmer temperatures and low flow or stagnant flow conditions. It is further enhanced by lack of chemical treatment of the cooling medium. Microbiologically induced corrosion is initiated by a metabolic process of the mocroorganisms. The influenced corrosion occurs when the growth of microorganisms create an environment for corrosion to exist by forming an oxygen-barrier or by producing metabolic by-products that attack metal surfaces. heat exchanger tubes, constructed of 90-10 Copper Nickel, located in two emergency Diesel Generators had to be replaced twice in less than two years. lack of effective chemical treatment was determined to be a contributing factor in both of the failures. The first failure was attributed to microbiologically induced corrosion and the second failure to a combination of microbiologically induced and influenced corrosion. This paper discusses the CPS heat exchanger tube failure analysis, the development and implementation of the MIC mitigation plan, various observations and the conclusions rendered

  14. The root caused analysis of leakaged heat exchanger tube; Ursachenanalyse einer Leckage an einem Waermeuebertraegerrohr

    Energy Technology Data Exchange (ETDEWEB)

    Shamsudin, Shaiful Rizam; Salleh, M.A.A. Mohd; Rahmat, Azmi; Anuar, Mohd Arif [Universiti Malaysia Perlis (UniMAP) (Malaysia). Center of Excellence Geopolymer and Green Technology (CEGeoGTech); Harun, Mohd; Zayid, Hafizal [Malaysian Nuclear Agency, Selangor (Malaysia). Industry Technology Div.; Noor, Mazlee Mohd [Universiti Malaysia Perlis (UniMAP) (Malaysia). School of Materials Engineering

    2015-05-01

    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.

  15. Analysis of simulated ECT signals obtained at tubesheet and tube expansion area

    International Nuclear Information System (INIS)

    Song, Sung Chul; Lee, Yun Tai; Jung, Hee Sung; Shin, Young Kil

    2006-01-01

    Steam generator(SG) tubes are expanded inside tubesheet holes by using explosive or hydraulic methods to be fixed in the tubesheet. In the tube expansion process, it is important to minimize the crevice gap between tubesheet and expanded tube. In this paper, absolute and differential signals are predicted by a numerical method for several different locations of tube expansion inside and outside the tubesheet and signal variations due to tubesheet, tube expansion and operating frequency are observed. Results show that low frequency is good for detecting tubesheet location in both types of signals and high frequency is suitable for sizing of tube diameter as well as the detection of transition region. Also learned is that the absolute signal is good for measuring tube diameter, while the differential signal is good for locating the top of tubesheet and both ends of the transition region.

  16. 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.

  17. 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

  18. Criticality analysis of the storage tubes for irradiated fuel elements from the IEA-R1 with the MCNP code

    International Nuclear Information System (INIS)

    Maragni, M.G.; Moreira, J.M.L.

    1992-01-01

    A criticality safety analysis has been carried out for the storage tubes for irradiated fuel elements from the IEA-R1 research reactor. The analysis utilized the MCNP computer code which allows exact simulations of complex geometries. Aiming reducing the amount of input data, the fuel element cross-sections have been spatially smeared out. The earth material interstice between fuel elements has been approximated conservatively as concrete because its composition was unknown. The storage tubes have been found subcritical for the most adverse conditions (water flooding and un-irradiated fuel elements). A similar analysis with the KENO-IV computer code overestimated the KEF result but still confirmed the criticality safety of the storage tubes. (author)

  19. 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

  20. Construction Simulation Analysis of 60m-span Concrete Filled Steel Tube arch bridge

    Science.gov (United States)

    Shi, Jing Xian; Ding, Qing Hua

    2018-06-01

    The construction process of the CFST arch bridge is complicated. The construction process not only affects the structural stress in the installation, but also determines the form a bridge and internal force of the bridge. In this paper, a 60m span concrete filled steel tube tied arch bridge is taken as the background, and a three-dimensional finite element simulation model is established by using the MIDAS/Civil bridge structure analysis software. The elevation of the main arch ring, the beam stress, the forces in hanger rods and the modal frequency of the main arch during the construction stage are calculated, and the construction process is simulated and analyzed. Effectively and reasonably guide the construction and ensure that the line and force conditions of the completed bridge meet the design requirements and provides a reliable technical guarantee for the safe construction of the bridge.

  1. Transient analysis of a U-tube natural circulation steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A J; Kumar, Rajesh; Bhadra, Anu; Chakraborty, G; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    A computer code has been developed, for transient thermal-hydraulic analysis of proposed 500 MWe PHWR steam generator. The transient behaviour of a nuclear power plant is very much dependent on the steam generator performance, as it provides a thermal linkage between the primary and secondary systems. Study of dynamics of steam generator is essential for over all power plant dynamics as well as design of control systems for steam generator. A mathematical model has been developed for the simulation of thermal-hydraulic phenomena in a U tube natural circulation steam generator. Fluid model is based on one dimensional, nonlinear, single fluid conservation equations of mass, momentum, energy and equation of state. This model includes coupled two phase flow heat transfer and natural circulation. The model accounts for both compressibility and thermal expansion effects. The process simulation and results obtained for transients such as step change in load and total loss of feed water are presented. (author). 5 refs., 7 figs.

  2. Analysis of the campaign videos posted by the Third Sector on YouTube

    Directory of Open Access Journals (Sweden)

    C Van-Wyck

    2013-04-01

    Full Text Available Introduction. Web 2.0 social networks have become one of the tools most widely used by the third sector organisations. This research article examines the formal aspects, content and significance of the videos posted by these organisations on YouTube. Methods. The study is based on the quantitative content analysis of 370 videos of this type, with the objective of identifying the main characteristics. Results. The results indicate that this type of videos are characterised by low levels of creativity, the incorporation of a great amount of very clear information, the predominance of explicit content and the use of very similar formats. Conclusions. Based on the research results, it was concluded that these organisations produce campaign videos with predictable messages that rely on homogeneous structures that can be easily classified in two types: predominantly informative and predominantly persuasive.

  3. Analysis of Applicability of Flow Averaging Pitot Tubes in the Areas of Flow Disturbance

    Directory of Open Access Journals (Sweden)

    Pochwała Sławomir

    2016-03-01

    Full Text Available The issues connected with the complex design of various facilities, including up-to-date boiler equipment as well as the ways of organizing the space around them, are the reasons why there is often a lack of room for mounting a flowmeter in accordance with the recommendations of manufacturers. In most cases the problem is associated with ensuring sufficient lengths of straight pipe leading into and out of a flowmeter. When this condition cannot be fulfilled, the uncertainty of measurement increases above the value guaranteed by the manufacturer of the flowmeter. This sort of operation problem has encouraged the authors of this paper to undertake research aimed at the analysis of applicability of averaging Pitot tubes in the areas of flow disturbance.

  4. Depth analysis of mechanically machined flaws on steam generator tubings using multi-parameter algorithm

    International Nuclear Information System (INIS)

    Nam Gung, Chan; Lee, Yoon Sang; Hwang, Seong Sik; Kim, Hong Pyo

    2004-01-01

    The eddy current testing (ECT) is a nondestructive technique. It is used for evaluation of material's integrity, especially, steam generator (SG) tubing in nuclear plants, due to their rapid inspection, safe and easy operation. For depth measurement of defects, we prepared Electro Discharge Machined (EDM) notches that have several of defects and applied multi-parameter (MP) algorithm. It is a crack shape estimation program developed in Argonne National Laboratory (ANL). To evaluate the MP algorithm, we compared defect profile with fractography of the defects. In the following sections, we described the basic structure of a computer-aided data analysis algorithm used as means of more accurate and efficient processing of ECT data, and explained the specification of a standard calibration. Finally, we discussed the accuracy of estimated depth profile compared with conventional ECT method

  5. Accuracy and cost-analysis of placental alpha-microglobulin-1 test in the diagnosis of premature rupture of fetal membranes in resource-limited community settings.

    Science.gov (United States)

    Eleje, George Uchenna; Ezugwu, Euzebus Chinonye; Ogunyemi, Dotun; Eleje, Lydia Ijeoma; Ikechebelu, Joseph Ifeanyichukwu; Igwegbe, Anthony Osita; Okonkwo, John E; Ikpeze, Okechukwu Christian; Udigwe, Gerald Okanandu; Onah, Hyacinth Eze; Nwosu, Betrand Obi; Ezeama, Chukwuemeka Okwudili; Ezenkwele, Eziamaka Pauline

    2015-01-01

    To determine accuracy and costs of placental α-microglobulin-1 (PAMG-1) test compared to standard clinical assessment (SCA) for diagnosing rupture of membranes (ROM). A multicenter double-blind study of consecutive women with symptoms and signs of ROM in Nnamdi Azikiwe University Teaching Hospital, Nnewi and University of Nigeria Teaching Hospital, Enugu, both in south-east Nigeria using SCA for ROM and the PAMG-1 test was done. ROM was diagnosed if two out of three methods from SCA (pooling, positive nitrazine test or ferning) were present and confirmed post-delivery based on presence of any two of these clinical criteria: delivery in 48 h to 7 days, evidence of chorioamnionitis, membranes overtly ruptured at delivery and adverse perinatal outcomes strongly correlated with prolonged PROM. A cost-analysis was also done. The outcome measures included sensitivity, specificity, accuracy and costs for the two tests. Accuracy, sensitivity and specificity for the PAMG-1 test were 97.2%, 97.4% and 96.7%, higher than for SCA which were 83.7%, 87.9% and 70.5%, respectively (P < 0.001). Accuracy of SCA was higher at less than 34 weeks than 34 weeks or more (88.3% vs 81.4%) while the PAMG-1 test performed equally at both gestational age categories (96.1% vs 97.7%). In women without pooling, accuracy of the PAMG-1 test was 96.7%, while it was 40.0% with SCA. Analysis showed that the overall cost of SCA was 45% higher than the PAMG-1 test. This study confirms that the PAMG-1 test has a consistently high diagnostic accuracy at all gestational ages and with equivocal cases of ROM. The PAMG-1 test appears less costly than SCA. © 2014 The Authors. Journal of Obstetrics and Gynaecology Research © 2014 Japan Society of Obstetrics and Gynecology.

  6. Stress analysis of glass-ceramic insulator and molybdenum cylinders in vacuum tube subassembly

    International Nuclear Information System (INIS)

    Spears, R.K.

    1980-01-01

    This study determined the state of stress between molybdenum cylinders and a glass-ceramic insulator of a vacuum tube during cooling when the glass-ceramic coefficient of expansion differed from molybdenum by +-2 x 10 -7 / 0 C. A thermoelastic stress analysis was performed on the vacuum tube subassembly using the finite element method. Two cases, which examined the effect of cooling over a 700 0 C range, were considered. In Case One, the expansion coefficient of the glass-ceramic was 2 x 10 -7 / 0 C less than that of molybdenum while for Case Two, it was 2 x 10 -7 / 0 C greater. For Case One, it was found that the tangential stresses in the insulator were entirely compressive but the maximum principal stresses in the r-z plane were mainly tensile. For Case Two, the tangential stresses were tensile in the insulator as were most of the maximum principal stresses in the r-z plane except for stress in the upper regions of the insulator. The magnitude of the stress at the maximum principal stress location appears to be substantially lower than what has been observed in practice (i.e., cracking of this design had never been a major problem, but it has been observed that if the coefficient of expansion of the glass-ceramic was 2 x 10 -7 / 0 C lower than molybdenum, cracking usually resulted). This analysis showed that the expansion coefficient of the glass-ceramic could be varied quite liberally from molybdenum before the ultimate strength (13,000 lb/in. 2 ) of the glass-ceramic was exceeded

  7. Storing of Extracts in Polypropylene Microcentrifuge Tubes Yields Contaminant Peak During Ultra-flow Liquid Chromatographic Analysis.

    Science.gov (United States)

    Kshirsagar, Parthraj R; Hegde, Harsha; Pai, Sandeep R

    2016-05-01

    This study was designed to understand the effect of storage in polypropylene microcentrifuge tubes and glass vials during ultra-flow liquid chromatographic (UFLC) analysis. One ml of methanol was placed in polypropylene microcentrifuge tubes (PP material, Autoclavable) and glass vials (Borosilicate) separately for 1, 2, 4, 8, 10, 20, 40, and 80 days intervals stored at -4°C. Contaminant peak was detected in methanol stored in polypropylene microcentrifuge tubes using UFLC analysis. The contaminant peak detected was prominent, sharp detectable at 9.176 ± 0.138 min on a Waters 250-4.6 mm, 4 μ, Nova-Pak C18 column with mobile phase consisting of methanol:water (70:30). It was evident from the study that long-term storage of biological samples prepared using methanol in polypropylene microcentrifuge tubes produce contaminant peak. Further, this may mislead in future reporting an unnatural compound by researchers. Long-term storage of biological samples prepared using methanol in polypropylene microcentrifuge tubes produce contaminant peakContamination peak with higher area under the curve (609993) was obtained in ultra-flow liquid chromatographic run for methanol stored in PP microcentrifuge tubesContamination peak was detected at retention time 9.113 min with a lambda max of 220.38 nm and 300 mAU intensity on the given chromatographic conditionsGlass vials serve better option over PP microcentrifuge tubes for storing biological samples. Abbreviations used: UFLC: Ultra Flow Liquid Chromatography; LC: Liquid Chromatography; MS: Mass spectrometry; AUC: Area Under Curve.

  8. Analysis of DC control in double-inlet GM type pulse tube refrigerators for detectors

    Science.gov (United States)

    Du, B. Y.

    2016-10-01

    Pulse tube refrigerators have demonstrated many advantages with respect to temperature stability, vibration, reliability and lifetime among cryo-coolers for detectors. Double-inlet type pulse tube refrigerators are popular in GM type pulse tube refrigerators. The single double-inlet valve may introduce DC flow in refrigerator, which deteriorates the performance of pulse tube refrigerator. One new type of DC control mode is introduced in this paper. Two parallel-placed needle valves with opposite direction named double-valve configuration, instead of single double-inlet valve, are used in our experiment to reduce the DC flow. With two double-inlet operating, the lowest cold end temperature of 18.1K and a coolant of 1.2W@20K have been obtained. It has proved that this method is useful for controlling DC flow of the pulse tube refrigerators, which is very important to understand the characters of pulse tube refrigerators for detectors.

  9. Numerical analysis of concrete-filled tubes with stiffening plates under large deformation axial loading

    OpenAIRE

    Albareda Valls, Albert

    2013-01-01

    Concrete-filled tubes have been increasingly used these recent decades thanks to their improved structural behavior, especially under compression.Concrete filling in these sections improves ¡ts compressive strength thanks to lateral pressure coming from confinement effect provided by the steel tube. At elevated percentages of loading,concrete suffers an important volumetric expansion, which is clearly restricted by the tube. Therefore, the core is subjected to a severe lateral pressure tha...

  10. Fracture mechanics analysis of the steam generator tube after shot peening

    International Nuclear Information System (INIS)

    Shin, Kyu In; Jhung, Myung Jo; Choi, Young Hwan; Park, Jai Hak

    2003-01-01

    One of the main degradation of steam generator tubes is stress corrosion cracking induced by residual stress. The resulting damages can cause tube bursting or leakage of the primary water which contained radioactivity. Primary water stress corrosion crack occurs at the location of tube/tubesheet hard rolled transition zone. In order to investigate the effect of shot peening on stress corrosion cracking, stress intensity factors are calculated for the crack which is located in the induced residual stress field

  11. Stress analysis in the tubes-tubesheet joint of the heat exchanger under hydraulic expansion

    International Nuclear Information System (INIS)

    Sanzi, H.; Carnicer, R.

    1994-01-01

    In the present work, we are presenting the stresses and displacement occurred in the tube/tubesheet joint of a heat exchanger under hydraulic expansion process. During this process a great amount of tubes cracked. An elasto-plastic finite element calculation was carried out in order to determine the exact deformations of the tube-tubesheet joint. The most important conclusions are presented and compared with the obtained by analytical procedures. (author). 2 refs, 11 figs

  12. 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

  13. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    Schulz, K.C.

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K Q due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  14. Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

    International Nuclear Information System (INIS)

    Thomas, J.M.; Cipolla, R.C.; Ranjan, G.V.; Derbalian, G.

    1978-07-01

    In some Pressurized Water Reactor (PWR) steam generators, a corrosion product has formed on the carbon steel support plate in the crevice between the tube and support plate. The corrosion product occupies more volume than the original metal; the tube-to-support plate crevice volume is thus consumed with corrosion product, and further corrosive action results in a radially inward force on the tube and a radially outward force on the corroding support plate. This has resulted in indentation (''denting'') of the tube, accompanied by occasional cracking. Large in-plane deformation and cracking of support plates has also been observed in the most severely affected plants along with some serious side effects, such as deformation and cracking of inner row tube U-bends caused by support plate movement. Mechanical aspects of the tube denting phenomena have been studied using analytical models. The models used ranged from closed form analytical solutions to state-of-the-art numerical elastic-plastic computer program for moderate strains. It was found that tube dents, such as those observed in operating steam generators, are associated with yielding of both the tubes and support plates. Also studied were the stresses in tube U-bends caused by support plate flow slot deformation

  15. Simulation and Analysis of ECT Signals Obtained at Tubesheet and Tube Expansion Area

    International Nuclear Information System (INIS)

    Song, Sung Chul; Lee, Yun Tai; Jung, Hee Sung; Shin, Young Kil

    2006-01-01

    Steam generator (SG) tubes are expanded inside tubesheet holes by using explosive or hydraulic methods to be fixed in a tubesheet. In the tube expansion process, it is important to minimize the crevice gap between expanded tube and tube sheet. In this paper, absolute and differential signals are computed by a numerical method for several different locations of tube expansion inside and outside a tubesheet and signal variations due to tubesheet, tube expansion and operating frequencies are observed. Results show that low frequency is good for detecting tubesheet location in both types of signals and high frequency is suitable for sizing of tube diameter as well as the detection of transition region. Also learned is that the absolute signal is good for measuring tube diameter, while the differential signal is good for locating the top of tubesheet and both ends of the transition region. In the case of mingled anomaly with tube expansion and tubesheet, low frequency inspection is found to be useful to analyze the mixed signal

  16. 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)

  17. 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.

  18. Tube plug

    International Nuclear Information System (INIS)

    Zafred, P. R.

    1985-01-01

    The tube plug comprises a one piece mechanical plug having one open end and one closed end which is capable of being inserted in a heat exchange tube and internally expanded into contact with the inside surface of the heat exchange tube for preventing flow of a coolant through the heat exchange tube. The tube plug also comprises a groove extending around the outside circumference thereof which has an elastomeric material disposed in the groove for enhancing the seal between the tube plug and the tube

  19. Expert system for eddy current signal analysis: non destructive testing of steam generator tubings

    International Nuclear Information System (INIS)

    Benoist, B.

    1991-01-01

    Automatic analysis, by computer, of defect signals in steam generator tubes, based on Eddy current multifrequency technique, is must often inefficient due to pilgrim noise. The first step is to use a method that allows us to eleminate the noise: the adaptative interpolation. Thanks to this method, which ensures reliable data on each channel, the analysis can be realised by taking into account the data corresponding to each basic or mixed channel. By correlating these diverse data, we can class the signals according to two types of defects: single defects (symmetrical), multiple defects (several in the same place). The second step is to use an expert system which allows a reliable diagnosis for whatever family the defect belongs to. According to this classification, analysis is continued and results in the characterization of the defect. The expert system has already been developed with the general purpose application expert system shell SUPER, which is briefly described. The knowledge base (SOCRATE) and the specific tools developed for this application are thoroughly described. The first results obtained with signals corresponding to real defects, that have been recorded in different places, are presented and discussed. The expert system is revealed efficient in all the studied cases, even with signals obtained in very noisy environments [fr

  20. MTHFD1 polymorphism as maternal risk for neural tube defects: a meta-analysis.

    Science.gov (United States)

    Zheng, Jinyu; Lu, Xiaocheng; Liu, Hao; Zhao, Penglai; Li, Kai; Li, Lixin

    2015-04-01

    Recently, the association between methylenetetrahydrofolate dehydrogenase 1 (MTHFD1) G1958A polymorphism and neural tube defects (NTD) susceptibility has been widely investigated; however, the results remained inconclusive. Hence, we conducted a meta-analysis to evaluate the effect of MTHFD1 G1958A polymorphism on NTD. The relative literatures were identified by search of the electronic databases PubMed, MEDLINE, and EMBASE. The extracted data were statistically analyzed, and pooled odds ratios (ORs) with 95 % confidence intervals (CIs) were calculated to estimate the association strength using Stata version 11.0 software. Finally, ten studies met our inclusion criteria, including 2,132/4,082 in NTD infants and controls; 1,402/3,136 in mothers with NTD offspring and controls; and 993/2,879 in fathers with NTD offspring and controls. This meta-analysis showed that, compared with the mothers with GG genotype, the women with AA genotype had an increased risk of NTD in their offspring, with OR values and 95 % CI at 1.39 (1.16-1.68), p < 0.001. Interestingly, fathers with AG genotype had a significant decreased risk of NTD offspring (OR = 0.79, 95 % CI = 0.66-0.94, p = 0.009). However, there was no significant association between the MTHFD1 G1958A polymorphism in NTD patients and the risk of NTD. In conclusion, the present meta-analysis provided evidence of the association between maternal MTHFD1 G1958A polymorphism and NTD susceptibility.

  1. Single cells for forensic DNA analysis--from evidence material to test tube.

    Science.gov (United States)

    Brück, Simon; Evers, Heidrun; Heidorn, Frank; Müller, Ute; Kilper, Roland; Verhoff, Marcel A

    2011-01-01

    The purpose of this project was to develop a method that, while providing morphological quality control, allows single cells to be obtained from the surfaces of various evidence materials and be made available for DNA analysis in cases where only small amounts of cell material are present or where only mixed traces are found. With the SteREO Lumar.V12 stereomicroscope and UV unit from Zeiss, it was possible to detect and assess single epithelial cells on the surfaces of various objects (e.g., glass, plastic, metal). A digitally operated micromanipulator developed by aura optik was used to lift a single cell from the surface of evidence material and to transfer it to a conventional PCR tube or to an AmpliGrid(®) from Advalytix. The actual lifting of the cells was performed with microglobes that acted as carriers. The microglobes were held with microtweezers and were transferred to the DNA analysis receptacles along with the adhering cells. In a next step, the PCR can be carried out in this receptacle without removing the microglobe. Our method allows a single cell to be isolated directly from evidence material and be made available for forensic DNA analysis. © 2010 American Academy of Forensic Sciences.

  2. Flow-induced vibration analysis of Three Mile Island Unit-2 once-through steam generator tubes. Volume 1. Final report

    International Nuclear Information System (INIS)

    Johnson, J.R.; Brown, J.C.; Harris, C.E.; McGuinn, E.J.; Simonis, J.C.; Thoren, D.E.

    1981-06-01

    Tube responses to flow-induced vibration were measured in the top two spans and the tenth span in the B once-through steam generator at Three Mile Island, Unit 2. This program evaluated the effects of flow-induced biration of OTSG tubes during steady-state and transient operation. Twenty-three tubes were instrumented with accelerometers and strain gages in tubes located along the open lane, in the bundle, and at the tenth span. Tube displacements, frequencies, dynamic strains, and mode shapes were determined during steady-state and transient operation. Pressure sensors were installed in the OTSG to measure pressure fluctuations and plant parameters, which were recorded for correlation with tube response. Data analysis results indicate that the steady-state tube response increases with increasing reactor power, with the maximum response (12 mils peak to peak at midspan) at the outer perimeter of the generator in the 16th span

  3. Atomic spectrometry based on metallic tube atomizers heated by flame: Innovative strategies from fundamentals to analysis

    International Nuclear Information System (INIS)

    Arruda, Marco Aurelio Zezzi; Figueiredo, Eduardo Costa

    2009-01-01

    This review describes recent developments in atomic absorption spectrometry using metallic tube atomizers heated by flames. Sample introduction in spray or gaseous form is emphasized, describing some proposed systems for this task and the fundamentals involved in each context. The latest challenges and future possibilities for use of metallic tubes in atomic/mass spectrometry are also considered.

  4. Strength and Stability Analysis of a Single Walled Black Phosphorus Tube under Axial Compression

    OpenAIRE

    Cai, Kun; Wan, Jing; Wei, Ning; Qin, Qinghua

    2016-01-01

    Few-layered black phosphorus materials recently attract much attention due to its special electronic properties. As a Consequence, the nano-tube from a single-layer black phosphorus has been theoretically built. The corresponding electronic properties of such black phosphorus nano-tube were also evaluated numerically.

  5. Numerical analysis of an experimental data base for tubes pulled in flexion

    International Nuclear Information System (INIS)

    Langlois, R.

    1998-01-01

    The aim of this study is the simulation and the interpretation of experimental results about maximal loading that tubes are able to carry. The tubes are products from primary circuit of german power reactors light water moderated boiling and not boiling cooled. The crack propagation is evaluate under loading. (A.L.B.)

  6. A preliminary stability analysis of MYRRHA Primary Heat Exchanger two-phase tube bundle

    Energy Technology Data Exchange (ETDEWEB)

    Castelliti, Diego [Studiecentrum voor kernenergie – Centre d’étude de l’énergie nucléaire (SCK-CEN), Boeretang 200, Mol (Belgium); GeNERG – DIME/TEC, University of Genova, Via all’Opera Pia 15/a, 16145 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, Via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, Via Dodecaneso 33, 16146 Genova (Italy)

    2016-08-15

    the channel. The stability assessment should take in consideration all the possible reactor operational power levels in order to prove the stable behavior under all operational conditions. The tube bundle stability assessment has been carried out by following a similar procedure used for BWR fuel channels, through a specific RELAP5-3D model representing the PHX and able to evaluate the propagation of a density wave in the tube length. A series of suitable boundary conditions, on both primary and secondary side, and perturbation triggers have been foreseen into the model, so to discover all kind of unstable behavior and to dimension the needed orifice to guarantee the flow stability in all operating conditions. The PHX stability analysis is initially performed on the original tube bundle without the adoption of any stabilizing devices, in order to check the natural behavior of the system. The possible adoption and design of an orifice is then conducted on the basis of this preliminary study. The system response against the various types of instabilities, before the introduction of an orifice, is not completely satisfactory: a stable flow is found within certain specific system parameters ranges. After the introduction of a suitable orifice, the system behavior becomes stable under all operating conditions against all types of two-phase flow instabilities.

  7. Three-dimensional quantitative analysis of adhesive remnants and enamel loss resulting from debonding orthodontic molar tubes

    OpenAIRE

    Janiszewska-Olszowska, Joanna; Tandecka, Katarzyna; Szatkiewicz, Tomasz; Sporniak-Tutak, Katarzyna; Grocholewicz, Katarzyna

    2014-01-01

    Aims Presenting a new method for direct, quantitative analysis of enamel surface. Measurement of adhesive remnants and enamel loss resulting from debonding molar tubes. Material and methods Buccal surfaces of fifteen extracted human molars were directly scanned with an optic blue-light 3D scanner to the nearest 2 μm. After 20 s etching molar tubes were bonded and after 24 h storing in 0.9% saline - debonded. Then 3D scanning was repeated. Superimposition and comparison were proceeded and shap...

  8. Experiment and analysis of instability of tube rows subject to liquid crossflow

    International Nuclear Information System (INIS)

    Chen, S.S.; Jendrzejczyk, J.A.

    1981-09-01

    A tube array subjected to crossflow may become unstable by either one or both of the two basic mechanisms: velocity mechanism and displacement mechanism. The significance of these two mechanisms depends on the mass-damping parameter. The velocity mechanism is dominant for tube arrays with a low mass-damping parameter, and the displacement mechanism is dominant for tube arrays with a high mass-damping parameter. This report presents an experimental and analytical investigation of tube rows in liquid crossflow. The main objective is to verify a mathematical model and the transition between the two mechanisms at the intermediate values of mass-damping parameter. Tests of two tube rows are conducted to determine the critical flow velocity as a function of system damping. Experimental and analytical results are found to be in good agreement

  9. 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.

  10. 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

  11. 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 ...

  12. Loading and Contact Stress Analysis on the Thread Teeth in Tubing and Casing Premium Threaded Connection

    Directory of Open Access Journals (Sweden)

    Honglin Xu

    2014-01-01

    Full Text Available Loading and contact stress distribution on the thread teeth in tubing and casing premium threaded connections are of great importance for design optimization, pretightening force control, and thread failure prevention. This paper proposes an analytical method based on the elastic mechanics. This is quite different from other papers, which mainly rely on finite element analysis. The differential equation of load distribution on the thread teeth was established according to equal pitch of the engaged thread after deformation and solved by finite difference method. Furthermore, the relation between load acting on each engaged thread and mean contact stress on its load flank is set up based on the geometric description of thread surface. By comparison, this new analytical method with the finite element analysis for a modified API 177.8 mm premium threaded connection is approved. Comparison of the contact stress on the last engaged thread between analytical model and FEM shows that the accuracy of analytical model will decline with the increase of pretightening force after the material enters into plastic deformation. However, the analytical method can meet the needs of engineering to some extent because its relative error is about 6.2%~18.1% for the in-service level of pretightening force.

  13. 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

  14. 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.)

  15. 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

  16. 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.

  17. Physical and Numerical Analysis of Extrusion Process for Production of Bimetallic Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Misiolek, W.Z.; Sikka, V.K.

    2006-08-10

    Bimetallic tubes are used for very specific applications where one of the two metals provides strength and the other provides specific properties such as aqueous corrosion and carburization, coking resistance, and special electrical and thermal properties. Bimetallic tubes have application in pulp and paper industry for heat-recovery boilers, in the chemical industry for ethylene production, and in the petrochemical industry for deep oil well explorations. Although bimetallic tubes have major applications in energy-intensive industry, they often are not used because of their cost and manufacturing sources in the United States. This project was intended to address both of these issues.

  18. 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.

  19. Double rupture disc experience

    International Nuclear Information System (INIS)

    1979-01-01

    Result of these observations, comparisons and evaluations can be summarized in the following list of concerns regarding the use of double rupture discs coupled to the liquid space of a steam generator that is subjected to a large leak sodium water reaction event. Single rupture disc show delayed collapse characteristics in LLTR Series I and double disc assemblies are presumed to be more complex with additional delay before opening to give pressure relief. Delayed failure increases pressures in the IHTS and must be adequately covered by design requirements. With CRBR design, the first disc may fail only partially reducing the loading on the second disc with the result that relief performance may not meet requirements

  20. Safety of pull-type and introducer percutaneous endoscopic gastrostomy tubes in oncology patients: a retrospective analysis

    Directory of Open Access Journals (Sweden)

    Pelckmans Paul A

    2011-03-01

    Full Text Available Abstract Background Percutaneous endoscopic gastrostomy (PEG allows long-term tube feeding. Safety of pull-type and introducer PEG placement in oncology patients with head/neck or oesophageal malignancies is unknown. Methods Retrospective analysis of 299 patients undergoing PEG tube placement between January 2006 and December 2008 revealed 57 oncology patients. All patients with head/neck or oesophageal malignancy were treated with chemo- and radiotherapy. In case of high-grade stenosis introducer Freka® Pexact PEG tube was placed (n = 24 and in all other patients (n = 33 conventional pull-type PEG tube. Short-term complications and mortality rates were compared. Results Patients' characteristics and clinical status were comparable in both groups. Short-term complications were encountered in 11/24 (48% introducer PEG patients as compared to only 4/33 (12% pull-type PEG patients (P vs. 0/33 (0%, P vs. 3/33 (9%, NS. Finally, 3/24 gastrointestinal perforations (12% resulted from a difficult placement procedure vs. 1/33 (3%, leading to urgent surgical intervention and admission to ICU. Two introducer PEG patients died at ICU, resulting in an overall mortality rate of 8% vs. 0% (P = 0.091. Conclusion The introducer Freka® Pexact PEG procedure for long-term tube feeding may lead to significantly higher complication and mortality rates in patients with head/neck or oesophageal malignancies treated with chemo- and radiotherapy. It is suggested to use the conventional pull-type PEG tube placement in this group of patients, if possible.

  1. 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

  2. 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.

  3. [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.

  4. Heat transfer analysis and effects of feeding tubes arrangement, falling film behavior and backsplash on ice formation around horizontal tubes bundles

    International Nuclear Information System (INIS)

    Sait, Hani Hussain

    2013-01-01

    Highlights: • Ice shape around the tubes. • Effects of accumulation of ice around the tubes. • Effects of parallel and series tubes arrangements. • Effects of ice accumulated around the tube surfaces. • Effects of backsplash on ice formation. - Abstract: Excessive electrical load has recently get a lot of attention from electric companies specially in hot countries like Saudi Arabia, where air-conditioning load represents about 75% from the total electrical load. Energy storage by freezing is one of the methods that used to tackle this issue. Ice is formed around horizontal cold tubes that are subjected to falling film of water. Ice quantity is measured, photographed and studied. In this studied the coolant inside the tubes flows in series tube arrangement. The results are compared with previous study in which parallel arrangement was used. In addition the falling film behavior and the resulted backsplash are also investigated. A mathematical model to predict ice formation around the tube is proposed. Comparison of the results of the model with that of the experiments showed that the agreement between the two is acceptable. The results also show a quite reasonable quantity of ice is formed in a short time and the series arrangement is more efficient than parallel one. The falling film shapes and its backsplash has also affected the ice formation

  5. Use of image analysis on the prediction of coal burnout performance in a drop tube furnace

    Energy Technology Data Exchange (ETDEWEB)

    R. Barranco; M. Cloke; E. Lester [University of Nottingham, Nottingham (United Kingdom). Nottingham Fuel and Energy Centre, SChEME

    2003-07-01

    An experimental investigation in a drop-tube furnace (DTF) into the combustion burnout performance of some South American coals was carried out. The coal samples, mainly from Colombia, were crushed and screened into three size fractions: 53-75 {mu}m, 106-125 {mu}m, and 150-180 {mu}m. These samples were characterised by standard tests along with a specially developed image analysis technique (grey-scale histogram). Pyrolysis of these samples was performed at a temperature of 1300{sup o}C, in a 1% of oxygen in nitrogen atmosphere for 200 ms. The chars obtained were then re-fired in the same apparatus, at the same temperature, at various residence times, in an atmosphere containing 5% of oxygen in nitrogen. The changes in the characteristics of the chars produced were assessed using a number of different techniques including intrinsic reactivity test and automatic char analysis. Despite the fact that all the coals used in this study were vitrinite-rich, variations in char morphology were evident. This demonstrated that it was impossible to assign any one char type to a single maceral group. It was apparent that vitrinite generates a wide range of char types depending upon the rank of the parent coal and on the maceral associations within the coal. In addition, a reactivity parameter, derived from the grey-scale histogram obtained by image analysis of the coal, was found to be important in the prediction of coal combustion behaviour. Some properties of the re-fired chars were compared with morphology and intrinsic reactivity data of the pyrolysed chars. The results showed that the poor burnout of one of the coals was clearly due to the formation of some particular chars during pyrolysis. This confirms the usefulness of high temperature pyrolysis chars as a predictor of burnout performance. 18 refs., 8 figs., 2 tabs.

  6. Mechanistic multidimensional analysis of two-phase flow in horizontal tube with 90 deg elbow

    International Nuclear Information System (INIS)

    Tselishcheva, E.A.; Antal, St.P.; Podowski, M.Z.; Marshall, S.

    2007-01-01

    The development of modeling and simulation capabilities of two-phase flow and heat transfer is very important for the design, operation and safety of nuclear reactors. Whereas a significant progress in this field has been made over the recent years, further advancements are clearly needed for new concepts of advanced (Generation-IV in particular) reactors. Difficulties in analyzing gas/liquid flows are due to the fact that such two-phase mixtures can assume several different flow patterns, each characterized by flow-regime specific interfacial phenomena of mass, momentum and energy transfer. The level of difficulty increases even further in the case of a complex tube geometries and spatial orientations. The purpose of this paper is to discuss the results of the analysis of a two-phase flow in a horizontal pipe with a 90-degree elbow. The overall objective of the present work is the development of a 3-dimensional computational model of a two-phase high-Reynolds number turbulent flow. The overall new model has been encoded in the next-generation Computational Multiphase Fluid Dynamics (CMFD) computer code, NPHASE. The model has been tested parametrically and the results of NPHASE calculations have been compared against experimental data. It has been demonstrated that the proposed model is consistent both physically and numerically, the predictions are in a reasonable agreement with the measurements

  7. Thermal analysis of a transmission line for Traveling Wave Tube TWT

    International Nuclear Information System (INIS)

    Chbiki, Mounir; Laraqi, Najib; Jarno, Jean-François; Herrewyn, Jacques; Silva Botelho, Tony da

    2012-01-01

    A new analytical method has been developed to study the delay line of Traveling Waves Tubes (TWT). Our study is focused on the analysis of the hot lines shrinking phenomenon. In the studied case, unlike brazed configuration, the contact areas are not perfect, resulting in a diminution of the heat transfer process. In this work, we highlight the influence of the macro-constriction on the heat transfer rate in the various parts of a TWT the geometry of which is also relatively complex. We propose in this work an analytical study of the thermal behavior of a transmission line in established regime. First, we determine the individual thermal resistance of each component. Secondly, we estimate the global resistance of the device according to the geometrical parameters and the respective conductivities of the various elements of this line. In this analytical model, we proceed to parametric studies in order to determine the geometrical configurations that will provide the lowest global thermal resistance. We will emphasize the potential gain according to the used materials and the increase of contact areas.

  8. Geological and Seismological Analysis of the 13 February 2001 Mw 6.6 El Salvador Earthquake: Evidence for Surface Rupture and Implications for Seismic Hazard

    OpenAIRE

    Canora Catalán, Carolina; Martínez Díaz, José J.; Villamor Pérez, María Pilar; Berryman, K.R.; Álvarez Gómez, José Antonio; Pullinger, Carlos; Capote del Villar, Ramón

    2010-01-01

    The El Salvador earthquake of 13 February 2001 (Mw 6.6) caused tectonic rupture on the El Salvador fault zone (ESFZ). Right-lateral strike-slip surface rupture of the east–west trending fault zone had a maximum surface displacement of 0.60 m. No vertical component was observed. The earthquake resulted in widespread landslides in the epicentral area, where bedrock is composed of volcanic sediments, tephra, and weak ignimbrites. In the aftermath of the earthquake, widespread dama...

  9. Analysis of coolant flow in central tube of WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Zsiros, G.; Toth, S.; Attila Aszodi, A.

    2011-01-01

    Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

  10. Process analysis of two-layered tube hydroforming with analytical and experimental verification

    International Nuclear Information System (INIS)

    Seyedkashi, S. M. Hossein; Panahizadeh R, Valiollah; Xu, Haibin; Kim, Sang Yun; Moon, Young Hoon

    2013-01-01

    Two-layered tubular joints are suitable for special applications. Designing and manufacturing of two layered components require enough knowledge about the tube material behavior during the hydroforming process. In this paper, hydroforming of two-layered tubes is investigated analytically, and the results are verified experimentally. The aim of this study is to derive an analytical model which can be used in the process design. Fundamental equations are written for both of the outer and inner tubes, and the total forming pressure is obtained from these equations. Hydroforming experiments are carried out on two different combinations of materials for inner and outer tubes; case 1: copper/aluminum and case 2: carbon steel/stainless steel. It is observed that experimental results are in good agreement with the theoretical model obtained for estimation of forming pressure able to avoid wrinkling.

  11. 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)

  12. 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)

  13. Vortex Tube: A Comparison of Experimental and CFD Analysis Featuring Different RANS Models

    Directory of Open Access Journals (Sweden)

    Chýlek Radomír

    2018-01-01

    Full Text Available The Ranque–Hilsch vortex tube represents a device for both cooling and heating applications. It uses compressed gas as drive medium. The temperature separation is affected by fluid flow behaviour inside the tube. It has not been sufficiently examined in detail yet and has the potential for further investigation. The aim of this paper is to compare results of numerical simulations of the vortex tube with obtained experimental data. The numerical study was using computational fluid dynamics (CFD, namely computational code STAR-CCM+. For the numerical study, a three-dimensional geometry model, and various turbulence physics models were used. For the validation of carried out calculations, an experimental device of the vortex tube of identical geometrical and operating conditions was created and tested. The numerical simulation results have been obtained for five different turbulence models, namely Standard k-ε, Realizable k-ε, Standard k-ω, SST k-ω and Reynolds stress model (RSM, were compared with experimental results. The most important evaluation factor was the temperature field in the vortex tube. All named models of turbulence were able to predict the general flow behaviour in the vortex tube with satisfactory precision. Standard k-ε turbulence model predicted temperature distribution in the best accordance with the obtained experimental data.

  14. Systematic review and meta-analysis of initial management of pneumothorax in adults: Intercostal tube drainage versus other invasive methods.

    Directory of Open Access Journals (Sweden)

    Min Joung Kim

    Full Text Available The ideal invasive management as initial approach for pneumothorax (PTX is still under debate. The purpose of this systematic review and meta-analysis was to examine the evidence for the effectiveness of intercostal tube drainage and other various invasive methods as the initial approach to all subtypes of PTX in adults.Three databases were searched from inception to May 29, 2016: MEDLINE, EMBASE, and the Cochrane CENTRAL. Randomised controlled trials that evaluated intercostal tube drainage as the control and various invasive methods as the intervention for the initial approach to PTX in adults were included. The primary outcome was the early success rate of each method, and the risk ratios (RRs were used for an effect size measure. The secondary outcomes were recurrence rate, hospitalization rate, hospital stay, and complications.Seven studies met our inclusion criteria. Interventions were aspiration in six studies and catheterization connected to a one-way valve in one study. Meta-analyses were conducted for early success rate, recurrence rate, hospitalization rate, and hospital stay. Aspiration was inferior to intercostal tube drainage in terms of early success rate (RR = 0.82, confidence interval [CI] = 0.72 to 0.95, I2 = 0%. While aspiration and intercostal tube drainage showed no significant difference in the recurrence rate (RR = 0.84, CI = 0.57 to 1.23, I2 = 0%, aspiration had shorter hospital stay than intercostal tube drainage (mean difference = -1.73, CI = -2.33 to -1.13, I2 = 0%. Aspiration had lower hospitalization rate than intercostal tube drainage, but marked heterogeneity was present (RR = 0.38, CI = 0.19 to 0.76, I2 = 85%.Aspiration was inferior to intercostal tube drainage in terms of early resolution, but it had shorter hospital stay. The recurrence rate of aspiration and intercostal tube drainage did not differ significantly. The efficacy of catheterization connected to a one-way valve was inconclusive because of the small

  15. Data analysis algorithms for flaw sizing based on eddy current rotating probe examination of steam generator tubes

    International Nuclear Information System (INIS)

    Bakhtiari, S.; Elmer, T.W.

    2009-01-01

    Computer-aided data analysis tools can help improve the efficiency and reliability of flaw sizing based on nondestructive examination data. They can further help produce more consistent results, which is important for both in-service inspection applications and for engineering assessments associated with steam generator tube integrity. Results of recent investigations at Argonne on the development of various algorithms for sizing of flaws in steam generator tubes based on eddy current rotating probe data are presented. The research was carried out as part of the activities under the International Steam Generator Tube Integrity Program (ISG-TIP) sponsored by the U.S. Nuclear Regulatory Commission. A computer-aided data analysis tool has been developed for off-line processing of eddy current inspection data. The main objectives of the work have been to a) allow all data processing stages to be performed under the same user interface, b) simplify modification and testing of signal processing and data analysis scripts, and c) allow independent evaluation of viable flaw sizing algorithms. The focus of most recent studies at Argonne has been on the processing of data acquired with the +Point probe, which is one of the more widely used eddy current rotating probes for steam generator tube examinations in the U.S. The probe employs a directional surface riding differential coil, which helps reduce the influence of tubing artifacts and in turn helps improve the signal-to-noise ratio. Various algorithms developed under the MATLAB environment for the conversion, segmentation, calibration, and analysis of data have been consolidated within a single user interface. Data acquired with a number of standard eddy current test equipment are automatically recognized and converted to a standard format for further processing. Because of its modular structure, the graphical user interface allows user-developed routines to be easily incorporated, modified, and tested independent of the

  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. 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...

  18. Analysis of the pumpkin phloem proteome provides insights into angiosperm sieve tube function.

    Science.gov (United States)

    Lin, Ming-Kuem; Lee, Young-Jin; Lough, Tony J; Phinney, Brett S; Lucas, William J

    2009-02-01

    Increasing evidence suggests that proteins present in the angiosperm sieve tube system play an important role in the long distance signaling system of plants. To identify the nature of these putatively non-cell-autonomous proteins, we adopted a large scale proteomics approach to analyze pumpkin phloem exudates. Phloem proteins were fractionated by fast protein liquid chromatography using both anion and cation exchange columns and then either in-solution or in-gel digested following further separation by SDS-PAGE. A total of 345 LC-MS/MS data sets were analyzed using a combination of Mascot and X!Tandem against the NCBI non-redundant green plant database and an extensive Cucurbit maxima expressed sequence tag database. In this analysis, 1,209 different consensi were obtained of which 1,121 could be annotated from GenBank and BLAST search analyses against three plant species, Arabidopsis thaliana, rice (Oryza sativa), and poplar (Populus trichocarpa). Gene ontology (GO) enrichment analyses identified sets of phloem proteins that function in RNA binding, mRNA translation, ubiquitin-mediated proteolysis, and macromolecular and vesicle trafficking. Our findings indicate that protein synthesis and turnover, processes that were thought to be absent in enucleate sieve elements, likely occur within the angiosperm phloem translocation stream. In addition, our GO analysis identified a set of phloem proteins that are associated with the GO term "embryonic development ending in seed dormancy"; this finding raises the intriguing question as to whether the phloem may exert some level of control over seed development. The universal significance of the phloem proteome was highlighted by conservation of the phloem proteome in species as diverse as monocots (rice), eudicots (Arabidopsis and pumpkin), and trees (poplar). These results are discussed from the perspective of the role played by the phloem proteome as an integral component of the whole plant communication system.

  19. A power system design and analysis of carbon nano-tubes field emission displays

    Science.gov (United States)

    Wang, Jong C.; Yao, W. C.

    2006-01-01

    In new generation Flat Panel Displays(FPD), a lot of design methods are being deployed, including OLED, PDP, TFT-LCD, Back Projection and Field Emission Display(FED) etc. These new generation FPDs have their respective pluses and minuses. Each has its selling points and market attractions. But among them, FED principles are most close to that of CRT displays. Not only FEDs are advantageous in their good degree of saturation of color, but also they have excellent contrast, luminance and electricity consumption etc. It has been considered as the main products of future generation FPDs. Japan and countries all over the world are successively proposing and launching related FED products in the fields. This will not only drive the FEDs into a wave of new trends, but also it will be able to replace most of the current FPD products within a short time. In this paper, based on these solid trends, we are determined to put into our resources and efforts to perform research on these important FEDs technologies and products, particularly in Carbon Nano-Tubes FEDs(CNT-FED). Our research group has already performed research on CNT-FED subjects for almost three years. During the course of our research, we have run into a lot of issues and problems. We have made every effort to overcome some of them. This paper performs comparative analysis of three power option for small size (4-inch) CNT-FEDs to drive the FED effects such as the direct current power, pulsed power and sinusoidal power respectively. This paper performs comparative analysis of three power options for small sized CNT-FEDs. It was concluded that the pulsed power option will produce the best results overall among the three power options. It is felt that these data presented can then be referenced and used to design a power system circuit to get an optimum design for better luminance and least power consumption for small sized commercial CNT-FED products.

  20. 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)

  1. Analysis of flow-induced vibration of heat exchanger and steam generator tube bundles using the AECL computer code PIPEAU-2

    International Nuclear Information System (INIS)

    Gorman, D.J.

    1983-12-01

    PIPEAU-2 is a computer code developed at the Chalk River Nuclear Laboratories for the flow-induced vibration analysis of heat exchanger and steam generator tube bundles. It can perform this analysis for straight and 'U' tubes. All the theoretical work underlying the code is analytical rather than numerical in nature. Highly accurate evaluation of the free vibration frequencies and mode shapes is therefore obtained. Using the latest experimentally determined parameters available, the free vibration analysis is followed by a forced vibration analysis. Tube response due to fluid turbulence and vortex shedding is determined, as well as critical fluid velocity associated with fluid-elastic instability

  2. Tubing For Sampling Hydrazine Vapor

    Science.gov (United States)

    Travis, Josh; Taffe, Patricia S.; Rose-Pehrsson, Susan L.; Wyatt, Jeffrey R.

    1993-01-01

    Report evaluates flexible tubing used for transporting such hypergolic vapors as those of hydrazines for quantitative analysis. Describes experiments in which variety of tubing materials, chosen for their known compatibility with hydrazine, flexibility, and resistance to heat.

  3. Gastric-tube versus whole-stomach esophagectomy for esophageal cancer: A systematic review and meta-analysis.

    Directory of Open Access Journals (Sweden)

    Wenxiong Zhang

    Full Text Available To conduct a systematic review and meta-analysis of studies comparing the gastric-tube vs. whole-stomach for esophageal cancer in order to determine the optimal surgical technique of esophagectomy.A comprehensive literature search was performed using PubMed, EMBASE, ScienceDirect, Ovid MEDLINE, Cochrane Library, Web of Science, Google Scholar, and Scopus. Clinical trials that compared the gastric-tube versus whole-stomach for esophageal cancer were selected. The clinical endpoints included anastomotic leakage, anastomotic stenosis, reflux esophagitis, pneumonia, delayed gastric emptying, and thoracic stomach syndrome.A total of 6 articles (1571 patients were included. Compared to the whole-stomach approach, the gastric-tube approach was associated with a lower incidence of reflux esophagitis (95% confidence interval [CI]: 0.16 to 0.81, p = 0.01 and thoracic stomach syndrome (95% CI: 0.17 to 0.55, p < 0.0001. The rates of anastomotic leakage, anastomotic stenosis, pneumonia, and delayed gastric emptying did not significantly differ between the two groups.The gastric-tube esophagectomy is superior to the whole-stomach approach, as it is associated with a lower incidence of postoperative reflux esophagitis and thoracic stomach syndrome. Our findings must be validated in large-scale randomized controlled trials.

  4. Numerical Thermodynamic Analysis of Two-Phase Solid-Liquid Abrasive Flow Polishing in U-Type Tube

    Directory of Open Access Journals (Sweden)

    Junye Li

    2014-08-01

    Full Text Available U-type tubes are widely used in military and civilian fields and the quality of the internal surface of their channel often determines the merits and performance of a machine in which they are incorporated. Abrasive flow polishing is an effective method for improving the channel surface quality of a U-type tube. Using the results of a numerical analysis of the thermodynamic energy balance equation of a two-phase solid-liquid flow, we carried out numerical simulations of the heat transfer and surface processing characteristics of a two-phase solid-liquid abrasive flow polishing of a U-type tube. The distribution cloud of the changes in the inlet turbulent kinetic energy, turbulence intensity, turbulent viscosity, and dynamic pressure near the wall of the tube were obtained. The relationships between the temperature and the turbulent kinetic energy, between the turbulent kinetic energy and the velocity, and between the temperature and the processing velocity were also determined to develop a theoretical basis for controlling the quality of abrasive flow polishing.

  5. Integration of finite element analysis and design of experiments to analyse the geometrical factors in bi-layered tube hydroforming

    International Nuclear Information System (INIS)

    Alaswad, A.; Olabi, A.G.; Benyounis, K.Y.

    2011-01-01

    Tube hydroforming (THF) is a type of unconventional metal forming process in which high fluid pressure and axial feed are used to deform a tube blank in the desired shape. Bi-layered tube hydroforming is suitable to produce bi-layered joints to be used in special applications such as aerospace, oil production and nuclear power plants. In this work, a finite element study along with response surface methodology (RSM) for design of experiment (DOE) has been used to construct models for three responses namely: bulge height, thickness reduction, and wrinkle height as a function of geometrical factors for X shape bi-layered tube hydroforming. A finite element model was built and experimentally validated. The models developed using finite element analysis (FEA) and RSM was found to be educated. The factors effect and their interactions on the three responses were determined and discussed. Such integration was proved to be a successful technique that can be used to predict the geometry of the hydroformed part.

  6. The effectiveness of different interventional methods for partial fallopian tube obstruction: an analysis of 186 cases

    International Nuclear Information System (INIS)

    Tan Yiqing; Wang Yase; Dai Hongxiu; Li Haitao; Deng Yi; Xiong Liqin

    2011-01-01

    Objective: To evaluate selective salpingo-catheterization recanalization therapy in treating partial fallopian tube obstruction through comparing its clinical effectiveness with that of non-interventional radiology methods. Methods: During the period from January 2008 to October 2010, a total of 186 infertility women with partial fallopian tube obstruction, which was confirmed with hysterosalpingography, were encountered in authors' hospital. This study protocol was approved by our hospital ethics committee, and informed consent was obtained from all patients. According of different treatment methods, 186 patients were divided into two groups. Patients (n=78) in group A received non-interventional radiology methods, including hydrotubation, enema, laparoscopy, physical therapy and traditional Chinese medication, while patients (n=108) in group B received selective salpingo-catheterization recanalization therapy. All 186 patients were followed up for more than six months. Close observation on the pregnancy incidence after different treatments and fallopian tube patency was carried out. The clinical findings were documented. The results were statistically analyzed by using paired 'x2' test. Results: Half a year after different procedures, in group A the pregnancy rate was 12.82% (n=10), and different degrees of fallopian tube obstruction were found in 31.58% patients. Whereas in group B, during the same period of observation, the pregnancy rate was 58.33% (n=63), and partial occlusion in cornual region was seen in one patient (0.47%). Significant difference in both pregnancy rate and fallopian tube occlusion rate existed between two groups (P<0.05). Conclusion: Because of its minimal invasiveness, high effectiveness and safety, the selective salpingography together with fallopian tube recanalization procedures are well accepted by the patients. The selective salpingo-catheterization recanalization therapy is superior to non-interventional radiology methods in

  7. Effectiveness of Tympanostomy Tubes for Otitis Media: A Meta-analysis.

    Science.gov (United States)

    Steele, Dale W; Adam, Gaelen P; Di, Mengyang; Halladay, Christopher H; Balk, Ethan M; Trikalinos, Thomas A

    2017-06-01

    Tympanostomy tube placement is the most common ambulatory surgery performed on children in the United States. The goal of this study was to synthesize evidence for the effectiveness of tympanostomy tubes in children with chronic otitis media with effusion and recurrent acute otitis media. Searches were conducted in Medline, the Cochrane Central Trials Registry and Cochrane Database of Systematic Reviews, Embase, and the Cumulative Index to Nursing and Allied Health Literature. Abstracts and full-text articles were independently screened by 2 investigators. A total of 147 articles were included. When feasible, random effects network meta-analyses were performed. Children with chronic otitis media with effusion treated with tympanostomy tubes compared with watchful waiting had a net decrease in mean hearing threshold of 9.1 dB (95% credible interval: -14.0 to -3.4) at 1 to 3 months and 0.0 (95% credible interval: -4.0 to 3.4) by 12 to 24 months. Children with recurrent acute otitis media may have fewer episodes after placement of tympanostomy tubes. Associated adverse events are poorly defined and reported. Sparse evidence is available, applicable only to otherwise healthy children. Tympanostomy tubes improve hearing at 1 to 3 months compared with watchful waiting, with no evidence of benefit by 12 to 24 months. Children with recurrent acute otitis media may have fewer episodes after tympanostomy tube placement, but the evidence base is severely limited. The benefits of tympanostomy tubes must be weighed against a variety of associated adverse events. Copyright © 2017 by the American Academy of Pediatrics.

  8. 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

  9. Seismic analysis for shroud facility in-pile tube and saturated temperature capsules

    International Nuclear Information System (INIS)

    Iimura, Koichi; Yamaura, Takayuki; Ogawa, Mitsuhiro

    2009-07-01

    At Oarai Research and Development Center, Japan Atomic Energy Agency (JAEA), the plan of repairing and refurbishing Japan Materials Testing Reactor (JMTR) has progressed in order to restart JMTR operation in the fiscal 2011. As a part of effective use of JMTR, the neutron irradiation tests of LWR fuels and materials has been planned in order to study their soundness. By using Oarai Shroud Facility (OSF-1) and Fuel Irradiation Facility with the He-3 gas control system for power lamping test using Boiling Water Capsules (BOCA Irradiation Facility), the irradiation tests with power ramping will be carried out to study the soundness of fuel under LWR Transient condition. OSF-1 is the irradiation facility of shroud type that can insert and eject the capsule under reactor operation, and is composed of 'In-pile Tube', 'Cooling system' and 'Capsule exchange system'. BOCA Irradiation Facility is the facility which simulates irradiation environment of LWR, and is composed of 'Boiling water Capsule', 'Capsule control system' and 'Power control system by He-3'. By using Saturated temperature Capsules and the water environment control system, the material irradiation tests under the water chemistry condition of LWR will be carried out to clarify the mechanism of IASCC. In JMTR, these facilities are in service at the present. However, the detailed design for renewal or remodeling was carried out based on the new design condition in order to be correspondent to the irradiation test plan after restart JMTR operation. In this seismic analysis of the detailed design, each equipment classification and operating state were arranged with 'Japanese technical standards of the structure on nuclear facility for test research' and 'Technical guidelines for seismic design of nuclear power plants on current, and then, stress calculation and evaluation were carried out by FEM piping analysis code 'SAP' and structure analysis code 'ABAQUS'. About the stress of the seismic force, it was proven

  10. Upper Airway Volume Segmentation Analysis Using Cine MRI Findings in Children with Tracheostomy Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Fricke, Bradley L.; Abbott, M. Bret; Donnelly, Lane F.; Dardzinski, Bernard J.; Poe, Stacy A.; Kalra, Maninder; Amin, Raouf S.; Cotton, Robin T. [Cincinnati Children' s Hospital Medical Center, Cincinnati (United States)

    2007-12-15

    The purpose of this study is to evaluate the airway dynamics of the upper airway as depicted on cine MRI in children with tracheotomy tubes during two states of airflow through the upper airway. Sagittal fast gradient echo cine MR images of the supra-glottic airway were obtained with a 1.5T MRI scanner on seven children with tracheotomy tubes. Two sets of images were obtained with either the tubes capped or uncapped. The findings of the cine MRI were retrospectively reviewed. Volume segmentation of the cine images to compare the airway volume change over time (mean volume, standard deviation, normalized range, and coefficient of variance) was performed for the capped and uncapped tubes in both the nasopharynx and hypopharynx (Signed Rank Test). Graphical representation of the airway volume over time demonstrates a qualitative increased fluctuation in patients with the tracheotomy tube capped as compared to uncapped in both the nasopharyngeal and hypopharyngeal regions of interest. In the nasopharynx, the mean airway volume (capped 2.72 mL, uncapped 2.09 mL, p = 0.0313), the airway volume standard deviation (capped 0.42 mL, uncapped 0.20 mL, p = 0.0156), and the airway volume range (capped 2.10 mL, uncapped 1.09 mL, p = 0.0156) were significantly larger in the capped group of patients. In the hypopharynx, the airway volume standard deviation (capped 1.54 mL, uncapped 0.67 mL, p = 0.0156), and the airway volume range (capped 6.44 mL, uncapped 2.93 mL, p = 0.0156) were significantly larger in the capped tubes. The coefficient of variance (capped 0.37, uncapped 0.26, p = 0.0469) and the normalized range (capped 1.52, uncapped 1.09, p = 0.0313) were significantly larger in the capped tubes. There is a statistically significant change in airway dynamics in children with tracheotomy tubes when breathing via the airway as compared to breathing via the tracheotomy tube.

  11. Upper Airway Volume Segmentation Analysis Using Cine MRI Findings in Children with Tracheostomy Tubes

    International Nuclear Information System (INIS)

    Fricke, Bradley L.; Abbott, M. Bret; Donnelly, Lane F.; Dardzinski, Bernard J.; Poe, Stacy A.; Kalra, Maninder; Amin, Raouf S.; Cotton, Robin T.

    2007-01-01

    The purpose of this study is to evaluate the airway dynamics of the upper airway as depicted on cine MRI in children with tracheotomy tubes during two states of airflow through the upper airway. Sagittal fast gradient echo cine MR images of the supra-glottic airway were obtained with a 1.5T MRI scanner on seven children with tracheotomy tubes. Two sets of images were obtained with either the tubes capped or uncapped. The findings of the cine MRI were retrospectively reviewed. Volume segmentation of the cine images to compare the airway volume change over time (mean volume, standard deviation, normalized range, and coefficient of variance) was performed for the capped and uncapped tubes in both the nasopharynx and hypopharynx (Signed Rank Test). Graphical representation of the airway volume over time demonstrates a qualitative increased fluctuation in patients with the tracheotomy tube capped as compared to uncapped in both the nasopharyngeal and hypopharyngeal regions of interest. In the nasopharynx, the mean airway volume (capped 2.72 mL, uncapped 2.09 mL, p = 0.0313), the airway volume standard deviation (capped 0.42 mL, uncapped 0.20 mL, p = 0.0156), and the airway volume range (capped 2.10 mL, uncapped 1.09 mL, p = 0.0156) were significantly larger in the capped group of patients. In the hypopharynx, the airway volume standard deviation (capped 1.54 mL, uncapped 0.67 mL, p = 0.0156), and the airway volume range (capped 6.44 mL, uncapped 2.93 mL, p = 0.0156) were significantly larger in the capped tubes. The coefficient of variance (capped 0.37, uncapped 0.26, p = 0.0469) and the normalized range (capped 1.52, uncapped 1.09, p = 0.0313) were significantly larger in the capped tubes. There is a statistically significant change in airway dynamics in children with tracheotomy tubes when breathing via the airway as compared to breathing via the tracheotomy tube

  12. CFD Analysis of The Hydraulic Turbine Draft Tube to Improve System Efficiency

    Directory of Open Access Journals (Sweden)

    Chakrabarty Spandan

    2016-01-01

    Full Text Available Demand of the power is increasing day by day with the development of the science and technology. Development of the renewable energy sector has become essential issue at the present situation due to the limited source of the non-renewable energy. Hydro energy power generation sector is superior over the other renewable sector due to the high efficiency, ability to continuous generation and low generation cost. In India a great amount of the power generation is taken care by the hydro power system but still some more potential have unexplored. The efficiency improvement of the hydro turbine system can be done for the new installation or installed system by the improvement in component level. The system can be installed by the state of the art equipment, like modern inlet guide vane (IGV control system, improved design of the runner, IGV system, draft tube, penstock to reduce the loss, hence improve the efficiency. The energy recovery in the draft tube depends on the design of draft tube. In the present work the optimized design of the draft tube shape through computational fluid dynamics (CFD simulation has been carried out in ANSYS FLUENT platform. The design objective of the draft tube is to reduce the flow loss and improve the energy recovery, hence to improve the efficiency.

  13. Transcriptome analysis of tube foot and large scale marker discovery in sea cucumber, Apostichopus japonicus.

    Science.gov (United States)

    Zhou, Xiaoxu; Wang, Hongdi; Cui, Jun; Qiu, Xuemei; Chang, Yaqing; Wang, Xiuli

    2016-12-01

    Tube foot as one of the ambulacral appendages types in Aspidochirote holothurioids, is known for their functions in locomotion, feeding, chemoreception, light sensitivity and respiration. In this study, we explored the characteristic of transcriptome in the tube foot of sea cucumber (Apostichopus japonicus). Our results showed that among 390 unigenes which specifically expressed in the tube foot, 190 of them were annotated. Based on the assembly transcriptome, we found 219,860 SNPs from 34,749 unigenes, 97,683, 53,624, 27,767 and 40,786 were located in CDSs, 5'-UTRs, 3'-UTRs and non-CDS separately. Furthermore, 12,114 SSRs were detected from 7394 unigenes. Target genes of four specifically expressed miRNAs (miR-29a, miR-29b, miR-278-3p and miR-2005) in tube foot were also predicted based on the transcriptome, which contain immune-related factors (MBL, VLRA, AjC3, MyD88, CFB), skin pigmentation (MITF), candidate regeneration factor (TRP) and holothurians autolysis-related factor (CL). These results develop a relatively large number of molecular markers and transcriptome resources, and will provide a foundation for further analyses on the function and molecular mechanisms underlying A. japonicas tube foot. Copyright © 2016 Elsevier Inc. All rights reserved.